LOCKHEED MARTIN ENERGY RESEARCH LIBRARIES | qt | i i \ : ; ! 4 [ h : i b : | R \ i ‘ i ! 4 | : B B i " B i iy 4 4 4 v i 1 3 445k 05155905 ORNL-4191 UC-80 — Reactor Technology Contract No. W-7405-eng-26 MOL TEN-SALT REACTOR PROGRAM SEMIANNUAL PROGRESS REPORT For Period Ending August 31, 1967 M. W. Rosenthal, Program Director R. B. Briggs, Associate Director P. R. Kasten, Associate Director DECEMBER 1967 OAK RIDGE NATIONAL LABORATORY Osk Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U. 5. ATOMIC ENERGY COMMISSION TIN ENERGY AESEARGH LIBRARIES [T 3 4454 0515905 7 i i § i ! i ' b i This report is one of a series of periodic reports in which we describe briefly the progress of the program. Other reports issued in this series are listed below. ORNL-3708 is an especially useful report, because it gives a thorough review of the design and construction and supporting development work for the MSRE. It also describes much of the general technology for molten-salt reactor systems, ORNL.-2474 ORNL.-2626 ORNL-2684 ORNL-2723 ORNL-2799 ORNL-2890 ORNL-2973 ORNL-3014 ORNL-3122 ORNL.-3215 ORNL-3282 ORNL.-3369 ORNL-3419 ORNL-3529 ORNL-3626 ORNL.-3708 ORNL-3812 ORNL-3872 ORNL-3936 ORNL-4037 ORNL-4119 Period Ending January 31, 1958 Period Period Period Period Ending October 31, 1958 Ending January 31, 1959 Ending April 30, 1959 Frnding July 31, 1959 Period Ending October 31, 1959 Periods Ending Jonuary 31 and April 30, Period Period Period Period Period Period Period Ending July 31, 1960 Ending February 28, 1961 Ending August 31, 1961 Ending February 28, 1962 Ending August 31, 1962 Ending January 31, 1963 Ending July 31, 1963 Period Ending January 31, 1964 Period Period Period Period Period Period Ending July 31, 1964 Ending February 28, 1965 Ending August 31, 1965 Ending February 28, 1966 Ending August 31, 1966 Ending February 28, 1967 1960 Contents TN R O DU T N et e ettt et e s s e e be e bt et e oo s e e e oo e e ea e et e e e ae e et s 1 Sl M M A R Y oottt et ettt et et 2ttt nE o e e A A< e et et s e e e e e e e e e e e ae s e e et e e e e e a e e e e nen e e 3 PART 1. MOLTEN-SALT REACTOR EXPERIMENT L. MO RE O E R A T IO S o oo e oo oo e oo oo e e e et 4 e e e ea et s es et 4o e n s £ bbb e 1t e e aeae e e 13 1.1 Chronological Account of Operations and Maintenance ...l e 13 1.2 Reactivity BalanCe ... 19 Balance s At P oWt e e 19 Balances at Zeto PowWer e 21 1.3 Thermal Effects of Operation ... ..ot et 22 Radiation Heating . oot ettt e e e e e ettt e e ettt e e e e 22 Thermal Cycle HISTOIY .o e et e e 22 Temperature MeaSurement. ... .. .o e e e 22 Fuel Salt Afterheat e e 24 1.4 Equipment PerformamOe et e et e e e 25 Heat T rans er o e e 25 MR B O OIS i ittt et e ettt et et e e 26 Radiator EnC oS Ure e e e e e e 26 O GaS Sy S MS i i e ettt e ettt a e 27 Cooling Water SySteMS ..o ettt et e e e 29 Component Cooling SyStRIM ... e e e e n e 29 Salt Pump O11 SySTemMS oo i e e e e e 30 Electrical System............... e e e e 31 HEATOIS ittt et e e e e e e e e e e e e e e e et e e et e e e e et e ea e 31 Control Rods and Drives ... e 31 Salt Samplers ... et ee et Meeeeeeeteiessueseteeetateessatetenehe et eeentn Lt et bt s e e st e e .31 oM A It L. i ettt et 33 2. COMPONENT DEVE L O PMEN T e et e et ettt 36 2.1 Off-Gas SamM P T e e 36 2.2 Remote Maint@nanee ... e e e 36 Preparations for Shutdown After Run 11, 36 Evaluation of Remote Maintenance After Run 11 e 37 Repair of Sampler-Enricher and Recovery of Latch ..., 37 2.3 Decontamination SIS . o e e e 40 111 2.4 Development of a Scanning Device for Measuring the Radiation I.evel of 2.5 3. INSTRUMENTS AND CONTROLS Remote Sources Experiment with a S-curie '37Cs Gamma SOUICE ..ot Gamma Scan of the MSRE Heat Exchanger .. ... Gamma Energy Spectrum Scan of the MSRE Heat Exchanger............................................. BT DS L Mark-2 Fuel Pump ...l UUTTI e Spare Rotaty Elements for MSRE Fuel and Coolant Salt Pumps Stress Tests of Pump Tank Discharge Nozzle Attachment MSRE Oil PUMPS o U 0il Pump Endurance Test 3.1 MSRE Operating ERPeriCnCe ... e e Control System Components ........................oco. e N AT I S T I O S oo e e e e Safety SyStem . 3.2 Contiol System Design ... 4. MSRE REACT OR AN ALY SIS e e B It oAU O IO o e 4.2 Neutron Energy Spectra in MSRE and MSBR ... . 4.3 Other Neutronic Characteristics of MSRE with 233U Fuel .. o 4.4 MSRE Dynamics with 233U Fuel 5. DESIGN 5.1 5.2 5.3 5.4 5.5 5.6 5.7 5.8 6. REACTOR PHYSICS 6.1 Critical Loading, Rod Worth, and Reactivity Coefficients ... Fission Rate and Thermal Flux Spatial Distributions ... ... Reactor-Average Fluxes and Reaction Cross Sections Effect of Circulation on Delayed-Neutron PrecurSors ... Samarium Poisoning Effeets ... Cell Arrangement Reactor Fuel Heat Exchanger Blanket Heat Exchanger Fuel Drain Tanks MSBR Physics Analysis Optimization of Reactor Parameters Useful Life of Moderator Graphite Flux Flattening Temperature Coefficients of Reactivity PART 2. MSBR DESIGN AND DEVELOPMENT .................................................................................................. 41 43 43 45 45 45 46 46 46 47 47 47 47 48 48 50 50 50 54 54 55 55 58 60 61 63 63 71 76 79 79 81 82 82 82 84 37 7. SYSTEMS AND COMPONENTS DEVELOPMENT ... oo e 90 7.1 Noble-Gas Behavior in the MSBR .. e 90 7.2 MSBR Fuel Cell Operation with Molten Salt ...t e 95 7.3 Sodium Fluoroborate Circulating Loop Test ... e et e e e 95 T4 MSBR PUIIDS oottt cie et ettt ee e et e ettt et e et e e e ettt e er et s s et e e et e s en e s e e e 96 Survey of Pump Experience Circulating Liquid Metals and Molten Salts ... 96 Introduction of MSBR Pump Program. ... ..ot e et e e 96 Fuel and Blanket Salt Pumps oo ettt et ee e e e ens 97 Coolant Salt PUMDPS ..ot e ettt et e e et £t ane s 99 Water Pump Test Facility . ... e 99 Molten-Salt Bearing TeSES ..ottt s s ra e e e e e e eeneae s s 100 Rotor-Dynamics Feasibility Investigation ... 100 Other Molten-5alt PUIPS ... ettt et s e ae s st et e e oo e ers e e e e e e e st ee s e s mnmeeiaee e s 101 PART 3. CHEMISTRY 8. CHEMISTRY OF THE MORE L. ittt e e s e ee e et s e aene e e ce e crien e 102 8.1 MSRE Salt Composition and PUrity .. ...t e ee e e e 102 L GBIt i e et e e b e e et e et aenene e 103 Co01ant Sall Lo e e e e e e taa e e e e e e e et e en e naaae o 103 ISR Sat oo et ettt ba ettt e 2 e st et nnae e enea 103 Implications of Current Experience in Future Operations ... .. 108 8.2 MSRE Fuel Circuit Corrosion ChemiStry ...ttt 110 8.3 Adjustment of the UF, Concentration of the Fuel Salt ... 110 9. FISSION PRODUCT BEHAVIOR IN THE MSRE ... s 116 9.1 Fission Products in MSRE Cover Gas .. ... 116 9.2 Fission Products in MSRE Fuel ..o e 119 9.3 Examination of MSRE Surveillance Specimens After 24,000 Mwhr ... 121 Examination of Graphite ... e 121 Examination of Hastelloy N ..o e e s 124 Fission Product Distribution in MSRE . et e 125 9.4 Deposition of Fission Products from MSRE Gas Stream on Metal Specimens................. 128 9.5 Deposition of Fission Products on Graphites in MSRE Pump Bowl ... 131 10. STUDIES WITH LiF-BeF , MELTS ..o 136 10.1 Oxide Chemistry of ThFd—UF‘4 MBS et e e et e 136 10.2 Containment of Molten Fluorides in SiliCaA. ...t st e 137 CREMIESEIY .ottt ettt e e et e s e e s ek et 137 Spectrophotometric Measurements with Silica Cells ... 139 10.3 Electrical Conductivity of Molten Fluorides and Fluoroborates ... 140 11. BEHAVIOR OF MOLYBDENUM FLUORIDES ... ... e 142 11.1 Synthesis of Molybdenum Fluorides ... 142 11.2 Reaction of Molybdenum Fluorides with Molten LiF-BeF , Mixtures ... 143 11.3 Mass Spectrometry of Molybdenum Fluorides ... 144 vi 12. SEPARATION OF FISSION PRODUCTS AND OF PROTACTINIUM FROM MOLTEN FLUORIDES 12.1 Extraction of Protactinium from Molten Fluorides into Molten Metals... ... ... . . . 12.2 Stability of Protactinium-Bismuth Solutions Contained in Graphite 12.3 Attempted Electrolytic Deposition of Protactinium 12.4 Protactinium Studies in the High-Alpha Molten-Salt Laboratory Reduction of Iron Dissolved in Molten LikF'- ThF Thorium Reduction in the Presence of an Iron Surface (Brillo Process) Thorium Reduction Followed by Filtration Conclusion 12.5 MSBR Fuel Reprocessing by Reductive Extraction into Molten Bismuth ............................. 12.6 Reductive Extraction of Cerium from LiF-BeF, (66-34 Mole %) into Pb-Bi Eutectic Mixture 13. BEMNAVIOR OF BF, AND FLUOROBORATE MIXTURES 13.1 Phase Relations in Fluoroborate Systems 13.2 Dissociation Vapor Pressures in the NaBF -Nal" System 13.3 Reactions of Fluoroborates with Chromium and Other Hastelloy N Constituents ... Apparent Mass Transfer of Nickel 13.4 Reaction of BF3 with Chromium Metal at 6507 C e e e, 13.5 Compatibility of BF with Gulfspin-35 Pump Oil at 150°F 14. DEVELOPMENT AND EVALUATION OF ANALYTICAL METHODS FOR MOL TEN-SALT REAC T ORS e e e 14.1 Determination of Oxide in MSRE Salts ... 14.2 Determination of U3* in Radioactive Fuel by a Hydrogen Reduction Method ........................ 14.3 In-Line Test Facility . ..., U OPR P 14.4 Electroreduction of Uranium(IV) in Molten LiF-Bel' -ZrF, at Fast Scan Rates and Short TransSition TImMES Lo e e e e e e e bt nan e 14.5 Spectrophotometric Studies of Molten Fluoride Salts ... 14.6 Analysis of Off-Gas from Compatibility Tests of MSRE Pump Oil with BF , ... 14.7 Development of a Gas Chromatograph for the MSRE Blanket Gas ... PART 4. MOLTEN-SALT IRRADIATION EXPERIMENTS 15. MOLTEN-SALT CONVECTION LOOP IN THE ORR ... e 15.1 Loop Description .. ... BT ST ST T U OO UP PP PRSPPI 15, O PEIA IO oo oo e 15.3 Operating TemMPEIALUIES .......cooiitiiiiiiie it e et e e e et 15.4 Salt Circulation by ConveTtIon i e e aeieee et e e e s s s a s a2 aaaassareree e aenaeans 15.5 Nuclear Heat, Neutron Flux, and Salt Power Density ... LS. OTTOSIOM . oo 148 167 16. 17. 18. 19, vii 15.7 Oxygen AnalySiS oo e 182 15.8 Crack in the Core Qutlet Pipe ... e e e 182 15.9 Cutup of Loop and Preparation of Samples ... 184 15.10 Metallographic EXaminalion .. ..o ik e 185 15.11 Isotope Activity Calculation from Flux and Inventory History ... 187 15.12 Isotope Activity BalanCe ... s 190 15.13 Uranium-235 Observed in Graphite Samples.. .. 190 15.14 Penetration of Fission Products into Graphite and Deposition onto Surfaces ... 192 15.15 Gamma Irradiation of Fuel Salt in the Solid Phase ... e 193 PART 5. MATERIALS DEVELOPMENT MSRE SURVEILLANCE PROGRAM L ottt s e e en e e e e e e samn e 196 16.1 General Description of the Surveillance Facility and Observations on Samples Removed ................oein e eeeeaseiutetete e e et eeaaaeante e e et ee 196 16.2 Mechanical Properties of the MSRE Hastelloy N Surveillance Specimens ... 200 GRA P H I TE ST U S et e et e e et e e e et eh et e ee it e bt ee e s et e e s an e eamsearneeaeeeean 208 17.1 Materials Procurement and Property Evaluation ... 208 17.2 Graphite Surface Sealing with MetalS ...t 211 17.3 Gas Impregnation of MSBR Graphites ... e 212 17.4 Trradiation of Graplit@ .. ... ettt 215 HA S T E L LY N ST U S . ittt e et e e ettt e s er e e e e 2 et eab ot e e et a2 etenesebre s s e baeaesaneenns 217 18.1 Improving the Resistance of Hastelloy N to Radiation Damage by Composition MOifiC@tIONS . . oo ittt et e et e et et et ae e e e e s eearasaeneaeeeranees 217 18.2 Aging Studies on Titanium-Modified Hastelloy N ... ..ot et 217 18.3 Phase Identification Studies in Hastelloy N . e ee e eee e ne s 2195 18.4 Hot-Ductility Studies of Zirconium-Bearing Modified Hastelloy N...........ooinn . 221 18.5 Residual Stress Measurements in Hastelloy N WeldsS..........ooo e 223 18.6 COrtOSION SEUAIES ..o oottt e et et e et 226 FE] SLES it ittt e e ettt e e s et e e e n s e e e 226 00 Mt SaIES it ee e e et et e et e et eaeet s ettt e e e et e eataeeae e etan e e e aee e e nnenne 227 Equipment Modifications ... e e e e nans 230 18.7 Titanium Diffusion in Hastelloy N .. e e 230 18.8 Hastelloy N—Tellurium Compatibility ... e 233 GRAPHITE-TO-METAL JOINING oottt et s e e oo 236 19.1 Brazing of Graphite to Hastelloy N et e 236 JOIOE DIESIEIL ..ot et ettt e e e a e e e e st 236 Brazing Development. . ... e et et e e e e, 237 19.2 Compatibility of Graphite-Molybdenum Brazed Joints with Molten Fluoride Salts............... 237 20. 21. 22. 23. 24. 25. 26. viil PART 6. MOLTEN-SALT PROCESSING AND PREPARATION VAPOR-LIQUID EQUILIRRIUM DATA IN MOLTEN-SALT MIXTURES . 239 RELATIVE VOLATILITY MEASUREMENT BY THE TRANSPIRATION METHOD . _..................... 242 DISTILILATION OF MSRE FUEL CARRIER SAL T o e 243 STEADY-STATE FISSION PRODUCT CONCENTRATIONS AND HEAT GENERATION IN AN MSBR AND PROCE SSING P L AN o e e e e e 245 Heat Generation in a Molten-Salt Still ... TR 247 REDUCTIVE EXTRACTION OF RARE EARTHS FROM FUEL SAL T o 248 MODIFICATIONS TO MSRE FUEL PROCESSING FACILITY FOR SHORT DECAY CYCLE ......... 251 PREPARATION OF 233UF4.7L1F FUEL CONCENTRATE FOR THE MSRE .. ... 252 IO G oo 252 Equipment and Operatlons ... e 252 Introduction The objective of the Molten-Salt Reactor Program is the development of nuclear reactors which use fluid fuels that are solutions of fissile and fertile materials in suitable carrier salts. The program is an outgrowth of the effort begun 17 years ago in the Aircraft Nuclear Propulsion (ANP) program to make a molten-salt reactor power plant for aircraft. A molten-salf reactor — the Aircraft Reactor Ex- periment ~ was operated at ORNL in 1954 as part of the ANP program. Our major goal now is to achieve a thermal breeder reactor that will produce power at low cost while simultaneously conserving and extending the nation’s fuel resources. Fuel for this type of re- actor would be ?33UF, or 235UrF'4 dissolved in a salt of composition near 2LiF-BeF . The blanket would be ThF4 dissolved in a carcrier of similar composition. The technology being developed for the breeder is also applicable to advanced con- verter reactors. Our major effort at present is being applied to the operation of the Molten-Salt Reactor Experi- ment {MSRE). This reactor was built to test the types of fuels and materials that would be used in thermal breeder and converter reactors and to pro- vide experience with the operation and maintenance of a molten-salt reactor. The experiment is demon- strating on a small scale the attractive features and the technical feasibility of these systems for large civilian power reactors. The MSRE operates at 1200°F and at atmospheric pressure and pro- duces about 7.5 Mw of heat. Initially, the fuel contains 0.9 mole % UF4, 5 mole % ZrF ,, 29 mole % Ber, and 65 mole % LiF, and the uranium is about 33% ?3°U. The melting point is 840°F. In later operation we expect to use ?3°U in the lower concentration typical of the fuel for a breeder. The fuel circulates through a reactor vessel and an external pump and heat exchange system. All this equipment is constructed of Hastelloy N, a nickel-molybdenum-chromium alloy with exceptional resistance to corrosion by molten fluorides and with high strength at high temperature. The re- actor core contains an assembly of graphite moder- ator bars that are in direct contact with the fuel. The graphite is new material of high density and small pore size. The fuel salt does not wet the graphite and therefore does not enter the pores, even at pressures well above the operating pres- sure. Heat produced in the reactor is transferred to a coolant salt in the heat exchanger, and the coolant salt is pumped through a radiator to dissipate the heat to the atmosphere. A small facility installed in the MSRE building will be used for processing the fuel by treatment with gaseous HF and F,. Design of the MSRE started early in the summer of 1960, and fabrication of equipment began early in 1962. The essential installations were com- pleted and prenuclear testing was begun in August of 1964. Following prenuclear testing and some modifications, the reactor was taken critical on June 1, 1965, and zero-power experiments were completed early in July. After additional modifi- cations, maintenance, and sealing of the contain- ment, operation at a power of 1 Mw began in January 1966. At the 1-Mw power level, trouble was experi- enced with plugging of small ports in control valves in the off-gas system by heavy liquid and varnish-like organic materials. These materials are believed to be produced from a very small amount of oil that leaks through a gasketed seal and into the salt in the tank of the fuel circulating pump. The oil vaporizes and accompanies the gaseous fission products and helium cover gas purge into the off-gas system. There the intense beta radiation from the krypton and xenon poly- merizes some of the hydrocarbons, and the products plug small openings. This difficulty was largely overcoimie by installing a specially designed filter in the off-gas line. Full power — about 7.5 Mw — was reached in May. The plant was operated until the middle of July to the eguivalent of about six weeks at full power, when one of the radiator cooling bloweis — which were left over {rom the ANP program — broke up from mechanical stress. While new blowers were being procured, an airay of graphite and metal surveillance specimens was taken from the core and examined. Power operation was resumed in October with one blower; then in November the second blower was installed, and full power was again attained. After a shutdown to remove salt that had acciden- tally gotten into an off-gas line, the MSRE was operated in December and January at full power for 30 days without interruption. A fourth power run was begun later in January and was continued for 102 days until terminated to remove a second set of graphite and metal specimens. The end of that run came almost a year after full power was first attained. In spite of the time required to replace the blowers, the load factor for that year was 50%. An additional operating period of 46 days during the summer was interrupted for maintenance work on the sampler-enricher when the cable drive mech- anism jammed. The reactor has performed very well in most re- spects: the fuel has been completely stable, the fuel and coolant salts have not corroded the Has- telloy N container material, and there has been no detectable reaction between the fuel salt and the graphite in the core of the reactor. Mechanical difficulties with equipment have been largely con- fined to peripheral systems and auxiliaries. Ex- cept for the small leakage of oil into the pump bowl, the salt pumps have run flawlessly for over 14,000 hr. The reactor has been refueled twice, both times while operating at full power. Because the MSRE is of a new and advanced type, substantial research and development effort is provided in support of the operation. Included are engineering development and testing of reactor components and systems, metallurgical develop- ment of materials, and studies of the chemistry of the salts and their compatibility with graphite and metals both in-pile and out-of-pile. Conceptial design studies and evaluations are being made of large power breeder reactors that use the molten-salt technology. An increasing amount of research and development is being di- rected specifically to the requirements of two- region breeders, including work on materials, on the chemistry of fuel and blanket salts, and on processing methods. Summary PART 1. MOLTEN-SALT REACTOR EXPERIMENT ’ 1. MSRE Operations There were two long runs at full power during this report period. The first, run 11, began in January and lasted into May. After 102 consecutive days of nuclear operation {over 90% of the time at full power), the reactor was shut down to retrieve and replace part of the graphite and metal specimens in the core. The six-week shutdown also included scheduled maintenance and annual tests of con- tainment, instruments, and controls. Run 12 in- cluded 42 days in which the reactor was at full power continuously except for two brief periods after spurious scrams. The run ended when the fuel sampler-enricher drive mechanism jammed, making it inoperative. The reactor was then shut down, the drive was removed, and the sampler latch, which had accidentally been severed from the cable, was retrieved from the fuel pump bowl. During the long runs at high power, interest focused primarily on reactivity behavior and on fuel chemistry. Slow changes in reactivity due to fission product ingrowth and uranium burnup fol- lowed expectations, and no anomalous effect was observed outside the very narrow limits of pre- cision of measurement (£0.02% 5k/k). Over 2 kg of #3°U was added to the fuel during full-power operation. The operation, using the sampler- enricher, demonstrated quick but smooth melting and mixing into the circulating fuel. Six additions of beryllium metal were made to the fuel during operation to maintain reducing conditions in the salt. Corrosion in the salt systems was practically nil, as evidenced by chromium analyses and exami- nation of the core specimens. Studies of the be- havior of certain fission products continued. Component performance, on the whole, was very good. There was no deterioration of heat transfer capability or evidence of unusual heat generation in the reactor vessel. Six thermocouples in the reactor cell began giving anomalous readings during mn 11, but all other themocouples showed no tendency to become less accurate. The new off- gas filter showed no increase in pressure drop and apparently remained quite efficient. Restrictions that built up slowly at the main charcoal bed in- lets were effectively cleared by the use of built-in heaters. While the reactor was down in May for sample removal, two conditions that had existed for some time were remedied: an inoperative posi- tion indicator on a control rod drive and a leaking space cooler in the reactor cell were replaced. Until the sampler failure at the end of run 12, the only delays in the experimental program due to equipment difficulties were brief ones caused by the main blowers and a component cooling pump. A main blower bearing was replaced in run 11, and shortly after the start of run 12 a main blower motor mount was stiffened to alleviate a resonance condition. Also at the start of run 12, low oil pressure made a component coolant pump inopera- tive until the relief valve was replaced. Secondary containment leakape remained well within pre- scribed limits, and there was no leakage from pri- mary systems during operation. During the six- month period, the reactor was critical 2925 hr (66% of the time), and the integrated power in- creased by 2597 to a total of 5557 equivalent full- power hours. 2. Component Development Extensive preparations were made for remote maintenance in the May-June shutdown, including training of 30 craftsmen and foremen. Work pro- ceeded during the shutdown on two shifts. Pro- cedures and tools prepared in advance worked well in replacing cote specimens, repairing a control-rod drive, replacing a reactor cell space cooler, and inspecting equipment in the reactor cell. When the sampler became inoperative, prepara- tions were first made for shielding and containment during replacement of the mechanism and retrieval of the latch. The mechanism was then removed, and a maintenance shield was set up for the latch retrieval. Various long, flexible tools were de- signed and tested in a mockup before use in the sampler tube. The latch was grasped readily, but difficulties were encountered in bringing it up until a tool was designed that enclosed the upper end of the latch. Tools removed from the sampler tube were heavily contaminated, and a shielded carrier with disposable liner was devised to handle them. The sample capsule had broken loose from the latch and cable and was left in the pump bowl after an effort to retrieve it with a magnet failed. The sampler repair and capsule retrieval were accom- plished without spread of contamination and with very moderate radiation exposures. A sampler manipulator was successfully decon- taminated for reuse in a test of decontamination methods. A scheme for mapping and identifying fission product sources remotely was tested in the reactor cell during the May shutdown. A lead-tube colli- mator and an ionization chamber mounted in the movable maintenance shield were vsed to map gamma-ray sources in the heat exchanger and ad- jacent piping; then a collimator and a gamma energy spectrometer were used to characterize the souice at various points. Results were promising. Installation of the off-gas sampler was delayed when the valve manifold had to be rebuilt because of imperfect Monel—stainless steel welds. Stress tests on a Mark-1 pump tank nozzle were completed. Results compared favorably with cal- culated stresses, and the design was judged ade- quate. The Mark-2 replacement fuel pump tank for the MSRE was completed, and preparations for a test with salt proceeded. Oil pumps removed from the MSRE were repaired and tested. A replacement rotary clement for the coolant salt pump was modified by seal welding a mechanical seal that might have become a path for oil leakage to the pump bowl. 3. Instruments and Controls During the May shutdown a complete functional check of instrumentation and control systems was made. Preventive maintenance at that time included modifying 139 relays and replacing capacitors in 33 electronic control modules. The type of com- ponent failures that occurred did not compromise safety or cause excessive inconvenience. Four of the eight neutron chambers were replaced, one because of a short and three because of moisture inleakage. Separate power supplies were installed for each safety channel tc improve continuity of operation and preclude a single compromising failure. Various other modifications to circnits or com- ponents were made to provide more infomation, to improve performance, or to increase protection. 4. MSRE Reactor Analysis As part of planning for future operation of the MSRE, computational studies were made of the neutronic properties of the reactor with 233U in the fuel salt instead of the present %3°U (33% enriched). The neutron energy spectrum was coni- puted and compared in detail with that for a core- lattice design being considered for a molten-salt breeder reactor. The strong similarities indicate that the results of the MSRE experiment will be useful in evaluating design methods for the MSBR. Other computations were made, with the following results. The critical loading will be 33 kg of 233U, compared with 70 kg of 2*°U in the first critical experiment. Control rod worth will be higher by a factor of about 1.3. The important reactivity coefficients will also be considerably larger than with 2°°U fuel. The themmal-neutron flux will be up by more than a factor of 2, and the steady-state samarivm concentmations will consequently be Since more samarium will be left in the salt from 23°U operation, it will act at first as a burn- able poison, causing the reactivity to rise for several weeks despite 233U burmup. Fission power densities and importance functions will be similar to those for 235U fuel. The effective delayed- neutron fraction in the static system will be 0.0026, decreasing to 0.0017 when fuel circulation starts. (Corresponding fractions for 23°U are 0.0067 and 0.0046.) The dynamic behavior with 233U was also ana- lyzed from the standpoint of the inherent stability of the system. Because of the small delayed- neutron fraction, the neutron level responds more sensitively to changes in reactivity, but the re- lower. sponse of the total system is such that the maigins of inherent stability are greater with 233U fuel. PART 2. MSBR DESIGN AND DEVELOPMENT 5. Design The conceptual design wotk on molten-salt breeder reactors during the past six months has been concemed largely with a general advance in the design of cells, containment, piping, and com- ponents, and with stress analysis. In addition, major effort has been devoted to preparation and evaluation of a reactor design in which the average core power density is reduced to 20 kw/liter from the 40 kw/liter we were using during the previous reporting period. At this lower power density the core life before replacement is required would be adequate even if the graphite behavior under irra- diation is no better than that which has been achieved to date. The performance at the lower power density is more nearly representative of current technology, and better perforimance should be achievable as better graphite is developed. Going from 40 to 20 kw/liter increases the capital cost by $6/kwhr (electrical). No new design work was performed on the steam system, but all salt systems (fuel, blanket, and coolant) have been in- vestigated more thoroughly than has been done heretofore. Afterheat removal and thermal shield cooling have been evaluated. 6. Reactor Physics Parametric studies have been carried out which reveal the dependence of MSBR performance on such key design features as the average core power density. They indicate that the power density may be reduced from 80 w/cm? to 20 w/em?® with a penalty not greater than 2%/year in annual fuel yield or 0.1 mill/kwhr (electrical} in power cost. At 20 w/cm? the life of the graphite will be in excess of ten years. Studies of power flattening in the MSBR core show that a maximum-to-average power density ratio of 2 or less can be achieved with no loss of performance. Calculations of tempetaiure ceefficients of re- activity show that the lamge negative component due to fuel expansion is dominant, and yield an overall temperature coefficient of —4.3 x 107 %/°C. 7. Systems and Components Development An analytical model was developed to compute the steady-state migration of noble gases to the graphite and other sinks in the MSBR. Work done to date indicates that the mass tmansfer coefficient from the circulating salt to the graphite is more important than the diffusion ceefficient of xenon in graphite in minimizing the poisoning due to xenon migration to the graphite. In addition, the work has shown that removal of xenon from molten-salt fuels is strongly controlled by the mass transfer coefficient to entrained gas bubbles as well as by the surface area of those bubbles. Studies indicate that the xenon poison fraction in the MSBR is greater than 0.5% with the parameter values con- sidered previously and that the poison fraction may be about 1% with those parameters. The xenon poisoning can be decreased slightly by increasing the surface area of once-through bubbles, decreased significantly by increasing the surface area of re- circulating bubbles, decreased significantly by increasing the mass transfer coefficient to circu- lating bubbles, decreased proportionately by re- ducing the graphite surface area exposed to salt, and decreased significantly if the diffusion co- efficient of xenon into graphite can be decreased to 1077 ft2/hr or less. Similar studies of the after- heat in the graphite from the disintegration of the radioactive noble gases and their decay products show that the afterheat is affected by a variation of the parameters in very much the same manuner as the xzenon poisoning. An experimental program was started to provide an early demonstration of the compatibility of a full-sized graphite fuel cell with a flowing salt stream. The cell will include the graphite-to- graphite and the graphite-to-metal joints. Ag part of a program to qualify sodium fluoro- borate (NaBF4) for use as a coolant for the MSBR, an existing MSRE-scale loop is being prepared to accept NaBF | as the circulating medium under isothermal conditions. The principal alterations are to the cover gas system, to include the equip- ment necessary for handling and controlling the required overpressure of BF ;. The objective will be to uncover any problems associated with the circulation of NaBF , and to devise and test suit- able solutions or corrective measures. A report was issued of a survey of experience with liquid-metal and molten-salt pumps. An ap- proach to producing the breeder salt pumps, which invites the strong participation of U.S. indastry, was evolved. The dynamic response and critical speeds for preliminary layouts of the MSBR fuel salt pump are being calculated, and a survey of fabrication methods applicable to the pump is being made. Preliminary layouts are being made of molten-salt bearing and water pump test facilities for the MSBR fuel salt pump. The pump with the molten-salt bearing was fitted with a new salt bearing and a modified gimbals suppost and was satisfactorily tested with oil. PART 3. CHEMISTRY 8. Chemistry of the MSRE Results of regular chemical analyses of MSRE fuel, coolant, and flush salts showed that after 40,000 Mwhr of power operation generalized corro- sion in the fuel and coolant circuits is practically absent and that the salts are currently as pure as when charged into the reactor. Although statisti- cally satisfactory, fuel composition analyses are much less sensitive to variations in uranium con- centration than is the reactivity balance, and im- proved methods will be required for future MSR fuels whose uranium concentrations need to be only 0.25 that of the MSRE fuel salt. A program for adjusting the relative concentra- tion of U3*/XU to approximately 1.5% by addition of small amounts of beryllium metal to the MSRE fuel was completed. Specimens of fuel salt taken from the pump bowl during this program showed occasional temporary perturbation in the chromium concentration, giving evidence that the identity and concentrations of the phases present at the salt-gas interface of the pump bowl are not neces- sarily typical of the salt in the fuel circuit. 9. Fission Product Behavior in the MSRE A second set of graphite and Hastelloy N long- terrm surveillance specimens, exposed to fissioning molten salt in the MSRE core for 24,000 Mwhr, was examined and analyzed. As for the first set, ex- posed for 7800 Mwhr, examination revealed no evidence of chemical damage to the graphite and metal. Very similar fission product behavior was observed, with heavy deposition of the noble- metal fission products — *"Mo, 132Te, '?%Ruy, 106Ru, 2°Nb, and !'1Ag — on both metal and graphite specimens. A refined method of sampling of the graphite surfaces showed that about 99% of the ?°Mo, ?°Nb, '%3Ru, and '9%Ru was deposited within the outer 2 mils of the surface. By con- trast, appreciable fractions of the !32Te, ?5Zr, 140B4 and %Sy penetrated 50 mils or farther into the graphite. Ten additional exposures of metal specimens in the MSRE pump bowl and five additional samplings of pump bowl cover gas were cariied out. The re- sults from tests under normal operating conditions were similar to those of previous tests; they showed heavy depositions of noble mctals on specimens exposed to the cover gas and the fuel phase. Of special interest were the observations under unusual operating conditions: nearly as much deposition occurred after reactor shutdown with the fuel pump stopped and with the reactor drained as occurred under nommal conditions. Analysis of the time dependence of fission product deposition on Hastelloy N indicated that there was a short-temm rapid process that reached saturation in about 1 min and a long-term process that proceeded at slow constant rate for over 3000 hr. Results from only three exposures of graphite specimens indicated that deposition rate decreased with exposure time for long exposures. 10. Studies with LiF-BreF2 Melts Equilibrium data have been obtained for the reaction U*HE) + Th*™ (o) == Th*™ (1) + U*¥0) ; (XTh)f(XU)o where (f) indicates that the species is dissolved in molten 2LiF . BeF and (o) indicates that the species is in the sparingly soluble oxide solid solution (U, Th)Oz. These data show that, over the interval 0.2 to 0.9 for mole fraction uranium in the oxide phase and 0.01 to 0.07 for mole fraction Th** in the molten fluoride, the equilibrium con- stant is in excess of 1000. Uranium is strongly extracted from the fluoride phase to the oxide solid solution. It seems very likely that protactinium is even more sirongly extracted. If so, equilibration of an LLiF-BeF Z—ThF4-UF 4»PaF4 melt with the prop- er (stable) (U,Th)O2 solid solution should remove protactinium, Recovery of 2?3Pa from a one-region breeder fuel would, accordingly, be possible. Vitreous silica (510 ,) has been shown to be a feasible container material for Lii*fi—BeF2 melts, especially when the system is stabilized by a small overpressure of SiF . Preliminary measure- ments have shown that the solubility of SiF ZLiF - BeF ) is moderately low (about 0.035 mole of 5K per kilogram of melt per atmosphere of SiE ) at 5504C and at least threefold less at 700°C. No ev1dence for silicon oxyfluoudes has been ob- served. [t appears that, at least for temperatures near 500°C and lor short times, an electrically in- sulating and optically transparent coatainer for LiF-BeF , solutions is available. Optical cells of transparent Si0, have been used toy establish, with a Cary model 14M spectrophotom- aeter, that solutions of UF , in 2LiF . Ber under 400 mm of SiF were stable for 48 hr at tempera- tures up to TOOOC These ztudies have led to a considerably more precise definition of molar ab- sorptivity of U*" as a function of temperature and inc ident wavelength than had previously been possible with windowless optical cells. In similar spectrophotometric studies with silica cells, the solubility at 550°C of Cr®” in 2LiF . BeF | was shown to be at least 0.43 mole %. ‘ Silica apparatus has also been shown to be feasible for studies of electrical conductivity of 2L1F-BE.-F2, of the I_Ji.F‘fI'}‘ll:“4 eutectic mixture, and of NaBF4. Preliminary values obtained in this study are to be refined in the near future by use of an improved cell design which will provide a much longer current path length through the melts. 11. Behavior of Molybdenum Fluorides Molybdenum hexafluoride, the only commercially available fluoride of this element, has been used as raw material for preparation of MoF | and MoF .. Direct reduction of Mo¥F . by molybdenum metal in glass appatatus at 30 to 100°C yields, as shown by other investigators, MoF _ of good quality. Dis- proportionation of Mo under vacuum at 2000C yields pure MoF , as the solid residue; we have prapared several samples of the material by this method, which seems not to have been described before. The MoF | reacts on heating with Lil to form at least two binary compounds; the optical and x-ray characteristics of these materials have been determined, but their stoichiome try has not vet been established. Molybdenum hexafluoride has been shown to re- act rapidly with UF _ in Lil"-BeF | solution and with nickel in conta(.t with such qolutmnf-, Molyb- denum trifluoride has been shown to be relatively stable when heated to 700°C under its vapor in sealed capsules of nickel or copper. However, when such heating is done in the presence of 2LiF . BeF , the MoF | reacts readily with nickel, yielding l\IiF‘2 and Mo; the reaction is less marked if the capsule is of copper. Molybdenum trifluoride has been shown to react completely at 500°C with UF3 in LiF . Ber mixtures; the products are UF4 and Mo. Vaporization behavior of MoF | has been shown, by examination with a time-of-flight mass spec- trometer, to be complex and temperature dependent. The behavior observed may suggest that the free energies of formation (per fluorine atom) of these intemediate molybdenum fluorides are so nearly equal that the descriptive chemistry of these sub- stances is dominated by kinetic factors. 12. Separation of Fission Products and of Protactinivm from Molten Fiuorides Very dilute solutions of %%3Pa in bismuth have been shown to be stable for extended periods in graphite containers, but the protactinium appears to be strongly adsorbed upon any added metal or any precipitated phase. More than 90% of the con- tained #?3Pa has been successfully transferred from LiF-BeF ThF blanket mixtures through a molten Bi-Sn metdl phase and recovered in an LiF-NaF-K¥F salt mixture by adding Th reductant to the blanket mixture and oxidant HF to the re- covery salt; successful operation of this experi- mental assembly suggests that a redox transfer process for Pa should be feasible. More concen- trated solutions of ?3'Pa plus 2°*Pa in realistic blanket mixtures continue to be successfully re- duced to insoluble solid material by the addition of thorium metal. Passage of such reduced mix- tures through sintered nickel filters produces a virtually protactinium-free filtrate but fails to localize the Pa in a readily manageable form. Preliminary attempts to reduce 27 'Pa plus 233pa solutions in simulated blanket mixtures to inscluble materials by electrochemical means were unsuccessful; such reduction certainly seems feasible, and the experiments will continue. Rate-earth fluorides in uraninm-free LiF-Bel solutions ate readily reduced to the metallic state and are transferred to the molten bismuth upon contact with a molten alloy of lithium in bismuth. Preliminary evidence suggests that separations of uranium from the rare earths and, perhaps, of uranium from zirconium may be possible by this reductive extraction technique. Material balances on the reductant are poor in experiments to date; this problem will receive additional attention in future experiments. Use of a Pb-Bi alloy with 51 at. % Bi as a substitute for pure bismuth in similar extractions gave generally unsatisfactory results. 13. Behavior of BF; and Fluoroborate Mixtures Recrystallization of NaBF , and of KBF | from dilute (usually 0.5 M) aqueous hydrofluoric acid solutions yields preparations which melt at higher temperatures and which are almost certainly more pure than those reported by previous investigators. These preparations, and our standard differential thermal analysis and quenching techniques, have been used to examine the binary systems Nal'- NaBF and KF-KBF , and the NaBF -KBF , and Nal-KBF joins in the temary system NaF-KF-BF .. The NaF—NaBF4 and the KF-KBF |, systems show single simple eutectics; phase diagrams which we consider to be correct, but which are at variance with data from other laboratories, are presented in this report. Pressures of BF | in equilibrium with NaF-NaBF mixtures over the composition interval 65 to 100 mole % NaBF, have been measured at temperatures of interest to the MSRE. Introduction of chromium metal chips into the system with the NaF-NaBF | eutectic (92 mole % NaBF ) led to perceptible re- action. After the sample bad been above 500°C for 26 hr, the BF | pressure observed was twice that from the melt without added chromium. Subsequent examination of the materials revealed NaCrF _ as one of the reaction products with an additional un- identified black material also present. Other ex- periments with Hastelloy N, iron, and molybdenum showed little or no visual evidence of attack; these tests (for which dissociation pressure was not monitored) did show perceptible weight losses for both the Hastelloy N and iron specimens. In addition, nickel vessels used in the routine de- composition pressure measurements showed shiny interior surfaces, which suggest that some mass transfer had occurred. Boron trifluoride gas has been showin to react at 650°C with essentially pure metallic chromiam in the foim of thin flakes. Weight gain of the chro- mium sample increased linearly with square root of time; x-ray diffraction techniques have revealed the mixed fluoride Cr¥ o CrF , as a reaction product, Gulfspin-35 pump oil (the type used in MSRE) has been exposed for 600 hr at 150°F to helium gas containing 0.1 vol % BF‘E. In these tests the gas mixture was bubbled at 1 liter/min through 1.5 liters of the lubricating oil. Some discoloration of the oil was noted, but there was no distinguishable increase in viscosity. 14. Development and Evaluatien of Analytical Methods for Mclten-Salt Reucters The determination of oxide in highly radioactive MSRE fuel samples was continued. The replace- ment of the moisture monitor cell was the first major maintenance performed since the oxide equip- ment was installed in the hot cell. The U*" concentrations in the fuel samples run to date by the transpiration technique do not re- flect the beryllium additions which have been made to reduce the reactor fuel. This may be accounted for by an interference stemming from the radiolytic generation of fluorine in the fuel samples. This problem will receive further investigation. Experi- mental work is also being carried out to develop a method for the remote measurement of ppm concen- trations of HF in helium or hydrogen gas streams. Design work was continued on the experimental molten-salt test loop which will be used to evaluate electrometric, spectrophotometric, and transpiration methods for the analysis of flowing molten-salt streams. Controlled-potential voltammetric and chrono- potentiometric studies were carried out on the re- duction of U(IV) in molten fluoride salts using a new cyclic voltammeter. It was concluded that the U(IV) - U(III) reduction in molten LiF-Bek - ZrF4 is a reversible one-electron process but that adsorption phenomena must be taken into account for voltammetric measurements at fast scan rates or for chronopotentiometric measurements at short fransition times. An investigation of the spectrum of U(VI) in molten fluoride salts has been initiated. It was found that the spectrum of Na,UF dissolved in LiF-Bel", in an SiO, cell with SiF, overpressure was identical to the spectrum of U0 2F , dissolved under identical conditions. It appears that the equilibrium concentration of 02~ may be sufficient to react with the components of the melt. An at- tempt to use the 510 ,-5iF | system in the spectro- photometiic investigation of electrochemically generated species in molten fluorides also met with difficulties. The SiF, overpressure interferes with cathodic voltammetric studies by causing very high cathodic currents. It is planned to install a spectrophotometric facility with an extended optical path adjacent to a high-radiation-level hot cell to permit the obser- vation of absorption spectra of highly radioactive materials. The basic spectrophotometer and asso- ciated equipment have been ordered. Measurements were made of increases in hydro- carbon concentrations of an He-BF | gas stream after contact with MSRE pump oil. A thermal con- ductivity detector was used to monitor the BF concentration in the test gas stream. Development studies are being made on the de- sign of a gas chromatograph to be used for the continuous determination of sub-ppm, low-ppm, and high concentrations of permanent gas impurities and water in the helium blanket gas of the MSRE. This . problem of analyzing radioactive gas samples prompted the design and construction of an all- metal six-way pneumatically actuated diaphragm valve, A helium breakdown voltage detector with a glass body was designed and constructed to per- mit the observation of the helium discharge. Under optimum conditions, this detector has exhibited a minimum detectable limit below 1 ppb of impurity. It appears to be possible that the detector will also operate in the less-sensitive mode necessary for the determination of high-level concentrations of impurities in the blanket gas. PART 4. MOLTEN-SALT IRRADIATION EXPERIMENTS 15. Molten-Salt Convection Loop in the ORR Irradiation of the second molten-salt convection loop in beam hole HN-1 of the Oak Ridge Research Reactor was terminated after the development of 8.2 x 108 fissions/cc in the 7LiF-BeF2rZrF4«UF4 (65.3-28.2-4.8-1.7 mole %) fuel. Average {uel power densgities up to 150 w per cubic centimeter of salt were attained in the fuel channels of the core of MSRE-grade graphite. The experiment was terminated after radioactivity was detected in the secondary containment systems as a result of gageous fission product leakage from a crack in the core outlet tube. Salt samples were removed routinely during irradiation, and the fuel salt was drained from the loop before removal from the reactor beam hole. Me tallurgical examination revealed a nonductile crack in the Hastelloy N core outlet pipe. The loop was made from unmodified material, and we believe that the failure was caused by loss of strength and ductility under operating conditions of high temperature (™~730°C) and irradiation (™5 x 101° avi). The distribution of various fission products in the system was obtained by the examination of sam- ples of core graphite and loop metal. Some ad- herence of fuel salt to the graphite and entry into cracks in the graphite were found. Molybdenum and tellurium (and probably ruthenium) were largely deposited on graphite and metal surfaces. Other isotopes, including 131y 89¢, 14%p, and 2°Nb, which could have been transported as gases, were found to have penetrated the graphite. Solid MSR fuel salt (LiF-BeF -Z¢F -UF , about £65-28-5-2 mole %) was subjected to very high- intensity gamma irradiation in a spent HFIR fuel element at a temperature of 320°C to determine possible radiation effects on the salt and its com- patibility with graphite and Hastelloy N. Post- irradiation examination did not reveal any signifi- cant effects. PART 5. MATERIALS DEVELOPMENT 16. MSRE Surveillance Program The materials surveillance program for following the changes in the properties of the two major MSRE structural materials — graphite and Hastel- loy N — has been maintained. Graphite and metal specimens were removed for examination on July 28, 1966 (7820 Mwhr), and on May 9, 1967 (32,450 Mwhr). We plan to run various physical and me- chanical property tests an the graphite, but we have not considered thig an urgent item since the doses are quite low (approximately 1 x 10%! neu- trons/cm?, E > 0.18 Mev). Extensive mechanical property tests have been run on the Hastelloy N. Its high-temperature creep-rupture life and rupture ductility were reduced, but these changes are quite comparable with what we have observed for Hastelloy N irradiated in the ORR. There was a slight reduction in the low-temperature ductility, which we attribute to the irradiation-induced pre- cipitation of intergranular M _C. A set of Hastelloy N specimens located outside the reactor core was removed on May 9, 1967, after about 11,000 hr of exposure to the cell environ- ment. There was some surface oxidation, about 0.003 in., but no evidence of nitriding. The surveillance program has been expanded to include some heats of modified Hastelloy N, and specimens that contained 0.5% Ti and 0.4% Zr were removed from the core on May 9, 1967. The me- chanical testing has not been completed, but metal- lographic studies revealed no significant corrosion. 17. Graphite Studies Much of our materials program is directed toward finding suitable materials for future molten-salt reactors. In our present concept of a molten-salt breeder reactor, graphite tubes will be the struc- tural element that separates the fuel and fertile salts. This will require a graphite with very special properties, particularly with respect to a small pore spectrum, low gas permeability, and dimensional stability under high neutron doses. We are looking closely at many grades of graphite that are available from commercial vendors. Sev- eral grades look promising, but none completely satifies our requirements. L.ow gas permeability in graphite seems very hard to obtain, and we feel that producing mono- lithic graphite bodies with helium permeabilities of <10~ % em?/sec will be quite difficult. However, we may be able to satisfy this requiremeat by surface-sealing techniques. Our initial efforts with pyrocarbon and molybdenum sealants look very promising. The proof test will be to demon- strate that graphite sealed in this manner retains its low permeability after neutron exposure. The dimensional instability of graphite continues to be a major problem. We are analyzing very critically all the data obtained to date in an effort to determine what types of graphite appear most stable. We have started our own experiments in the HF[RR, where we can obtain doses of 4 x 1022 nvt (E > 0.18 Mev) in one year. 18. Hastelloy N Studies Although the Hastelloy N will not be in the core, it will be located in peripheral arcas where it will receive rather high doses. We have found that the 10 properties of this basic alloy can be improved sig- nificantly by slight modifications in the composi- Reducing the molybdenum from 16 to 12% suppresses the formation of M C, and small amounts (approximately 0.5%) of either Ti, Zr, or Hf improve the resistance to radiation damage. The titanium-modified alloy looks very good, and we are proceeding further with its development. Experiments are being run to determine the sta- bility of this alloy at elevated temperatures, and specimens aged at 1200 and 1400°F actually show some improvement in ductility. Our electron micros- copy studies show that TiC and Ti O precipitates ‘“solution annealed’ condition. tion. are present in the The changes in distribution and quantity of these precipitates in the aged specimens will be de- terminad., Since titanium can be leached from Hastelloy N by fluoride salts in a manner analogous to chromium, we must consider the corrosion resistance of the titanium-modified alloy. The process is likely controlled by the diffusion rate of titanium in Hastelloy N, and measurements we have made in- dicate that titanium diffuses at a rate comparable with that of chromium at 2000°F. Thus, our small titanium addition will probably not adversely affect the corrosion resistance of the alloy. Our welding studies have shown that Ti and Hf additions to Hastelloy N do not affect the welda- bility adversely, but that Zr is quite detrimental. However, the postirradiation ductility of the Zr- modified alloy is quite high, and we have tried to find a suitable technique for joining this alloy. Since residual stresses from welding can cause dimensional changes and even cracking, we have developed a technique for measuring these stresses. We now can adjust welding parameters and post- weld heat treatments to minimize the magnitude of the residual stresses. We have two thermal convection loops munning which contain an I,,iF-BeFQ-ZrF4—UP‘4-ThF4 fuel salt. One loop is constructed of Hastelloy N and has operated satisfactorily at 1300°F for 47,440 hr. The second loop is constructed of type 3041 stain- less steel with removable hot-leg specimens of the same material. The loop has operated at 1250°F for 36,160 hr, and the removable specimens have indicated a corrosion rate of 2 mils/year. Two loops have also been run using NaF—KF-BFg, which is a possible coolant salt. One loop, constructed of Croloy 9M, plugged in 1440 hr because of mass transfer and the deposition of iron crystals in the cold leg. The second loop, of Hastelloy N, was terminated as scheduled after 8765 hr ot operation, but it was partially plugged and considerable cor- rosion had occurred. Effort is being concentrated on the compatibility of Hastelloy N and the fluoro- borate salts. Of the various fission products that will be pro- duced in the MSBR, tellurium appears to be the only one that may not be compatible with Hastelloy N. We have coated specimens with tellurium and annealed them for long periods of time. There is a very slight penetration of tellurium into the metal, but the mechanical properties are not affected adversely for the conditions investigated. 19. Graphite-to-Metal Joining We are investigating several joint designs for brazing graphite to Hastelloy N. One approach has proven successful, but we are trying to develop a cheaper and simpler type of joint. One promising braze is the 60 Pd-35 Ni--5 Cr alloy, and we have n corrosion tests that confirm its compatibility with molten salts. PART 6. MOLTEN-SALT PROCESSING AND PREPARATION The concept of processing the fuel salt con- tinuously by fluorination and distillation persists esgentially in its initial form. The critical opera- fion in this flowsheet is the distillation of the carrier salt, and most of the effort in this period has been concentrated here. 20. Vapor-Liquid Equilibrium Data in Molten-Salt Mixtures The relative volatilities of ZrF4, NdF‘S, CeF , BaFZ, YFE,, LaFS, and Ser in the ternary system REF -LiF-Bel have been measured using an equilibrium still at 1000°C. Most values are in close agreement with those predicted by Raoult’s law. 11 21. Relative Yolatility Measurement by the Transpiration Method Results from initial experiments using the trans- piration method for measuring the vapor pressure of LiF-BelF over the range 920 to 1055°C con- formed to the correlation of log vapor pressure vs 1/T. These data are also in good agreement with data obtained from equilibrium still measurements. 22. Distillation of MSRE Fuel Carrier Salit Equipment for demonstration of vacuum distilla- tion using MSRE fuel salt has been built and as- sembled in its supporting framework. It is being installed in a test facility to perform nonradioactive experiments. 'FThis unit has been subjected to ex- tensive examination, and numerous dimensional measurements have been taken to afford a reference for postoperational examination. Only if the unit appears to be in good condition after nonradio- active tests will it be installed at the MSRE for carrier salt distillation demonstration. 23. Steady-State Fission Product Concen- trations and Heat Generation in an MSBR and Processing Plant A computer code that considers individual fission products has been prepared to provide information on fission product heat generation in the various components of an MSR processing plant. This pro- gram allows [or the generation and removal of fis- sion products by several different processes which can differ according to their chemical nature. It has been used to compute heat-generation curves for a fuel processing still, and the results com- pare favorably with other programs based on gross fission product heat data. 24. Reductive Extraction of Rare Earths from Fuel Salt One alternative to the distillation process for decontaminating MSBR fuel salt uses the reductive extraction of the rare earths from the salt after uranium has been recovered by fluorination. Ex- periments have been performed using lithium dis- solved in molten bismuth as a reductant. Although the results are complicated by an unexplained loss of metallic lithium to the salt phase, the distribu- tion of rare earths between the salt and metal phases can be correlated with the lithium metal concentration in the metal phase. 25. Modifications to MSRE Fuel Processing Facility for a Short Decay Cycle Provisions are being made for processing the MSRE fuel salt for uranium recovery on the shortest possible cycle after shutdown of the MSRE in early 1968. The flush salt will be processed first, and then the fuel salt will be treated with H -HF to establish the oxygen concentration. Allowing time for these operations, the fuel salt may be fluorinat- ed after 35 days (initial plans called for a 90-day cooling time). This shorter cooling time requires some modification of the processing facility at the MSRE. The higher concentration of iodine requires improvement of the off-gas system, and the pres- 12 ence of molybdenum requires increased shielding around the UF . product absorbers. 26 Preporotion of 233UF4-7LiF Foel Concentrate for the MSRE Refueling and operating the MSRE with 233U fuel early in 1968 is planned; this will require approximately 40 kg of ?°°U as *¥3UF -TLif (27 and 73 mole %) eutectic salt. This fuel concen- trate will be prepared in a cell in the TURF build- ing because of the radiation from the 32U daugh- ters in the 33U, The uranium will arrive as an oxide in cans, which will be opened and dumped into a reaction vessel. Lithium fluoride will be added, and the mixture will be treated with hydro- gen and finally HF to produce the entectic melt. Three 12-kg 23°U batches will be prepared for the major additions to the barren MSRE salt and one 7-kg *#3U batch will be loaded into 60 enriching capsules. The engineering design is almost com- plete, and most of the equipment has been fabri- cated. Part 1. Molten-Salt Reactor Experiment P. N. Haubenreich 1u the six-month period reported here the promise of the MSRE as a practical and reliable reactor was, in a large measure, realized. From the be- ginning, operation of the reactor had strengthened our confidence in the basic technical feasibility of molten-salt reactors. At first, however, me-~ chanical problems with the peripheral equipment did not allow the practical virtues of the molten- salt system to be emphasized by a long period of sustained operation at high power. But the delays were not excessive, and within a year after the first operation at full power, the reactor did com- plete a very satisfactory demonstration of sus- tained operation, Between Jjanuary and May 1967, there were 102 consecutive days of auclear opera- 1. tion with remarkably few difficulties; operation was terminated only because of scheduled re- moval of specimens from the core, The first part of this report details the experience with operation and maintenance of the MSRE. Then it covers development efforts directly related to the reactor. - Finally there is a section relating to a future experiment, namely, the predicted nuclear characteristics of the MSRE with 223U fuel. We plan to strip the present uranium from the fuel salt and replace it with *??U in the spring of 1968 in an experiment that promises to lend worth- while support to design calculations for *3*U- fueled breeder reactors. MSRE Operations P. N. Haubenreich 1.1 CHRONGLOGICAL ACCOUNT OF OPERATIONS AND MAINTENANCE Robert Blumberg C. K. McGlothlan J- L. Crowley R. R. Minue R. H. Guymon M. Richardson P. H. Harley H. C. Rolier T. L. Hudson R. C. Steffy A. 1. Krakoviak B. H. Webster Run 11 began in January and continued into May for 102 consecutive days of nuclear operation (see Fig. 1.1). Between February 1 and May §, the reactor was at full power (7.3 Mw) 93% of the time. The longest interruption in full-power operation was four days, initiated because of excessive vibration in a bearing on a main blower. On this 13 occasion the reactor operated at 5.9 Mw on one blower for a day; then the power was lowered to 10 kw for the beating replacement and was held there for three days to allow the xenon to strip out for a special reactivity measuremeni. Once the power was reduced to 10 kw for 7 hr to permit replacement of the coolant off-gas filter, and once one blower was off for 9 hr after unusual cold (97F) caused bearing vibrations, Twice, spurious scrams caused by false signals produced brief interruptions (1 to 2 hr). Three times the power was lowered for periods from 6 to 38 hr for experimental purposes. In addition to demonstrating the capability of the MSRE for sustained operation, the lengthy period at high power in run 11 afforded usefu! information on long-term reactivity changes due to samarium POWER (Mw) ———— SALT CIRCULATING { FUEL Fig. 1.1. There were no signif- Fuel chemistry, in and other fission products. icant reactivity anomalies. particular the behavior of volatile fission products, was investigated throughout the run by taking aa average of three samples per week from the fuel pump. Twice, a few grams of beryllium was added to the fuel to counteract the tendency of the fis- sjon process to make the salt chemically less re- ducing. An adequate margin against corrosive, oxidizing conditions was maintained, and chromium analyses showed practically no corrosion, Run 11 ended with a scheduled shutdown to re- move core samples. After the fuel system was flushed and cooled, the array of metal and graphite specimens was removed to a hot-cell facility, There the array was disassembled, new samples were substituted for one of the three stringers, and the array was reassembled. Meanwhile, maintenance and inspection were cairied ont on the reactor. Several maintenance and inspection jobs were performed in the reactor cell with semi- remote techniques (see p. 37 on development and evaluation of procedures and tools). One control rod drive was removed for replacement of a posi- tion-indicating device and inspection of the grease, A set of metallurgical specimens adjaceat to the reactor vessel was replaced, and a new americium-curium-beryllium neutron source was installed in the source tube in the thermal shield, One of the two space coolers in the reactor cell was replaced after it proved to be leaking as suspected. The maintenance shield was then set up over the salt heat exchanger to test an ex- perimental device for mapping radiation sources. 14 ORNL-DWGC 57-11785 Outline of MSRE Power Operntion from January to August 1967, Equipment in the reactor cell was viewed in an attempt to determiiie the cause of some anomalous thermocouple readings, and four new thermocouples to read ambient temperature were installed, White dust observed in the reactor cell was analyzed and found to be aluminum oxide, presumably from thermal insulation in the cell, but the source could not be located, After the core samples were reinstalled and the inspection of the cells com- pleted, the reactor and drain-tank cells were sealed on June 9. Other maintenance work at the same time in- cluded overhaul and repair of the radiator door brakes and enclosure, overhaul of the main blower motors, inspection of the main blowers, preventive maintenance on component coolant pump 2, re- placement of a differential pressure element on the fuel off-gas system, and planned modifications and improvements in the instrumentation and con- trol systems. During the shutdown the annual tests of instru- mentation and control systems and secondary con- tainment were conducted. The latter included leak-testing all containment valves and measuring the reactor cell leak rate at 20 psig. The shutdown work was completed ahead of schedule, and nuclear operation in run 12 began on June 19, 39 days after the reactor was taken subcritical at the end of run 11. Run 12 was another period of extended opera- tion at full power. The first week of nuclear operation was marked by difficulties with some of the equipment. These were remedied, however, and there followed 42 days in which the reactor was at full power continuously except for two brief periods following scrams — one accidental and one from loss of normal power due to lightning,. Part of the delay in the lirst week was caused by vibration of a main blower motor. After over- haul the motor had a slight imbalance which would have been acceptable, except that the resonant frequency of the motor mount was very neart the operating speed. Stiffening the mount by welding on reinforcing plates solved this problem. During the first weekend, a component coolant pump lost oil pressure, so it was necessary to switch to the standby. A few hours later the reactor scrammed when lightning knocked out the main power supply and damaged a period safety amplifier. Full- power operation was suspended for two days for modifying the blower motor mount, repaiting the oil system on the component coolant pump, and restoring the safety amplifiers to service. Then began the seven weeks at full power. During the weeks at full power, there were no other eguipment problems that threatened con- tinuity of operation, and interest focused pri- marily on the studies of the fuel salt. Four , additions of beryllium, ranging from 8 to 12 g each, were made in the first three weeks. After the fourth addition, there was an anomalous, temporary rise in chromium concentration in the salt samples, and over the next week ten fuel salt samples were taken to follow the behavior as the chromium con- centration returned to normal. Nex! came a series of uranium additions: 18 capsules in seven days. This brought the 23°U inventory up enough for =ix months of power operation without further additions. Operating for a period of considerable. burnup without refueling will make it possible to determine the capture-to-fission ratio for 225U in the MSRE neutron spectrum from the changes in uranium isotopic ratios. - Run 12 was brought to an end because of dif- ficulties with the fuel sampler-enricher. During an attempt to take a routine 10-g fuel sample on August 5, the cable latch became hung as the capsule was being lowered. There was no external sign of trouble; however, as the cable unreeled, it coiled up in the drive unit housing. Then, as it was being rewound, it tangled in the gears. The exact situation could not be diagnosed, and when the isolation valves between the sampler and the pump bowl were closed, the drive cable was severed just above the latch (see discussion on p. 32). After two days of low-power operation 15 to obtain reactivity data in the absence of xenon, the fuel was drained, and the loop was flushed and cooled down to permit replacement of the sampler mechanism and retrieval of the latch. A temporary containment enclosuwe was erected around the sampler, and a filtered exhaust system was connected to the sampler housing to minimize contamination problems. After the sampler mecha- nism was removed in a shielded carrier to the equipment storage cell, a steel work shield was set up on top of the sampler to permit insertion of retrieval tools down the sampler tube. By this time several long, flexible retrieval {ools had been designed and tested in a mockup (see p- 38). A noose-type tool was used first, but broke because the Jatch was stuck at the latch stop. After an effort to dislodge the tatch, it was enpaged with another noose tool. The latch was still stuck, and 1t was necessary to heat up the pump bowl to loosen it. (Apparently salt mist on the latch stop had frozen the latch in place.) The latch was lifted until it became hung in the tube, and the noose again broke, The latch was picked up again with a corkscrew-type tool, but it pulled loose at the first bend in the sampler tube. Then another tool was designed to slip down over the latch and clutch it with a knob on the end of a cable. Fig- ure 1.2 shows workers atop the shield, inside the enclosure, manipulating this tool onto the latch 20 ft below. The latch was retrieved successfully this time, but as shown in Fig. 1.3, the capsule was missing. After the latch was removed, a go gage was to be inserted to determine whether the tube was clear. It had already been concluded that leaving the capsule in the sample cage in the pump bowl would cause no harm, but it was simple to medify the go gage to house a retractable magnet that could pick up the capsule. When the tool was in- serted, the tube was clear, but the capsule was not found in the cage. Figure 1.4 shows details of the sampler installation in the pump bowl. There is enough cleatance for the capsule to slip out under the ring at the boftom of the cage, but the capsule is then confined by the baffle. The capsule, with a copper body and nickel- plated steel cap, should not deteriorate in the salt, nor are the salt currents strong enough to cause movement and erosion., Therefore, no more efforts were made to remove the capsule, Startup for run 13 then began while a new sampler mechanism was being installed and checked out. 16 PHOTO 88992 Fig. 1.2. Team Fishing for MSRE Sampler Latch. Fig. 1.3. Sampler Latch, Key, and Cable After Retrieval. Analysis and details of operations and main- 17 tenance are given in the sections which follow. Table 1.1, Summary of Some MSRE Operating Statistics IPHOTO 89000 Table 1.1 summarizes some operating statistics. March—August 1967 Total Through Aug. 31, 1967 Critical time, hr Integrated power, Mwhr Equivalent full power hours Salt circulation Fuel loop, hr Coolant loop, hr 2025 (66%) 18,795 2597 (59%) 3024 (68% 3113 (71% 7018 40,307 357 wn 10,361 12,059 18 ORNL-DWG 67-10766 — T ] g LATCH ASSEMBLY CABLE SHEARED OFF 1 |1 APPROXIMATELY AT THIS LOCATION A 11 1 W 1 % L ! g 12 in i LATCH ASSEMBLY NORMAL PQOSITION LATCH STOP. & TOP OF FUEL PUMP 7 —==-BAFFLE NS POSSIBLE MOVEMENT OF CAPSULE\fi SAMPLE CAPSULE / CABLE P L SAMPLE CAPSULE N 5 N NORMAL OPERATING _ /& Ameorienpdmron SALT LEVEL//f“' 1A CAGE FOR L~ N F—— SAMPLE CAPSULE CAPSULF CAGE"" o 74 aLor BAFFLE 7z’ & } SECTION A-A / || A ‘_\\h 2 1HE | ~SAMPLE CAPSULE LOST SAMPLE CAPSULE SALT INTAKE SLOT E3/|6 in. MAX / 1 0 1 2 3 4 INCHES Location of Latch and Sample Capsule in Fuel Pump Bowl. Fig. 1.4. Location of Sampler Latch and Capsules in Fuel Pump Bowl. 1.2 REACTIVITY BALANCE J. R. Engel The extended periods of full-power reactor operation in runs 11 and 12 have provided the most severe tests to date of the on-line reactivity balance calculation. Runs 11 and 12 increased the integrated power by 16,200 and 7650 Mwhr, respectively, to a total of 40,307 Mwhr. In addi- tion to the usual calculations of power- and time- dependent factors, calculations were required in each of these runs to compensate for 233U addi- tions that were made with the reactor at full power, The overall performance of the calcula- tion was highly satisfactory, and no anomalous reactor behavior was indicated at any time. How- ever, some additional calculation modifications were required to eliminate errors that developed. 19 Balances at Power Figures 1.5 and 1.6 summarize the results of the on-line calculations during this report period. These results are reproduced exactly as they were generated, with no corrections for computer- induced errors, For legibility, only about 2% of the data points are shown, but each plotted point ig the result of an individual calculation. Thus the scatfer in the plotted points is an indication of the precision of the calculation. The points at which changes were made to correct errors are indicated by notes. Except for one negative excursion caused by circulating voids immediately after a power shut- down in run 11 (Fig. 1.5), all the calculated values of residual reactivity were between - 0,03 and +.0.10% 8k/k. An apparent gradual decline in ORNL-DWS 67-4044R2 Pedanal I | REACTIVITY Y 8k/k T T .L..u S ot AReL e, N T e g.._u ? . | el S esp ™, nlatne, Mw 0 ‘ 1 0.0 25 * COMPUTER QUT OF SERVICE G5 - L P i b el | Yo 3K/ K - | -005 --1- Lo ] REACTIVITY .......... L] r CORARFCTED KTEMP, KSAM MORFED KXE . o L . o TR e +‘f B et e e e et e et ‘ " Mw 15 i7 MARCH, 1967 % Sk/k 13 15 17 APRIL 1967 Fig. 1.5. Residual Reactivity During 21 23 ), i s | SeaTnas weluns -~ L } P AT e aa ! e ! SSnatre, Y g et St - : ? . . 7 29 1 24 23 MAY, 1367 MSRE Run 11, 015 ciC 0.05 psommai™ag e, o % Bk /k oft——m—n e REACTOR SUBCRITICAL - - . ; o o A8 - E 20 ~ ORNL-DWG &7-9191 REACTIVITY . . . * o8¢ $ 2 e '.fla atat et g n e . P -0.05 -0.0 - - e el 8 {#fi . T T T s ‘ i POWER Z 4 ‘ 2 0| L . b . o i Lo 19 21 23 25 27 29 1 3 5 7 2 15 JUNE, 1967 JULY, 1987 o5 —— - e e s _ REACTIVITY 0.10 : s et e e : SHIFT IN COMPUTER. b tanee o et - READINGS % 005 ‘ o R PP . z O - CHANGE CONTROL? RGD ,,;fwv . -:T-.'.'f"_'j.’f:.."-_-_" "'.‘?“’.’Mfi%{%‘fi{’: ,‘ &2 CONFIGURATION -0.05 . . -o40lL e - : TR 8 — T T T o L i T POWER H‘ = - : . = 4 . 2 . . . . ‘. ol el S _ 16 18 20 22 24 26 28 30 1 3 5 7 2 JULY, 1967 AUG 1967 Fig. 1.6. Residual Recctivity During MSRE Run 12, residual reactivity occurred during the first several weeks of run 11. Detailed analysis of the individual terms revealed two sources of error. One was a gradual downward drift in the tempera- ture indicated by two of the four thermocouples used to calculate the average reactor outlet tem- perature. These two thermocouples were elimi- nated and replaced by one other that had not drifted. The second error was caused by loss of significance in the calculation of the '#?Sm con- centration. In the program, only the change in samarium concentration is computed, and that change is added to the last value to obtain the current value. As the '*?Sm concentration ap- proached 85% of its equilibrium valiue, the incre- mental concentration change computed for the 5- min time step between routine reactivity balances was outside the five-decimal-digit precision of the computer. As a result, these increments were lost when the concentration was updated. To avoid using double-precision arithmetic, the pro- gram was modified to only update the *#%Sm and the !°!Sm concentrations every 4 hr while the reactor is at steady power. Summary calcula- tions made off line were used to verify the ade- quacy of this change. When these corrections were introduced on March 17, the apparent downward drift in re- activity disappeared. At the same time, minor changes were made in some of the !35Xe stripping parameters to make the calculated steady-state xernon poisoning agree more closely with the ob- served value. Other small reactivity variations were observed in run 11, for example, from March 29 to April 9. These changes are directly related to changes in the helium overpressure on the fuel loop; a 1-psi pressure increase leads to a reversible reactivity decrease of slightly less than 0.01% 8k/k. The mechanism through which pressure and reactivity are coupled has not yet been established. The direct reactivity effect of the change in circulat- ing voids caused by a change in absolute pressure is at least a factor of 10 smaller than the observed effect of pressure on reactivity. The time con- stant of the pressure-reactivity effect is relatively long, suggesting a possible connection through the Xenon poisoning. Fuel additions were made for the first time in run 11 with the reactor at full power. Nine cap- sules containing a total of 761 g of 23°U were added between April 18 and 21. The reactivity- balance results during this time show good agree- ment between the calculated and observed effects of the additions. The transient effects of the actual fuel additions were very mild. Figure 1.7 shows an on-line plot of the position of the regulating control rod made during a typical fuel Con- trol rod movement to compensate for the additional uranium in the core started about 30 sec after the fuel capsule reached the pump bowl, and the entire transient was complete about 2 min later. This indicates rapid melting of the enriching salt and quick, even dispersion in the circulating fuel. The weights of the emptied fuel capsules indi- cated that essentially all their contained *33U was transferred to the fuel loop. The reactivity-balance results in run 12 (Fig. 1.6) were essentially the same as those in the addition with the reactor on servo control. preceding run. Minor variations, associated with pressure and power changes, were again observed. Another series of fuel additions at full power was made in this run between July 19 and 26, This series consisted of 18 capsules containing 1527 g of 232U, The purpose of this large addi- tion was to provide sufficient excess uranium so that a large amount of integrated power could be produced without intermediate fuel additions. We plan to perform a detailed evaluation of the uranium isotopic-change effects associated with power operation, and substantial burnup is required to make the analyses of isotopic composition useful. A secondary result of this large fuel addition (0.5% 6k/k) was a drastic change in the control rod configuration. At the end of the additions the separation between the tips of the shim rods ORNL-DWG 710132 ol | 37 - — ’ \“‘MJJ\,\““;‘"‘*/‘ l\fi.f-f‘.f\...‘_r-‘-t'-\/\.\, {inches withdrown) REGULATING RCD POSITION G 1 2 3 4 5 s 7 TIME {min) O=4150 hr APRIL 20,1967 Fig. 1.7. Regulating Control Rod Position During Fuel Addition. and that of the regulating rod was 15.5 in., whereas the normal separation has been 4 to 8 in. The variation in apparent residual reactivity as a function of control rod configuration was reexamined, and we observed a decrease of 0.02% 6k/k when the more usual configuration was established. This was consistent with an earlier evaluation (May 1966) of the accuracy of the analytic expression used in the computer to calculate control rod poisoning as a function of rod configuration. On August 3 a computer failure occurred which required recalibration of the analog-signal ampli- fiers after service was restored. As a result of this recalibration, there were small shifts in the values of several of the variables used in the re- activity balance. Errors in reactor-outlet tem- perature and regulating-rod pesition caused a down- ward shift of 0.03% dk/k in the residual reactivity. Balances at Zero Power Figure 1.8 shows the long-term variation in residual reactivity since the start of power opera- tion (December 1965). The values shown are average results at zero power with no xenon present, Corrections have also been applied for computer-induced errors such as those at the end of run 12. The results are plotted to show their relationship to the reactor operating limits at +0.5% Ok/k. The discovery of a 0.5-in, shift in the absolute position of rod 1 at the end of run 12 (see p. 31) adds some uncertainty to the last point in this figure. This shift represents a reactivity effect of 4-0.02% 8k/k, which would have been de- tected if it had occurred during a run. However, the dilution corrections which must be applied ORNI-DWE 67-3803RA o ACTIVITY % 8k/k? o W : . . - os | | v . o e RESIDUAL RE | 400 T | B 16 20 24 28 32 36 40 (x100) INTEGRATED POWER (Mwnr) Fig. 1.8. Long-Term Drift in Residual Reactivity of the MSRE at Zero Power. between runs contain enough uncertainty that an error of this magnitude could be lost. Thus the shift in rod position cannot be assigned to either the beginning or end of run 12. Even with this uncertainty in residual reactivity, the zero-power results fall within a very narrow band, which demonstrates the continuing good performance of both the reactor system and the reactivity-balance calculation. 1.3 THERMAL EFFECTS OF OPERATION C. H. Gabbard Radiation Heuting Reactor Yessel. — The temperature differences between certain thermocouples on the reactor vessel and the reactor inlet temperature are moni- tored by the computer to determine whether there is any evidence of a sedimentation buildup in the lower head or on the core support {lange. In the previous semiannual report,’ it was stated that these temperature differences had increased. Fulil- power data were reviewed from runs 6 thiough 12, and it now appears that the increase reported is within the data scatter. The average temperature differences for run 6 were 2.11 and 1.54°F/Mw for the core support flange and the lower head, respectively, and were 2.205 and 1.55°F/Mw for run 12, Fuel Pump Tonk. — An unexplained downward shift in the temperature of the upper pump-tank surface was mentioned in the previous semiannual report.? Past data for the pump-tank temperature and for the heat removal by the oil system were reviewed to determine if a better thermal coupling could have developed between the pump tank and the shield-plug oil cooler. No evidence of in- creased heat removal by the oil system was found. The temperature distribution remained essentially the same throughout run 11, with pump operation continuing without cooling air. When the reactor was taken to power in run 12, the full-power tem- perature distribution had shifted downward another 15 to 30°F, and the pump tank continued through IMSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4119, p. 19. Ibid., p. 18. 22 the run at the lower temperatures. The tempera- tures at zero power were consistent with the run 11 zero-power data. This would seem to indicate that less fission product activity was being re- leased in the pump tank. The lower temperatures are not detrimental to the operation or to the life of the pump tank. Thermel Cvele History The accumulated thermal cycle history of the various components sensitive to thermal cycle damage is shown in Table 1.2, Approximately 63% of the design thermal cycle life of the fuel system freeze flanges has been used to date; 54% had been used at the time of the previous semiannual report.3 Temperature Mecsurement Salt Systems. - Approximately 330 thermocouples are used to measure the temperature at various locations on the fuel and coolant circulating salt systems. Only two thermocouple wells are pro- vided, one each in the coolant radiator inlet and outlet pipes. The remaining thermocouples are attached to the pipe or vessel walls., The thermo- couples on the radiator tubes are insulated to pro- tect them from the effects of the high-velocity air that flows over them during power operation; the others are not insulated and thus are subject to error because of exposure to heater shine and to thermal convection flow of the cell atmosphere within the heater insulation. In March 1965, with the fuel and coolant systems circulating salt at isothermal conditions, a complete set of readings was taken from all the thermocouples that should A similar set of data was taken in June 1967 at the start of run 12. The results of the two sets of measurements are shown in Table 1.3. Compari- son of the standard deviations for the radiator read the temperature of the circulating salt. thermocouples with those for the other thermo- couples shows the effect of insulation on reducing the scatter. Comparison of the sets of data taken over two years apart shows very little change, certainly no greater scatter, Figure 1.9 shows that the statistical distribution of the deviations SIbid., p. 20. 23 Table 1.2, MSRE Cumulative Thermal Cycle History Through August 1967 -20 0 20 4G DEVIATION FROM AVERAGE (°F) -60 —-40 Fig. 1.9. Comparison of MSRE Thermocouple Daita from March 1965 and June 1947. Thaw Component Heat /Cool - Fill/Drain Power On/Off Thaw and Transfer Fuel system 8 37 57 Coolant system 6 11 53 Fuel pump 9 32 57 428 Coelant pump 7 12 53 113 Freeze flanges 100, 101, 102 3 33 57 Freeze flanges 200, 201 7 11 53 Penetrations 200, 201 7 11 53 Freeze valve 103 6 29 35 104 14 9 25 105 16 18 43 106 18 26 38 107 10 11 18 108 9 17 14 109 9 20 18 110 2 2 3 111 5 4 112 2 1 2 204 8 15 26 20 8 L 13 24 100 ‘ --------------------- proe d&Q.OR:L.DWG:—??WB:; Tabie 1.3, Comparison of Readings of Thermocouples ! { ‘ DO;"‘ | of Salt Piping and Yessels Taken with o ) - | o | the Salt Isothermal L 0 ;20 | o Lol Thermocouple Indicated Temperature ("F) 'E § [ ( ‘ o ! Location March 1965 June 1967 e b S w Radiator 1102.6 £ 6.7 1208.5 +3.3 '; || ® DATA TAKEN 1 ) ' 777777777 tubes £ 60 MARCH 1965 4 Sy = © DATA TAKEN . Other 1102.1 £13.0 1206.7 £12.3 @ JUNE 1967 | | ‘ 3 80 e — S R All 1102.3 £10.6 1207.4 £9.8 Z 40— — — \ B e | é ’ \‘ ! of individual thermocouples from the mean also Boapl L i & changed little in the two years. 'E ‘ l The scatter in the various thermocouple read- g 50 J , B o o ir.tgs is reduced to 31.1 acce[{table level by using i | ’ biases to correct each reading to the overall ol \ 7777777777777777777777777 average measured while both fuel and coolant systems are circulating salt at isothermal condi- 5 L,.J—_o‘¥gi‘,f’o§{"§i... 0 tions. These biases are entered into the computer and are automatically applied to the thermocouple readings. The biases ate revised at the beginning of each run and are checked when isothermal con- ditions exist during the run, Generally the biased PHOTO 87924 Fig. 1.10. East Side of Heat Exchanger Showing Heater Box HX-1 and the Cocked Spacer on the Right. thermocouple readings have been reliable, but there have been a relatively few cases when there have been shifts in thermocouple readings that have resulted in calculation errors. Temperature Disturbance in Reactor Cell. — Dur ing run 11, a shift upward of a reactor cell ambient thermocouple was noted. This upward shift, which occurred on only one of ten ambient couples, took place the day after reaching maximum power. A rather extensive investigation revealed that several other thermocouples were affected at the same time, all in the area between the fuel pump and the heat exchanger. Many tests were performed to determine the cause of this temperature disturbance, but none gave any conclusive answers. This area was viewed with closed-circuit television during the run 11 shutdown in May and June. The only ab- normality noted which might have caused this in- crease was a cocked heater spacer between heaters HX-1 and HX-2 on the heat exchanger. This spacer is shown in Fig. 1.10, a photograph of the television screen. It was concluded that this cocked spacer, which was viewed only after the fuel system was drained and cooled, was an indication of an even larger opening which existed during operation. Fuel Salt Afterheat At the conclusion of run 11 power operation, an experiment was run to determine the amount of fission product afterheat in the fuel salt. Power operation of run 11 was terminated by a rod and load scram from full power, and the temperature 100 -y T 1787 e T T T T | | L"H;T 5 - —~ CALDRON CAL(,ULAnom WITH Kr AND Xe STRI ppwe L I z —4/__'__‘#_—_‘1__"1 PP e e o e f 41 9 5o [ CALDRON CALCULATIONS WITHOUT STHIPF‘INb Tl = I i T | & 11 ] mere Measurements N0r2 ] ui ) /M/:.i IZED 1O CALDRON cAL- | | | 20 FISSION PRODUA 20 O':iNL DWG 87~147 4 " TORAIN TANK - - IR ETTL .)O 100 200 500 1000 TIME AFTER SHIEITOOWAL (ke Fig. 1.11. transient that followed was recorded by the com- puter. The net heat input fo the system was evaluated several times from the combined effects of the temperature slope and thermal capacity of the system, the power to the electric heaters, and the 10 kw of nuclear power when the reactor was critical. The fuel was drained shortly after the final heat input data were taken in the fuel loop 55.5 hr after the scram. Two additional sets of heat input data were taken in the fuel drain tank at times of 168.5 and 745 hr, but there was no ex- perimental method to cortelate the drain-tank data with the fuel-loop data because of the difference in heat losses. The analysis of the experimental data gave the change in afterheat between the 55.5-hr data and the various other sets of data taken in the fue!l loop and between the two sets of data in the fuel drain tank. The computer program CALDRON was used to check the experimental results and to provide reference points at decay times of 55.5 and 765 hr. The results of the CALDRON calculations and the afterheat measurements are shown in Fig. 1.11. Two sets of CALDRON calculations are shown, one set without krypton or xenon stripping and the other with krypton and xenon stripping at a rate equivalent to the removal from the MSRE fuel salt. The MSRE experimental data were normalized to the 55.5- and 745-hr CALDRON cal- culations that included stripping. (The heat losses required to make the observations agree with the calculation at these points were assumed to exist at all other times in the same system.) We had hoped to obtain useful data within about 10 min after the scram. However, there was ap- Results of MSRE Afterhect Measurement. parently an air leak through the radiator enclosure which healed itself in about 1.5 to 2 hr. Since the calculation procedure required that the heat losses {rom the reactor system be nearly con- stant, the first 2 hr of data could not be used. Actually the calculations made 2 hr after shutdown also appear to be somewhat low, as s=en in Fig. 1.11. The thermal capacity of the fuel and coolant systems for the afterheat calculation was cali- brated during the run 12 startup. The temperature transient was recorded following a step increase in nuclear power of 148 kw. The thermal capacity of the system was found to be 16.22 Mw-sec/°F. 1.4 EQUIPMENT PERFORMANCE Heat Transtfer C. H. Gabbard The monitoring of the heat transfer performance of the salt-to-salt heat exchanger continued, both by periodic measurement of the heat transfer coef- ficient and by practically continuous observation of the ‘‘heat transfer index,”” (The heat transfer index is defined as the ratio of reactor power to the tempetature difference between the fuel leav- ing the core and the coolant leaving the radiator.)? Six sets of data were taken for evaluation of the heat transter coefficient during runs 11 and 12. Ibid., p. 21. CRNL-DWG 37-11404 ; 5" 700 — =, 007 o - 800 } o | DoB ¥ £ | J g; 2500 ¢ - l‘ i — 005 .- ‘ 5 o 400 J - 004 = 2 ‘ | b Lu'] 300 -t 003 & S | 2 E 200 — - o HEAT TRANSFER COEFFICIENT (REVISED CALCULAIION) [ : 0.02 f 9 t ® HEAT TRANSFER COFFFICIENT [ORIGINAL CALCULATION){ : < & 00— L 1 " = HEAT TRANSFER INDEX : ; J 0ot T ool Loy Ty & A M J J A S O N S S . 12€6 1 | -+ {267 - Fig. 1.12. QObserved Performance of MSRE Main Heat Exchanger. Coefficients were computed from these data by a procedure used since the beginning of power operation and by a revised procedure whose principal difference is that it uses only the most reliable thermocouples. Coefficients computed both ways are shown in Fig. 1.12 along with the heat transfer index. The coefficients and the index indicate that the performance of the heat exchanger has remained practically unchanged. (The downward shift in the heat transfer index in March 1967 is the result of revising the tempera- ture biases in the computer.) Main Blowers C. . Gabbard The rebuilt main blowers, MI3-1 and MB-3, have now accumulated 4640 and 4220 hr of operation, respectively, since they were installed in October and November 1966. The main bearing on MB-3 was replaced in early March after 1800 hr of operation, when the vibration amplitude started increasing. The balls and races of the bearing were severely scored and pitted. The replace- ment bearing also gave an indication of trouble and was scheduled for replacement during the run 11 shutdown. However, the problem turned out to be the result of a loose vibration pickup. A complete inspection of the blowers and drive motors was made after the run 11 shutdown. Both blowers were again in excellent condition after 3585 and 3162 hr of operation, with no indication of cracking in the blades or hubs. The slip rings and brushes on the drive motors had become scored, and the motors were removed for repair. The re- pairs included refinishing the slip rings, replacing the brushes and bearings, and balancing the rotors. Vibration pickups were added at each motor bearing, and filters were installed to pro- tect the slip rings from dirt and grit. When the blowers were test run, there were excessive vi- brations on the drive motor of main blower 3. The motor vibration had been satisfactorily low when the motor was loosened on its mount, indicating that the motor was not badly unbalanced. The rotation speed of the motors was found to be very near the natural frequency of the motor mount. The vibration amplitude was reduced to an accept- able level (below 1 mil) by stiffening the mount, Insulation Dust in the Reactor Cell. ~- During observation in the reactor cell between runs 11 and 12, a nonuniform coating of white material was seen on most of the horizontal surfaces of the re- actor cell (see Fig. 1.13). Samples of the white coating were obtained with long-handled tools and identified as being mostly AlZO3 (insulation). Attempts to further identify it as one of the two specific types of insulation known to be in the cell were unsuccessful. The possible sources are the insulation covering the fuel pump and over- flow tank, the reactor vessel, the fuel drain line, or the fuel line under the heat exchanger, The drain-tank cell was also viewed, but no covering of insulation dust was noted. Radiator Enclosure M. Richardson The brake shoes in the brakes of the radiator door lifting mechanism were found to be worn and PHOTO 87750 Fig. 1.13. Motor of Reactor Cell Cooler No. 1 Showing Dust Accumulation. were replaced after run 11. This reduced the coastdown after the doors were partially lowered to 3 in. It was also necessary to replace part of the outlet door soft seal gasket material which had burned and blown loose. Operation of the doors has been without incident, and the radiator seals have been adequate for operation. Off-gas Systems . B. Engel Operational difficulties with the off-gas sys- tems were greatly reduced during this period of operation. One 7-hr power reduction was required to replace a filter in the coolant off-gas line. Otherwise, only minor inconvenience, which had no effect on power operation, was experienced. Particle Trap. — The new off-gas filter® (particle trap) installed in the fuel off-gas line before the start of run 11 continued to function satisfactorily with no evidence of increasing pressure drop. The pressure drop across this unit, with one sec- tion valved out, remained below 0.1 psi through- out the operation. The temperatures near the various filter media depend to some extent on operating conditions other than power. Increased pump tank pressure or reduced purge-gas flow increases the transport time for fission products from the pump tank to the particle trap. This per- mits more radioactive decay en route and results in lower temperatures at the particle trap. Never- theless, under similar conditions the steady-state temperature near the coarse filtering material *Ibid., p. 42. (Yorkmesh) was ~ 275°F when the particle trap was first used and ~ 380°F near the end of run 12, The temperatures decrease rapidly when the re- actor power is reduced, however, and the zero- power steady-state temperatures are essentially unchanged. These effects indicate the accumula- tion of some material, presumably organic, on the filtering media that enhances the retention of short-lived fission products. Main Charcoal Beds. — The performance of the charcoal beds in holding up noble-gas fission products has continued to be satisfactory. The gradual development of restrictions at the inlet ends of the beds has also continued, but this has not limited the reactor operation in any way, since effective measures can be taken to reduce the restriction when necessary. _ Run 11 was started in January 1967 with the charcoal bed sections 1A and 1B in service with an initial pressure drop of 2.5 psi at normal off- gas flow. The pressure drop increased very slowly, reaching 7 psi on March 29, two months after the start of the run. At that time the standby beds, 2A and 2B, were put in service, and the restricted sections were valved out. The pressure drop across these sections built up from 2.6 to 9 psi in only ten days. We then cleared the restrictions from all four sections by forcing clean helium through sections 2A and 2B in the normal flow direction and heating the inlet ends of sections 1A and 1B with previously installed® electric heaters. These operations did not require a re- actor shutdown but only a temporary lowering of the water level in the charcoal bed pit to allow the heaters to function. After the restrictions in both sets of beds had been cleared, sections 1A and 1B were put back in service. In the ensuing three weeks the pres- sure drop increased from 2.4 to 3.5 psi. At that time, we decided to increase the helium purge flow by 1 liter/min to see if the xenon poisoning would be affected by lower concentrations in the fuel pump gas space. To accommodate the higher gas flow without an increase in fuel pump pres- sure, sections 1A and 1B were valved out and 2A and 2B were put in service. The next day section 1A had to be reopened to keep the fuel pump pres- sure at 5 psig. The normal purge flow was re- stored after three days, and, just before the power shutdown at the end of run 11, the partial restric- ®Ibid., pp. 30-31. 28 tions in all four beds were again removed by heat- ing sections 1A and 1B and forward blowing sec- tions 2A and 2B. Owing to the success of the heaters in clearing the restrictions from sections 1A and 1B, we in- stalled similar heaters at the inlets of sections 2A and 2B during the shutdown between runs 11 and 12. The differential pressure transmitter that senses charcoal bed pressure drop directly was also replaced. This instrument had failed earlier, possibly because of the pressure differ- ences imposed during blowouts of the charcoal beds. However, all these pressure differences were within the specified overrange capability of the instrument. Power operation in run 12 was started with sec- tions 1A and 1B in service. The gradual increase in pressure drop made it necessary to change to sections 2A and 2B after about three weeks. The pressure drop across the second sections reached an unsatisfactory level after only six days. Then the restrictions were cleared from all four sections by heating the inlet ends. The remainder of run 12 was completed with sections 1A and 1B in service. The development of flow restrictions at the char- coal beds appears to be related to the accumula- tion of volatile organic matter on the steel wool packing at the bed inlets. Physical variations in this packing probably account for the different times required to plug various individual sections. The experience in runs 11 and 12 indicates that the restrictions can be effectively removed by electrically heating the inlet ends of the beds. Presumably, this heating drives the volatile matter off the steel wool packing in the inlets and moves it farther downstream where the flow areas are larger. There is no evidence from the charcoal temperatures that this material has reduced the fission product retention capability of the charcoal. Since the heating operations do not affect reactor performance, there are no plans at present to make further modifications at the charcoal beds. Coolant Off-gas System. — Very slow plugging of the coolant off-gas system at the filter that precedes the coolant-loop pressure control valve has been encountered throughout the reactor opera- tion. The originally installed filter was replaced in February 1965, during the preoperational check- out of the system. Subsequent replacements were made in March and September 1966 and on March 1, 1967, The replacement on March 1, 1967, re- quired a reactor power reduction for 7 hr to permit personnel access to the area where the filter is located. By the end of run 11 (May 1967) the filter was plugged again, and periodic venting of the coolant system through an auxiliary line (1.-536) was required to keep the loop overpressure below 10 psig. During the shutdown between runs 11 and 12, the filter was replaced again, and a minor piping modification was made to permit the coolant off- gas activity monitoring to monitor gas vented through line 536. Before this change, any activity release would have been detected and stopped by another monitor on the combined fuel and coolant off-gas, but identification of the source of the activity would have been more difficult. No activity has ever been detected in the coolant off-gas, Cooling Water Systems A. L. Krakoviak The cooling water systems performed satis- factorily during this report period. The systems functioned relatively trouble free except for a few leaks. In July a 15-gpd leak from the treated water system was detected and was traced to a faulty pressure-relief valve in the line leading to one of the reactor cell coolers. Replacement of the faulty relief valve restored the system to normal operation. Space Coolers, — As reported previously,” leaks have occurted in the reactor cell space coolers at the brazed joints on the brass tubing headers, During the scheduled shutdown at the end of run 11, both space coolers were leak-tested. One cooler (RCC-1) leaked less than 125 cm?®/day and was not replaced; however, the other (RCC-2), which leaked at the rate of 9 liters/day, was re- placed with a cooler whose headers and nipples were fabricated of copper. Copper weldments were used on the new cooler instead of the brazed joints. Radiation levels around the removed unit were sufficiently low that it could be disassembled directly. The radiator was the most radioactive component, with readings up to 1000 millirems /hr at contact. (The radiation was very soft and caused no contamination problem.) This unit was discarded. However, the fan motor was retained "Ibid., p. 32. 29 for possible future use, and the new fan, motor, and radiatot were mounted on the original frame for installation in the cell. Reactor Cell Annulus. — Sometime prior to or during run 11, the fill line to the biological shield plugged, and water additions to the reactor cell annulus were made through the level measuring line. Since the plug in the fill line could not be cieared, the overflow pipe from the cell annulus was modified to also serve as a fill line. Steam Dome Feedwater Tanks. — Water is dumped automatically from a feedwater tank (FWT) to a steam dome if cooling of a fuel drain tank (FD) is required after a fuel drain. - During run 12, small amounts of water from FWT-1 had randomly ap- peared in the steam dome of FI-1, causing a tem- perature decrease in the fuel drain tank. To en- sure that this drain tank remained available for a possible emergency drain, the water was removed from the feedwater tank, which is now in normal service after having a faulty temperature switch replaced. Component Cooling System P. H. Harley Although some difficulties were encountered, the component cooling system operated satis- factorily during this report period. The two main blowers (CCP-1 and CCP-2) operated 1536 and 2375 hr respectively; CCP-2 has operated for a total of 3340 hr without a failure. The discharge check valve on CCP-2 was re- placed and the belt drive was tightened as part of the preventive maintenance program. There was no indication of any significant aging of the sili- cone rubber in the removed check valve. Trouble was encountered in the CCP-1 oil cir- culating system. First, a loose tubing connection caused the loss of ~ 2 gal of il during run 12; this irregularity was easily repaired. Then, fol- lowing the run 12 shutdown, intermittent low-oi}- pressure alarms again occurred. An investigation indicated no significant loss of cil, but a slow oil- pressure response was observed when the blower was started. The suspected oil pump and pressure- relief valve on CCP-1 were replaced with spare units to cormrect the trouble. The removed pressure- relief valve was found to be relieving before the normal oil pregsure developed. In spite of these difficulties in the system, there was sufficient lubrication, and no noticeable damage was ob- served, Although the temporary strainer in the CCP dis- charge line worked satisfactorily, a more efficient strainer, which had been on order for a year, was received and installed in the line. Over an eight- month period the temporary strainer accuniulated ~ 30 to 50 g of black, dry powdery material that appeared to be dust from abrasion of the drive belts. The new strainer, however, has a 100-mesh screen and a 0.2-psi pressure drop as compared with a 1/“,)-in. pore size and a 0.5-psi pressure drop in the old strainer. The improved performance of the blower belt drives, which have not needed replacing during the past nine months, is attributed to less frequent starting and stopping of the blowers as well as to the reinforced motor support. During early opera- tion, the blowers were alternated twice a month; now one blower is operated continuously during a run., The stainless steel strainer which was removed had been in contact with condensate containing dilute HNO, while in service (see ‘‘Containment,”’ p. 33). After being decontaminated, the strainer was examined and was found to be in very good condition. The surface was slightly etched, but no more than would be caused by the decontamina- tion process. Blower CCP-3, which coaols out-of-containment freeze valves, failed on April 19 after more than 5000 hr of operation, A bearing galled and damaged the drive shaft. Operation continued without inter- ruption by using air from the service air compres- sor. Blower CCP-3 has heen repaired and can now be used when required. Salt Pump Oil Systems A. 1. Krakoviak The lubricating oil systems for both salt pumps have been in continuous service except during the planned oil change during the shutdown after run 11. At this time the oil was sampled, drained, and replaced with new oil. The oil, which had been in service since August 1966, showed no signifi- cant change in its physical or chemical properties. During the steady full-power operation in run 11, very good balances were obtained on the oil sys- tem inventory changes, indicating little or no loss 30 by leakage into the pump bowl. The measured amounts removed for analysis and accumulated in the catch tanks actually slightly exceeded the ob- served decreases in supply reservoir contents. In March and April the difference was 65 cm?® in the fuel pump system and 210 cm® in the coolant pump 8ystein, The oil leakage through the lower seal of the fuel-pump shaft had previously accumulated at the rate of 5 cm®/day; it has now decreased to ~ 1 cm?®/day. The leakage past the lower seal of the coolant salt pump averaged 17 cm?®/day during run 11 and 30 cm?®/day during run 12; the present ac- cumulation rate is 15, Automatic siphons were originally installed on the oil collection tanks from both pumps to meas- ure and dispose of seal leakage without manual draining to keep the level in the sensitive (reduced cross-sectional area) range of the level-measuring leg of the collection tanks. These siphons have failed to function properly at the low oil leakage rates that actually occurred. The oil simply flows over the high point of the siphon tube, much like a liquid flowing over a weir, without bridging the tube to form a siphon. The overflow points for the coolant pump and fuel pump oil collection tanks were reached in March and April respectively. For the remainder of run 11 the only indicators of leakage rate were the supply reservoirs, which are much less sensitive than the leakage collec- tion tanks. After run 11 the collection tanks were drained, and the average leakages for the latter part of the run were determined by measuring the total accumulated oil leakage. The collection tanks were drained again after run 12, and periodic drainings are planned to keep the oil level below the siphon (overflow) level. Because there is a high radiation field at the collection tanks, during power operation the draining operations must be performed only when the nuclear power is low. If high leakage rates (500 to 1000 cm®/day) develop, the auntomatic siphons are expected to function as designed, Although the oil from the coolant pump seal when sampled had a dark appearance, spectro- graphic and infrared spectrophotometric analyses showed no significant difference between the seal leakage oil and a sample of unused oil. The dark appearance and the somewhat increased leakage rate past the seal could be an indication of ab- normal wear at the Graphitar—stainless steel rotary seal of the salt pump. The somewhat lower accumulation rate at present indicates that the seal may have reseated itself. Electrical System T, L. Hudson Power to the MSRE electrical system is supplied from the ORNIL. substation by either of two 13.8~ kv TVA power lines, a preferred line or an alter- nate. During this report period, while operating on the preferred feeder, there were two unscheduled electrical interruptions when the reactor was at power, During a thunderstorm on June 25, the reactor operation was interrupted by the loss of hoth feeders. Two amplifiers were damaged on the reactor period safely system. Approximately 32 hr later, the reactor was returned to critical opera- tion after the period safety amplifiers had been tepaired. On July 12, the other interruption was caused by the loss of the preferred feeder during another thunderstorm, Emergency power from the diesel generators was in gervice within 2 min, and low-power nuclear operation was resumed in about 13 min. After repairs had been completed on the preferred feeder, full-power operation was resumed in approximately 6 hr. Difficulty was experienced in run 12 when re- starting main blower 1. The breaker tripped off the blower during the starting sequence for several attempted starts, On a later occasion, the breaker tripped several times before main blower 1 was started. This erratic behavior was explained when a loose gasket was found in one of the time- delay orifices when the breaker was checked in August. Tests have been made that indicate the total time of the starting sequence is too long when compared with maximum time delay of the breaker overload element. Therefore, the total time of the start sequence will be reduced from 25 to 12 sec. Heaters T. L. Hudson The last of six heating elements in heater HX-1 failed on October 28, 1966. Satisfactory heat ex- changer temperatures were maintained without this heater, even with the fuel loop empty. However, continuous circulation of the coolant salt was maintained until after run 11, so the full effect of the heater failure could not be determined. Tests were performed with both the fuel and coolant loops empty after run 11 to determine the need for this heater in preheating the system from a cold condition. When helium circulation was stopped in the fuel system, the temperature distribution was satisfactory for a fill without heater HX-1 operating. However, with helinum circulation in the fuel system, ‘‘cold’” helium was introduced - possibly from the fuel pump -~ and one temperature decreased to below 800°F, Since satisfactory temperatures could be achieved without it, the failed heater was not replaced. Radiator heater CR6-48 failed on March 30, 1967, and was replaced with a spare heater (CR4-3C). During the shutdown after run 11 the electrical lead to heater CR6-4B was repaired, and the heater was placed back in service. Several broken ceramic bushings were alsc replaced at this time. Contro! Rods and Drives M. Richardson Performance of the control rods and drives has been within the operating limits this period. No mechanical failures of the rods or drives have occurred. The fine-position synchro of rod 2, which had failed during the previous period, was replaced prior to the beginning of run 12, The drive unit for this rod was inspected at the same time and found to be in excellent condition. The grease in the drive unit appeared unchanged after a total radiation dose of about 10? rads and there- fore was not replaced, After run 12, routine checks of absolute rod position using the single-point indicators in the rod thimbles revealed an apparent upward shift of 0.5 in. for rod 1. Since there wag an equjvalent shift in the upper and lower limit switches, it is believed that this shift was caused by slippage of the drive chain on one of the sprockets after a rod scram. The exact time of the shift is not known because, between the absolute position measurements, there were control rod scrams be- fore and after run 12. Salt Samplers R. B. Gallaher Fuel Sampler-Enricher. — The fuel sampler- enricher was used intensively during this report period with only a few minor difficulties until the failure that brought run 12 to an end. Between March 1 and August 5 there were 111 operations, as follows: Salt samples 71 Freszze-valve samples of cover gas 6 Exposure of graphite to salt and cover gas Beryllium additions 6 Uranium additions 27 Of the salt samples, 14 were 50-g samples and 2 were taken in special three-compartment capsules. The special samples are described in the section ‘‘Reactor Chemistry.”” These operations brought the total, since the sampler-enricher was installed in March 1965, to 114 uranium enrichments and 279 samples and special exposures. The uranium additions, the first made with the reactor critical, provided information on mixing between salt in the sample enclosure and the main circulating stream. Response of the re- activity (Fig. 1.7) revealed a time constant close to that for mixing betweesnt the pump bowl and the main stream, indicating rapid melting of the en- riching salt and good circulation through the sample enclosure. Near the end of run 11 the manipulator arm and boot assembly was replaced after a small leak appeared in one ply of the boot. The arm was quite contaminated (300 r/hr at 3 in.), but it was successfully decontaminated and saved for pos- sible future use (see p. 40). At the start of run 12, one ply of the boot was ruptured when ex- cessive differential pressure was inadvertently applied, and again the assembly was replaced, This time a slightly different boot was used. The new boot was thicker and more durable (at the expense of ease of manipulation), and the outside was coated with a white plastic spray which ef- fectively improved viewing in the sampler by de- creasing light absorption. One of the beryllium addition capsules was dropped while it was being removed from the 1-C area with the manipulator. The capsule fell onto the gate of the operational valve, where it was retrieved by a magnet lowered on a cable. Enriching capsules occasionally jammed in the transport container until the difficulty was elimi- nated by increasing the length of the cavity in the disposable portion to give more room for the long capsule and attached key. 32 The neoprene seals on the 1-C access port be- gan to show increased leakage, possibly due to radiafion damage. Fission products, mostly the species found in cover-gas samples, produced radiation levels in area 3A of several hundred rads per hour. Before run 12, the radiation in this area was reduced a factor of 10 by using the manip- ulator to wipe down surfaces with damp sponges. In run 12, increased leakage from the buffer zone between the seals on the 1-C access port was met by increasing the size of the helium supply flow restrictor so that a satisfactory buffer pressure could be maintained. Operation of the pneumatic clamps on the 1-C access door occasionally re- sulted in gaseous fission products being vented through the operator vent line. A small charcoal filter was added to prevent this activity from reaching the stack. As described on p. 15, a situation developed on August 5 that led to a shutdown of the reactor and replacement of the 1-C assembly. It now appears that the trouble started when the latch hung at the maintenance valve. Indications were that hoth isolation valves weie fully open and that the capsule key was hanging properly in the latch at the outset; so the exact cause of the hangup is not known. But there is convincing evidence that the drive cable was severed by the operational- valve gate while the latch was at the maintenance valve. The length of the latch and remaining drive cable (Fig. 1.3) equals the distance from the maintenance valve up to the gate of the operational valve, and the cable end appeared to have been cut. It was believed that the operational valve was practically closed before substantial resist- ance was felt, but it would have been easy to mistake the torque and motion involved in shear- ing the cable and seating the valve for simply seating the valve, (The handwheel torque re- quired to shear the cable was subsequently cal- culated to be less than 100 in.-1b.) Once before, in December 1965, the latch ap- parently hung up at one of the isolatioa valves, causing the drive cable to coil up in the 1-C area and in the drive unit. Following that occasion, the travel of the valve gates on opening was in- creased slightly, and when the 1-C assembly was removed after the recent trouble, the valve gates were observed to completely clear the sample tube, The valves were also obseived to function reli- ably, so no changes were made, The retrieval of the sample latch would have been facilitated had it been magnetic. Therefore, the replacement latch was made of magnetic type 430 stainless steel instead of type 304 stainless steel. Another change in the replacement unit was the addition of a sleeve to bridge the 2-in. gap between the cable drive reel and the floor of the drive compartment. This will prevent the cable from escaping into this compartment if it meets resistance while being unreeled. As explained on p. 15, the detached capsule is positively confined by the batfle so that it cannot escape. Swirl velocities observed in the bowl of the prototype salt pump were less than 0.1 fps, indicating that continual movement of the loose capsule is very unlikely. Nor should the capsule corrode. The body of the capsule is copper, and the cap is mild steel with a thin nickel plating that probably leaves some ferrous metal exposed. On exposure to the molten salt, the copper-iron- nickel assemblage in contact with the chromjum- containing Hastelloy of the baffle and Tower head will temporarily constitute an electrochemical cell, in which Cr? from the Hastelloy surfaces will be oxidized to Cr? " and reduced once again to Cr® metal on the capsule surface. This re- action slows down as the chromium activity in the capsule surface approaches that of the Hastelloy surfaces, Eventually the transfer diminishes to the rate at which the chromium can diffuse into the capsule metal. The amount of chromium that can be transferred in this way will have negligible effect on nearby Hastelloy surfaces., The con- clusion is, thetefore, that no ill effects will re- sult from the presence of the capsule in the pump bowl. Coolant Sampler. — During this report period, seven 10-g samples were isolated from the coolant salt pump, bringing the total with this sampler to 60. There were no operating difficulties and no maintenance was required, Fuel Processing Sampler. — The shielding around the fuel processing sampler was finished, completing the installation of this sampler. The nylon guide in the removal area and the manipula- tor arm were subsequently removed for use in the fuel sampler. Other spares are now on hand but will not be installed until near time for use of the processing facility. 33 Containment P. H. Harley R. C. Steffy Secondary Containment. — During run 11, the containment cell inleakage was measured to be ~ 10 scf/day with the cell at —2 psig. This de- termination was made by monitoring the cell pres- sure with the Hook gage (a sophisticated water manometer) and by measuring all known purges into and out of the containment cell. The cell pressure and system purges remained virtually constant, with small pressure oscillations occur- ring as the outside temperature fluctuated, These oscillations were particularly noticeable during the colder months. Between May 16 and June 13, an extensive containment check was made. Thitteen block valves out of a total of 160 were found to be leaking and were repaired. Five were instrument air block valves, three were in the cover-gas (helium) system, and five were in the treated- water system, The most common cause of leak- age was aging elastomer O-rings; a few, however, had scale, metal filings, or dirt on the seating surfaces. None of the valves leaked excessively, and seven of them are backed by a closed system. Two others ate in use <0.1% of the time and are normally backed by closed hand valves. At the beginning of run 12, the secondary con- tainment vessel leakage was checked with the containment cell at 20 psig. At this pressure the cell leaked only ~ 28 scf/day. The cell was then evacuated to — 2 psig, and reactor filling opera- tions were started. As during earlier runs, there was a water leak in the cell during run 12 (this leak is discussed later in the section). As a result of the water leak the first determinations placed the cell leak rate at ~ 70 scf/day. However, as soon as the cell air became saturated, the indicated leak rate dropped to <20 scf/day. This took only about seven days. For the major part of run 12, the indicated cell leak rate remained essentially constant at ~ 14 scf/day. For both runs 11 and 12, the cell leak rate values were well below the 85 scf/day permissible (extrapolated from accident conditions at a 39- psig cell pressure). Although the Hook gage has been the primary means of determining the cell leak rate, an on- line oxygen analyzer (Beckman model F3) was installed at the MSRE with the expectation that it could be used as a megans of calculating the cell leak rate as well as keeping track of the cell oxygen content. Cell oxygen content is held > 3% to prevent nitriding of the Hastelloy N and <5% to eliminate the possibility of combustion in the cell in case of an oil leak. To obtain consistent data from the analyzer, we have found it necessary to calibrate the in- strument every 8 hr. In spite of this calibration frequency the indicated oxygen content often changes 0.2 or 0.3% between readings taken every 4 hr, Since a 0.1% change in oxygen coiresponds to ~ 60 scf of air, it is evident that only data taken on a long-term basis are meaningful. During both runs 11 and 12, data from the oxygen analyzer indicated a negative cell leak rate (the cell was leaking out instead of in). A possible explanation for this negative cell leak is that something is combining with the oxygen in the cell. If it is assumed that the Hook gage is cor- rect, then ~ 3.7 scf/day of oxygen must be re- moved from the cell atmosphere to explain the oxygen results. Oxidation of a small amount of oil on hot surfaces could easily consuine this amount of oxygen. Additional investigation will be required to resolve the discrepancy between the two methods of cell leak-rate measurement. Activity Releases. — The total activity release to the atmosphere during this repoit period con- sisted of 3.99 mc of iodine and <0.23 mc of paiticulate matter. The largest single release was 1 mc of iodine, released while removing graphite samples from the reactor core on May 15. Other measurable releases included 0.7 mc of iodine activity while preparing to remove the broken sample-cable latch and 0.05 mc of iodine released when the Fuel Storage Tank was vented to the stack prior to modification of the fuel processing piping. Ventilation. — No difficulties were encountered 34 with the ventilation fans during the past six months. The ventilation discharge filters showed an in- crease in AP from 1.13 to 1.85 in. H, O across the roughing filter section. There was no increase across the absolute filter section. The annual filtering efficiency test of the absolute filters in the ventilation discharge was performed on June 5, 1967. Results of the standard DOP (dioctyl phthalate) test for the three banks of filters were 99,994, 99,998 and 99.979%; the minimum ac- ceptable efficiency is 99.95%. Contamination experience in the vent house during maintenance operations® and the prepara- tions for the off-gas sampler led to several re- visions in the ventilation piping in that area. A new 0-in. line was installed between the main ventilation line and the instrument and valve box of the reactor off-gas system in the immediate vicinity of the particle traps. The existing vent line from the valve box was increased from 2 in. to 4 in. in diameter, and a 4-in. branch was in- stalled to vent the off-gas sampler box. These lines will improve the ventilation of the affected areas and reduce the release of contamination during maintenance when the areas are open to the atmosphere. Moisture in Reactor Cell Atmosphere. — In both runs 11 and 12, it appeared that the atmosphere in the reactor and drain-tank cells increased in humidity for the first few days after the contain- ment was sealed, until finally moisture began to condentse in cooler points in the component cool- ing system. Condensate was drained daily from the inlets of the component cooling blowers and the shell of the air cooler at the discharge of the blowers. During runs 11 and 12 the total conden- sate collected was 100 and 50 gal, respectively, averaging about 0.9 gpd. At least part of the inleakage in run 11 was from a space cooler (RCC-2) in the reactor cell which was definitely leaking before it was re- placed at the end of that run. The source of the leakage during run 12 was not located, not even as to whether it was in the reactor or drain-tank cell, since no water appeared in the sump in either cell in run 11 or 12, To determine whether the leak might be in the nuclear instrument penetration, the water for shielding and cooling this penetration was spiked with 6 kg of D, O during run 12. Analysis of the condensate showed no detectable increase in deuterium content above the normal water concen- tration. The sensitivity of the deuterium analyses was such that a leak of 0.01 gpd into the con- tainment from the instrument penetration would have been detectable. 81pid., p. 38. The moisture condensed from the cell atmos- phere was mildly acidic and contained consider- able tritium, Results of analyses during each run were as follows: Date Sample 3 {dis min~ ] ml“l) pH NO,™ (ppm) 2.21-67 LW-11-1 2.8 % 10° 2.7 221 7-27-67 LW-12-1 2.7 x 10° 2.5 3721 The condensate collected in runs 11 and 12 contained a total of about 700 curies of tritium, which was stored in the MSRE waste tank before being transferred to the Melton Valley waste system. The source of the tritium is presumably neutron reactions with °Li in the thermal insula- tion around the reactor vessel. A spectrographic analysis of Careytemp insulation like that used in the thermal shield shows the lithium content to be (.1%. Calculations show that with approxi- mately 1200 Ib of insulation containing 0.1% 4> 35 natural lithium, exposed to a thermal-peutron flux of about 10'? neuntrons cm™? sec™ !, the amount of tritium observed is entirely reasonable. The nitrate ion in the condensate, which makes it acidic, is presumably caused by ionization of nitrogen in the cell atmosphere. (The nitrogen content is kept at 95 to 97% by addition of nitrogen through the cell sump bubblers.) The acid con- centration in the condensate is relatively low, and little corrosion is evident. A general spectro- graphic analysis of a condensate sample taken in August gave the following results: Element Element ppm ppm Al 0.03 Mg 0.062 B 0.20 Mn 0.04 Ca 0.80 Ni 1.60 Cr 0.02 Pb 1.00 Cu 0.44 Zn 0.20 Fe 2.20 2. Component Development Dunlap Scott 2.1 DFF-GAS SAMPLER R. B. Gallaher A. N. Smith The system developed for sampling and analyz- ing the fuel off-gas was described in the last progress report.' During this report period, in- stallation of all peripheral equipment was com- pleted, but the installation of the actual sampling mechanism was delayed. (Peripherals include shielding, external tubing and connection points, power and instrument wiring, panels, and instru- ments.) During final inspection and checkout of the sampling mechanism, a failure was detected in a Monel-to-stainless-steel weld. Fuither detailed radiographic inspection revealed some other welds, mostly Monel to stainless steel, that were judged to be marginal or unacceptable for primary reactor containment. Therefore, the mechanism was modified to eliminate many of the dissimilar- metal welds (by substituting stainless steel valves for Mone!l in the valve manifold) and to reduce mechanical stresses at the welds. The valve substitution eliminated brazed joints at the valve bodies, and other minor modifications were made to improve the containment integrity of the assembly and the sample transport bottle. The sampler was reassembled, with all welds fully certified, but installation was deferred because of other work during the shutdown following run 12, 2.2 REMOTE MAINTENANCE Robert Blumberg During this report period the trend for more re- mote maintenance to be handled routinely was ac- 1 . MSE Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL.-4119, pp. 41—42. 36 celerated, but development efforts remained con- siderable and varied. The shutdown in May and June was planned, and considerable effort was spent in advance on de- velopment of procedures, preparation of tools and equipment, and training of maintenance persounnel. As a result, most of the remote maintenance work during this shutdown was handled by the normal maintenance forces, Development personnel also paiticipated and afterward reevaluated procedures and tools. The next shutdown was unscheduled. Although some equipment and general techniques already developed for other jobs proved to be adaptable, intensive work was required to develop tools and procedures for retrieving the sample latch. Preparations for Shutdown After Run 11 At the beginning of MSKE operation, when tools and procedures for all remote maintenance jobs were still being tested and revised, development personnel provided direct gunidance and much of the actual work. But the work planned for the shutdown after run 11 consisted mainly of jobs for which tools and procedures had alteady been proved. Therefore, the contribution of the develop- ment group shifted more toward training main- tenance personnel, helping plan the operations, and putting the maintenance equipment in readiness. The training program consisted of a series of lectures, demonstrations, and practice sessions conducted by members of the maintenance develop- ment group and attended by some 30 people, in- cluding 5 craft skills and their supervision. The lectures covered health physics, safety considera- tions, general requirements of all remote main- tenance, remote maintenance sirategy used at the MSRE, details of some of the equipment, and de- scriptions of specific problems. The demonstra- tions involved the viewing equipment, the remote cranes, the portable maintenance shield, and some of the long-handled tools. core sample removal and replacement served both to train personnel and to shake down the equipment. Practice sessions on The planning part of the preparation involved the writing or updating of step-by-step procedures for each job. These procedures were used to draw up lists of required tools, equipment, and material, and to estimate time and manpower requirements. Possible problem areas were recognized for prior testing and mockup practice. Before the shutdown began, the procedures were used to provide informa- tion to the foremen and the working crews. A large effort was required to put &ll the physical equipment into readiness. This meant fabrication of new equipment; cleaning, maintepance, and repair of existing equipment; and procurement of some special items. Numerous long-handled tools, containers for hot pulls, materials for contamina- tion control, handling equipment, and devices for shielding, lighting, and viewing were prepared. £valuation of Remote Maintenance After Run 11 The maintenance work during this shutdown was done on a two-crew, two-shift-per-day basis. Al- though thete were drawbacks, such as spreading more thinly the knowledgeable personnel and lost motion at shift change, the work proceeded safely and smoothly. Radiation levels were about the same as at the last shutdown and did not require any changes in procedure. Some bothersome de- lays were caused by breakdowns of the cranes and maintenance shield, but the schedule was not seriously affected. Comments on individual jobs follow. Removing and replacing the sample array in the core was the most difficult task of the shutdown, involving as it did an extremely intense source of radiation and contamination, The standpipe above the reactor access flange contained the assembly and maintained an inert atmosphere on it. Before the operation a new charcoal filter was installed io the vent line from the standpipe to prevent iodine releases, The procedure used previcusly was changed to minimize the time of exposure for the uncontained sample within the standpipe and to provide complete containment 37 and shielding during the transfer of the sample from the standpipe to the shielded carrier. Two unanticipated situations arose. The bushing which supports the sample assembly and which is sup- posed to stay locked in position in the vessel outlet strainer actually came out on the sample basket. After visually ascertaining that the strainer was not damaged, a new bushing was inserted. When difficulty was encountered later in obtaining a satisfactory seal while replacing the reactor access flange, four bolts were re- placed, and all the threads on the nuts were cleaned by tapping. This permitted an increase in gasket loading which eliminated the leak. Removal, repair, inspection, and replacement of a control rod drive went smoonthly. The leaking east space cooler was disconnected through the portable maintenance shield and then was removed by using the crane, television, and remotely operated maintenance-shield slide from the maintenance control room. A new unit, reas- sembled on the old frame, was reinstalled the same way without excessive difficulty. The metallurgical specimens which hang in the reactor vessel fumace were replaced without dif- ficulty. The carrier for the core samples was used to transport the specimens to the hot cell. The fresh neutron source was transferred from a carrier and placed in the source tube (on top of the original source) without incident. Visual inspection in the cell was by use of a periscope, binoculars, and remote television. Thermocouple junction boxes were tested, and four thermocouples were plugged into a spare dis- connect box by using long-handled tools through the maintenance shield, As a result of the experience during this shut- down, procedures were reviewed and revised where desirable, and some tools and equipment were revised or overhauled. The television camera mounts were changed for greater flexibility, and revisions were made to the track and to the roller and guide bearings on the maintenance shield. Repair of Sampler-Enricher and Recovery of Latch After the failure of the sampler-enricher drive unit, a lead carrier was built, and detailed pro- cedures for the removal of the defective unit were prepared and reviewed. The actual removal of this component was without incident. Meanwhile, retrieval tools for the latch were designed and tested in a mockup of the sampler tube. All the tool shafts had flexible sections 16 ft long to negotiate the two bends in the 11/2- in. pipe. One noose, one corkscrew, and two gripper tools were tested originally. All the tools proved capable of grasping a dummy latch, but the noose tool (Fig. 2. 1) gave the most secure grip and a positive indication when the latch was snared. After the latch was found to be stuck in place (see p, 15), the dislodging tool shown in Fig. 2.1 was developed and proved capable of greater lateral force and considerable impact. ff/g—in CABLE {/32-in, CABLE ‘/2 ~in. FLEX CABLE SHEATH C/ CABLE BEING NOOSE READY PUSHED OUT FOR FISHING NOCSE TOOL '/1g —in. CABLE / 16 72 in. SCHED 40 PIPE LATCH RETRIEVAL TOOL 38 3 in. DIAM x 3% in. | Later, when the latch became hung in the tube as it was being lifted with the noose and again while it was being lifted with the corkscrew, the con- clusion was that the latch stem was being forced against the tube wall, causing it to hang. A tool was then devised that slipped down over the latch to hold it securely and keep it from hanging on the way up the tube. This is the third tool in tig. 2.1 and the one that eventually brought up the latch. The tool that was developed to check the sampler tube for obstructions above the latch stop and to retrieve the capsule is shown last in fig. 2.1 ORNL-DWG &67-11788 {/4-in. CABLE .::/ —{Y/a-in. DIAM DISLODGING TCOOL STEEL TAPE— =116 —in. CABLE s — 1 in. 1D COPPER 7 L e MAGNET —~—{. 375 in. DIAM BRASS S LSS L /7?‘7;\-.\\\ < GO - GAGE AND CAPSULE RETRIEVAL TOOL Fig. 2.1. Tools Developed for Retrieval of Latch and Capsule from Sampler-Enricher. 39 PHOTO 88999 Fig. 2.2. General View of Latch Recovery Operation with Tool Retrieval Shield in Place. Some delays were encountered in the retrieval operation because of the necessity of building shielded carriers into which to pull the contami- nated tools. As a result a very convenient ar- rangement was developed. A hollow lead cylinder, 24 ft long with 1- to 11/2-in. walls, strapped to a steel I-beam formed the body of the carrier. A 24-ft length of 11/2-ir1. pipe, with a gate valve at the lower end, fitted into the shield and could easily be dropped out at the burial ground with the contaminated tool safely contained inside. Figure 2.2 shows the containment enclosure around the restricted work area, with the shielded carrier suspended from the crane bridge. The workers on the bridge are in position to pull a tool out of the sampler tube with a cable dropped through a seal on the upper end of the pipe liner. This system was used for five hot pulls, with no spread of contamination or excessive radiation. When the replacement sampler drive unit was installed, some difficulties were encountered in fitting up pipe and tubing connections, but con- tainment joints were made up leak-tight. The procedures and tools developed for this shutdown proved effective. The tool retrieval shield, in particular, was a useful development. Measures for control of contamination and radia- tion prevented spread of activity outside the restricted work zone, and the highest quarterly dose for any worker was only 300 millirems. In view of the intense sources involved, this record attests to the effectiveness of the controls, 2.3 DECONTAMINATION STUDIES T. H. Mauney The maintenance scheme for molten-salt breeder reactors proposes that components which can be easily decontaminated will be repaired and reused as spare parts. As an aid in evaluating this proposal, a study was started to determine the effectiveness of decontamination procedures in reducing the activity of contaminated parts from the MSRE. The first item used in the study was the manip- ulator hand which had been used in the sampler- enricher system during months of power operation, until it was replaced because of a leak in the boot. When removed from the sampler-enricher, the unit was contaminated by mixed fission prod- ucts to a level of approximately 300 r/hr at 3 in, 40 It was necessary to first remove the two-ply plastic boot which had provided fission product containment for the manipulator while in use. Since the ORNL facility used for routine decontami- nation can handle only up to 50 r/hr of gamma radiation, it was necessary to use a facility at the High-Radiation-Level Analytical Laboratory. This facility was equipped with remote handling equipment and high-pressure sprays which were used in the procedure. The decontamination began with spraying the manipulator hand with a 500-psi jet of detergent. The unit was then soaked in several solutions. After the radiation level was reduced to less than 3 t/hr, the manipulator was removed to a laboratory ’ hood and hand scrubbed with a wire brush and another detergent. The radiation level was finally y reduced from the initial reading of ~ 300 r/hr to a final level of 300 mr/hr, which was low enough to permit controlled direct contact for repair. The detailed results of the various treatments in re- ducing the radiation level are shown in Table 2.1, One of the complicating factors affecting the decontamination of the manipulator hand was the presence of many crevices in the linkages and pins which could not be reached by the jet spray. Extending the cleaning time per step might have resulted in greater decontamination factors. Fur- ther studies will be made of this effect as other components become available, The decontamination of the manipulator hand served a dual purpose. It not only helped de- termine the effectiveness of decontamination procedures in reducing the activity of MSRE con- taminated components, but also made available at a decontamination cost of $300 a spare part that would have cost about $2000 to duplicate. 2.4 DEVELOPMENT OF A SCANNING DEVICE FOR MEASURING THE RADIATION LEVEL OF REMOTE SOURCES Robert Blumberg T. H. Mauney Duniap Scott A method for locating and evaluating concentra- tions of radioactive materials in areas having high background radiation would be useful in fol- lowing the deposition of fission products in com- ponents of circulating fuel reactors and in chemi- cal process plants. Location of unusual deposits would aid in understanding the operation of the 41 Table 2.1, Readiation Levels Following Steps in Decontamination of an MSRE Sampler-Enricher Manipulator Hand Cleaning Radiation Level Cleaning Treatment Time After Treatment (min) (r/hr) None (as received) 300 Detergent (Duz) jet spray 5 100 Rinse (water) Detergent {(Duz) jet spray ‘ 5 75 Formula 50 10 25 Dilute HNOS 10 7.5 Bolt freeing solvent with acetone ringe 5 5.0 Concentrated HNO, 10 3.0 Scrub with scouring powder (BaB-0Q)7 5 0.5—-1.0 Scrub with scouring powder (BaB-0) 5 0.2-1.0 Formula 50 5 No reduction Concentrated HN(}3 5 No reduction Scrub with scouring powder (BaB-0) 5 No reduction Clean with concentrated HN03 5 0.2—-0.3 Acetone rinse 5 Ne reduction Smearsb 800—13,000° fRemoved to laboratory hood. bzirconium and ruthenium. ®In disintegrations per minute. system and in planning for maintenance of the components. On the basis of these needs, a study of possibly useful devices was started. The concept of the pinhole camera using radia- tion shielding and gamma-ray-sensitive film was evaluated along with a system using a portable collimator and a small gamma-sensitive dosimeter. It was determined that while the camera method was feasible, the results to date did not provide very good resolution of position, intensity, and size, The problem of calibration of the film for use in radiation fields of unknown energy appeared difficult. However, the time required to survey an area would be short, approximately equal to the required exposure time for the film. The use of the collimator method would take longer to com- plete a survey, but possibly could yield better resolution. This method aiso offered the potential of giving information on the gamma-ray energy spectrum of the source. A series of experiments with the collimator was conducted to evaluate the method, with the results described below, Experiment with a 5-curie '37Cs Gamma Source A collimator was constructed by casting lead in a 4-in.-diam steel pipe 48 in. long and providing a l-in.-diam hole the length of the tube. The col- limator tube was placed in a suitable hole in the portable shield which was positioned over a pit containing a S5-curie 12®7Cs source. Radiation measurements were made with the equipment and source arranged as shown in Fig. 2.3. The radiation source could be raised and lowered, and the collimator tube could be moved horizontally over the source. Measurements were made with the source 13 ft and 15 ft 6 in. below the top of the portable shield. The source was about 1/4 in. in diameter and 2 in. long and was enclosed in a 1-in.-diam tube 5 in. deep in the carrier, The results of the radiation measurements are shown in Fig. 2.4. It will be noted that as the coliimated ion chamber was moved horizontally over the source, the radiation readings changed PORTABLE SHIELD —.__ 42 CORNL-DWG 67— 11789 - GAMMA CHAMBER {'%4 in. 00) s N < imr/ hr) : 1 RADIATION LEVEL 30 25 20 4-in. LEAD COLLIMATOR .~ ‘ ~—4-in. COLLIMATOR HOLE 48 in. LONG - | | COLLIMATOR MOVEMEMNT = PR | e R i 1" CENTERLINE -~ : | | _~CESIUM -7 GAMMA SOURCE e Fig. 2.3. Collimator-Source Geometry for Point-Source Test. ORNL-—-DWG €7—11790 ““““ ; e = 1 | | | | i ' | : ! ‘ i ; ; f ? e DISTANCE 43.0ft VERTICALT[™ 17 77— (SOURCE OUT OF CARRIER) ; e DISTANCE 455t VERTICAL | {SOURCE OUT OF CARRIER) ‘ . DISTANCE 155ft VERTICAL | % (SOURCE IN CARRIER) 3 5 X ! ; Z LJ © Ll O o 2 | o] } v | ‘i i | " L J - 5 4 3 2 1 o 1 2 3 4 5 COLLIMATED SOURCE POSITION TO LEFT OR RIGHT (in) Fig. 2.4. Collimated Radiation from 5-curie 137¢. source. significantly, indicating clearly the collimating effect. The flat top at the peak was of the same width as the collimator hole and represents the resolution of the collimator when used with a point source, Gamma Scon of the MSRE Heat Exchanger A gamma radiation scan was made on May 19, 1967, over the area in the reactor cell containing the fuel-to-coolant-salt heat exchanger and fuel salt line 102, To accomplish the radiation scan, the portable maintenance shield was located over an opening in the reactor cell above the heat exchanger, and the collimator and gamma ion chamber were in- stalled in a hole in the portable shield. Measure- ments were made by moving the portable shield until the ion chamber was directly above the de- ORNL-DWG &7-11721 PORTABLE l SHIELD., .7” Z ION CHAMBER Y 1it-71in. 1 ft-2in. RO O —, /I 48-in. COLLIMATOR TUBE (4% in. OD) 18 £1-9 in, 13 fi=2 in. ~—— = |NCREASING REACT_OR— - > VESSEL HEAT EXCHANGER (_(N* - 5 ft—3in, ~ | LINE 102 YT T T/ - Y — U )L FF 102 Fig. 2.5. Source-Detector Geometry for Gamma Scan of MSRE Heat Exchanger. sired points in the cell. Figure 2.5 shows the vertical distances from the tip of the ion chamber to items in the cell. The orientation of the various ion chamber traverses of the cell along with lines of equal radiation levels are shown superimposed over the heat exchanger and line 102 in Fig. 2.6. It will be noted that the maximum radiation levels were 3.5 t/hr above the heat exchanger center line and 300 mr/hr above line 102, When the collimator was removed, the radiation level at the top of the hole in the portable shield was 100 r/hr. These results show that the collimated gamma ion chamber measures radiation levels with very good resolution of position, even in the presence of high background radiation levels. Therefore, this method of radiation measurement should be useful in locating accumulations of fission products in reactor components and in planning maintenance operations. Gamma Energy Spectrum Scan of the MSRE Heat Exchanger On May 20, 1967, a gamma energy spectrometer was set up over the same area in the reactor cell, and measurements were made at several points over the fuel-to-coclant-salt heat exchanger. This exploratory experiment was made primarily to evaluate the method. The measurements were made by using a 3- by 3-in, Nal crystal and photomultiplier tube mounted in a lead shield which had a collimating hole 1/32 in. in diameter and 7 in. long under the crystal. The array was mounted on the hole in the portable shield in the same manner as the collimator for the gamma scan described earlier. There were strong, well-defined peaks at about 0.48 and 0.78 Mev corresponding to '93Ru and 93 Zr-9°Nb. The resolution of the Nal crystal was not good enough to determine whether the second peak was from °°Nb only or from the °°Zr decay chain. The energy resolution could be improved by using a different crystal, since the collimator did not appear to degrade the peaks. The problem of assigning absolute disintegration rates to these isotopes is complicated by the geometrical ar- rangement of the heat exchanger and heaters, the effect of gamma energy on absorption by the col- limator, the source distribution in the heat ex- changer, and the counter efficiency. It is be- lieved, however, that some measure of the relative 44 ORNL-DWG 6711742 « x X < s < w 0 M~ - - 2 < ! v ol a Q o 1l il Il il I I i I Il > > 3 > S S S | HEAT EXCHANGER - X =99, "REEZE FLANGE 101~ | | ;,.g,/" FREEZE FLANGE 102 LINES OF EQUAL RADIATION LEVEL Fig. 2.6. Results of Gowmma Scon of the MSRE Heat Exchanger. disintegration rates can be obtained by comparing is believed that these differences in the ratios the ratio of the areas under the counting peaks represent a difference in the deposition of the two for different isotopes at one position with the isotopes at the two positions. Much additional ratio found at other positions. For the measure- work would he necessary to establish the method ments made at point 4 on Fig. 2.6, the ratio of for quantitative measurement under a particular Ru to Zr-Nb was 0.66, while at point B the ratio set of conditions, but it might be the only non- was 1.20, destructive method for making such deposition Although not clearly understood at this time, it measurements. 45 2.5 PUMPS needed at the MSRE. The pump tank is being in- stalled in the pump test facility, and the facility P. G. Smith A. G. Grindell is being prepared for tests with molten salt to investigate the adequacy of the running clearances Mark- 2 Fuel Pump and the performance of the xenon stripper. As reported previously,? the Mark-2 (Mk-2) fuel § v Koty Elewants fee MERE Epel pump was designed to give more salt expansion space in the pump bowl, thus eliminating the need for an overflow tank. As a consequence of the deeper bowl, the Mk-2 rotary element has a longer unsupported length of shaft than the Mk-1. The device for removing gaseous fission products from the salt is also different in the Mk-2 bowl. Fabrication of the Mk-2 pump tank was com- pleted, and the rotary element was modified so that the joint between the bearing housing and 2Ibid., p. 64. shield plug can be seal-welded should the pump be 3tbid., p. 65. and Coolant Salt Pumps The spare rotary element for the coolant salt pump® was assembled, tested, and fitted for service at the MSRE. The joint between the bearing housing and the shield plug was seal- welded to eliminate the possibility of an oil leak- age path from the lower seal catch basin to the PHOTO 74199 TENSION OR COMPRESSION . TSB : ; NOZZLE | ey AT TACHMENT [MOMENT } : Fig. 2.7. Experimental Apparatus for Nozzle Stress Tests. pump tank, It is now available, along with a spare fuel pump rotary element, for service in the MSRE. Stress Tests of Pump Tank Discharge Nozzle Attachment The stress tests of the discharge nozzle on the Mk-1 prototype pump tank® were completed, and the tank assembly was transferred to the Metals and Ceramics Division for examination. A photo- graph of the experimental apparatus is shown in Fig. 2.7. Strain data were obtained under a pres- sure of 50 psi in the tank and a moment of 48,000 in,-1b applied to the nozzle. The measured strains were converted into stresses by use of conventional relationships. For comparison, the stresses were calculated by means of the CERL- IT code, which is written for a nozzle attached radially to a spherical shell. The actual configura- tion is a nonradial nozzle attached to a cylindri- cal shell near the junction of the shell to a formed head. The computer code underestimated the ex- perimentally measured pressure stresses by a factor of about 4.6 and overestimated the moment stresses by about 3.5 times. When the stresses are combined, the code-calculated stresses are higher than the measured stresses for the test 46 conditions. Based on our present knowledge, the nozzle attachment appears to be adequate for the intended service. MSRE Oil Pumps Two oil pumps were removed from service at the MSRE, one because of excessive vibration and the other because of an electrical short in the motor winding. The one with excessive vibration had one of two balancing disks loose on the shaft. The loose disk was reattached, and the rotor, in- cluding shaft assembly and impeller, was dy- namically balanced. This pump has been reas- sembled, has circulated oil up to 145°F, and is ready for service at the MSRE. The pump with the shorted motor winding has been rewound, reas- sembled, and is circulating oil. Oil Pump Endurance Test The oil pump endurance test* was continued, and the pump has now run for 35,766 hr, circulat- ing oil at 160°F and 70 gpm. *Ibid., p. 66. 3. Instruments and Controls L.. C. Qakes 3.1 MSRE OPERATING EXPERIENCE C. E. Mathews J. L. Redford R. W. Tucker Instrumentation and control systems continued to perform their intended functions. Component fail- ures did not compromise safety nor cause excessive inconvenience in the reactor operation. A moderate amount of maintenance was required, as described in the sections that follow. Control System Components The 48-v dc relays continued to suffer damage because of heat developed in the resistors built into the operating coil circuits. Because of this problem, redesigned relays had been procured and some had been installed.! After several months, some of these also showed signs of detenoration. Therefore, following run 11, all 139 relays operat- ing from the 48-v dc system were modified by re- placing the built-in resistors with exterally mounted resistors. This was done with the relays in place, without disconnecting control circuit wiring, thus avoiding the possibility of wiring mistakes inherent in relay replacement. No trouble was experienced after the modification. Component failures were as follows. The coil in a solenoid valve for the main blower 3 backflow damper developed an open circuit. The fine-position synchro on rod drive 2 developed an open circuit in the stator winding and was replaced at the shutdown in May. The fuel overflow tank level transmitter was replaced after a ground developed in the feed- back motor. The ground was subsequently found MSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4119, p. 71. 47 and corrected. The span and zero settings of the differential pressure transmitter (PdT-556) across the main charcoal beds shifted drastically. Al- though the pressure capability had never been ex- ceeded, the diaphragm had apparently suffered damage. Thus, when a new transmitter was in- stalled in June, hand valves were installed in the lines connecting to the off-gas system to be used in protecting against unusual pressures during system operations. The 1-kw 48-v-dc—to~120-v-ac inverter, which supplies ac power to one of the three safety channels, failed during switching of the 48-v dc supply. It was repaired by replacement of two power transistors. Only one themmocouple (TE-FP-10B) failed due to an open circuit. (See p. 22 for discussion of thermocouple petformance.) Water was admitted to the steam dome on fuel drain tank 1 on several occasions before the cause was found to he an intermittent fault in a control module. A freeze-flange temperature control module failed because of excessive condenser leakage. At the next shutdown, all similar condensers were checked, and 33 were replaced because of leakage. Nuclear Instruments Four of the eight neutron chambers had to be replaced. The fission chamber for wide-range counting channel 1 was replaced because of a short circuit from the high-voltage lead to ground. The fission chamber for wide-range counting channel 2 was replaced after moisture penetrated the protec- tive cover on the cable. Moisture leakage into the cable to the BF , chamber required that the cable and chamber be replaced. The output of the ion chamber in safety channel 2 decreased drastically during a nonoperating period, and the chamber was replaced prior to resumption of operations. The 48 trouble proved to be a failure in a glass seal that allowed water to enter the magnesia insulation in the cable, which is an integral part of the chamber. Safety System A period safety amplifier failed when lightning struck the power line to the reactor site, and a replacement amplifier failed as it was being in- stalled. The field-effect transistor in this type of amplifier is susceptible to damage by transient voltages, and it was found that under some condi- tions, damaging transients could be produced when the amplifier is removed from or inserted into the system. A protective circuit was designed, tested, and installed on the spare unit; in the interim, a different and more stable field-effect transistor was installed. The module replacement procedure has been modified to reduce the possibility of damage incurred on installation of the module. Experience at other sites with the type of flux safety amplifiers used in the MSRE had shown that the input transistor could be damaged, causing erroneous readings, if the input signal became too large too rapidly. Therefore, a protective network of a resistor and two diodes was added to the input circuit of each of these amplifiers. Two relays in the safety relay matrices failed, both with open coil circuits. A chattering contact on the fuel pump motor current relay caused safety channel 2 to trip several times before the problem was overcome by paralleling two contacts on the same relay. A defective switch on the core outlet temperature also caused several channel trips and one reactor scram before the trouble was identified and the switch was replaced Four rod scrams, all spurious, occurred during operation in the report period. Two were caused by general power failures during electrical stoms. Ancther occurred during a routine test of the safety system when an operator accidentally failed to resef a tripped channel before tripping a second channel. The other scram, also during a routine test, came when a spurious signal from the switch on the core outlet temperature tripped a channel while another channe! was tripped by the test. 3.2 CONTROL SYSTEM DESIGN P. G. Hemdon As experience showed the need or desirability of more information for the operators, improved per- formance, or increased protection, the instiumenta- tion and controls systems were modified or added to. During the repoit period there were 25 design change requests directly involving instruments or Six of these required only changes in Fourteen coitrols. process switch operating set points. requests resulted in changes in instruments or controls, one was canceled, and the remaining four were not completed. The more important changes are described below. The single £32-v dc power supply formerly serv- ing the nuclear safety and controls instrumentation was replaced with three independent power supplies, one for each safety channel. Each obtains #32-v dc from 110-v ac, but the ac power sources are dif- ferent. One is the nomal building ac system. Another is 110-v ac from a 50-kw static inverter powered by the 250-v dc system. The third comes from a 1-kw inverter operating on the 48-v dc sys- tem. Thus continuity of control circuit operation is ensured in the event of a single power supply or power source failure. It also increased the re- liability of the protection afforded by the safety system by ruling out the possibility that malfunc- tion of a single voltage-regulating circuit could compiomise more than one channel. (Before the change, on one occasion, a failure in part of the regulating circuit caused its voltage output to in- crease from 32 to 50 v, and only a second regulator in series prevented this increase from being im- posed on all the safety circuits.) To prevent the reactor from dropping out of the “‘Run’’ control mode when a single nuclear safety channel is de-energized, the ‘““nuclear sag bypass’’ interlocks were changed from three series-connected contacts to a two-of-three matrix. Circuits were installed to annunciate loss of power to the control rod diive circuits. This re- minds the operator that he cannot immediately re- tum to power simply by manually withdrawing the rods after the controls have dropped out of the “Run’’ mode. A wiring error in a safety circuit was discovered and corrected. Interlocks had recently been added in the ‘‘load scram’’ channels to dwop the load when the control rods scram. A wiring design error re- sulted in these interlocks being bypassed by a safety jumper. Although the circuits were wired this way for a time before being discovered, the scram interlocks were always operative during power operation, since the reactor cannot go into the ‘‘Operate’’ mode when any safety jumper is inserted. The rate-of-fill interlocks, including the pneu- matic instrument components in the drain-tank weighing system, were removed from the control circuits for the fuel drain tank helium supply valve. Experience had shown that these interlocks, in- tended to limit the rate at which fuel could be transferred to the core, were not required, since physical restrictions in the helium supply posi- tively prevent filling as fast as the design set point on the jinterlock. Four new photoengraved graphic panels incorpo- rating recent circuit changes were installed on the control and safety circuit jumper board. The main control board graphics were also revised to show changes in the fuel off-gas system. Test circuits were originally provided to check the operability of electronic safety instruments on fuel loop pressure and overflow tank level. These were revised to make it possible to observe the value of the test signal current at which the re- lays operate. The measuring range of the matrix-type flow ele- ment on the upper seal gas flow from the coolant 49 pump was increased from 5 to 12 liters/hr. This was to keep the instrument on scale at pump bowl pressures higher than originally anticipated. A new capillary flow element was designed to replace the clement that measures the reactor cell evacuation rate, raising the range from 1 to 4 liters/min. To avoid reducing the reactor system helium supply pressure below the minimum limit when large quantities of helium are required for fuel processing, the fuel processing system helium supply line was removed from the 40-psig supply header and reconnected upstream to the 250-psig main supply header. Pressure-regulating instm- ments were installed. To provide more information to the operator, 20 additional pairs of themocouple lead wires were installed between the vent house and the data logger. Connections at the patch panel were made available by removing nnused lead wires from some radiator thermocouples. A new conduit for these lead wites was also installed between the vent house and the existing cable trough in the coolant drain cell. 4. MSRE Reactor Analysis 4.7 INTRODUCTION Table 4.1, lsotopic Composition of 233y Available for MSRE® I?. N. Haubenreich B. E. Prince o Fraction The experimental program planned for the MSRE Uranium Isatope (at. %) includes operation for over half a year with 233U as the fissile material. The fuel salt presently in 230 0.022 the reactor will be fluorinated in the MSRE process- 233 91.5 ing facility to remove the uranium (33% 233U). 234 7.6 Then 233U will be added to the stripped carrier 235 0.7 salt as the LiF-UF , eutectic (73-27 mole %). 236 0.05 During this report period, we calculated most of 938 0.14 the important neutronic properties of the reactor, operating with 233U fuel. These results will be - . . . . ORNL communication, J. M. Chandler to J. R. Engel, used in a safety analysis and in planning the start- May 10967, up experiments. We also made some computations of neutron spectra to use in evaluating the relevance of MSRE 233U experiments to molten-salt breeder design calculations. The techniques of analysis for the new fuel (Table 4.1) is the actual composi- tion of the uranium available for use. were similar to those used previously for the MSRE Wlth.235U fl:lel. The results are discussed 'm the 4.2 NEUTRON ENERGY SPECTRA IN sections which follow, and references are given ) . i .. MSRE AND MSBR which provide more detailed descriptions of the methods of analysis. The fluorination process removes uranium and some fission products, but leaves plutonium and most fission products in the carrier salt. It was necessary, therefore, to make some assumptions regarding these concentrations at the time of the startup with ?33U. It was assumed that all the uranium is removed and that all the plutonium and samarium remain in the salt. We assumed that the changeover would be made at an integrated power of 60,000 Mwhr and computed the plutonium and samarium concentrations from the time-integrated production and removal rates for these nuclides. Fission products other than samarium were not c?nsidere‘d explicitly in the base-line ?a‘lcul-ations, ORNLA4119, . 70, s1r?ce their net effect c?n th.e core_ react1v1t3.z ¥s 2SR Program Semiann. Progr. Rept. Feb. 28, 1967, gquite small. The uranium isotopic composition ORNI.-4119, p. 193. B. E. Priince In a recent report, we described the results of computational studies of MSRE neutron spectra for the present 23°U fuel salt.! We have extended these studies to the spectra with the 233U fuel and have compared the results with corresponding spectra for a typical MSBR core lattice currently under design study.? The characteristics of the MSBR core design used for these calculations were taken from ref. 2. The graphite-moderated portion of the core was 10 ft long and 8 ft in diameter and contained fuel and 'MS& Program Semiann. Progr. Rept. Feb. 28, 1967, 50 51 ORNL-DWG 67-14793 LETHARGY 8 7 6 5 4 3 ™ e~ 0 4 i [ T 1 T [ 1| BT T : T4 ; bl‘il'f\ » - | Jd \ MSRE- 27U FUEL W 3 ‘ l‘-—-“"‘- —6 | - ; R UNIT LETHARGY (normalized to one fission source neutron £ FLUX P M O b C 10! ,,,,,, 108 ENERGY (ev) Fig. 4.1. Calculated Epithermal Neutron Spectra in the 2331)_Fyeled MSRE and in the MSBR. fertile salt volume fractions of 16.48 and 5.85% respectively. The fuel salt contained approxi- mately 2.84 mcle % UF , (~64% of total U is 233y and ~5.9% is 225U), and the fertile salt contained 27 mole % ThF . By compatson, the model of the MSRE core used for this analysis was a cylinder with effective dimensions of 58 in. in diameter by 78 in. in height. The fuel salt volume fraction is 22.5%, and there is no fertile salt. content of the fuel salt was approximately 0.125 mole % UF (™ 01.4% 2331)), based on calculations summarized in the following section. As described in ref. 1, the epithermal and themal neutron spectra calculations were made with the GAM and THERMOS programs respectively. In the choice of an effective ‘“‘thermal cutoff’’ energy separating these spectra, there is a considerable degree of latitude, with the single restriction that the cutoff energy be high enough that important calculated neutronic properties at operating tem- perature are not strongly influenced by neglect of The uranium upscattering corrections above the cutoff energy. In previous studies of the MSRE, '3 we used 0.876 ev {(for which the present energy group structures for the GAM and THERMOS programs coincide). This value was found to be sufficiently high to satisfy the preceding criterion.?® However, because current design calculations for the MSBR are based on the higher value of 1.86 ev, to make a valid comparison we calculated the MSRE thermal spectra for this higher cutoff energy. The results of GAM-II calculations of the epi- thermal neutron spectra in the two reactors are shown in Fig. 4.1. As in ref. 1, use is made of the lethargy variable, proportional to the logarithm of the neutron energy, in plotting the results. This relation is lethargy = — In (energy/10 Mev) . For a unit fission sourtce the calculated fluxes per unit lethargy for each GAM-II neutron group are 3P. N. Haubenreich et al., MSRE Design and Opera- tions Repaort. Part IIl. Nuclear Analysis, ORNL-TM- 730, pp. 38--40. (normalized) R UNIT LETHARGY ,_ - FLUX P 0024 0.020 52 16 LETHARGY 17 I 0024 3 Q N = 0020 - £ & £ > 0016 | & o 3 2 0004 0.0 0.2 05 ENERGY (ev) Fig. 4.2, Calculated Thermal Neutron Spectra in the MSRE with 20 19 o MSRE - 233 ) FUEL THERMAL LETHARGY 17 ; ORNL-DWG €7 -11724 T 1117 FLUX SPECTRA AT THE CENTER OF THE FUEL CHANNEL ARE PLOTTED. FLUXES AREL NORMALIZED TO EQUAL. MAGNITUDES OF TOTAL NEUTRON FLUX i L BELOW 1{.86 ev, AVERAGED OVER THE LATTICE CELL. MSRE—-23%U FUEL 10 235 U and 233U Fuel Salts. ORNL--DWG 6714795 14 FLUX SPECTRUM AT CENTER OF FUEL CHANNEL GRAPHITE 0.046 0.042 0o.008 c.004 POSITIONS FOR THERMAL FLUX PLOTS A *B oC - — l ! 2 4 & 8 RADIAL POSITION (cm) MSBR CELL GEOMETRY MSRE AND MSBER FLUX PLOTS ARE NCRMALIZED TC EQUAL MAGNITUDES OF TOTAL NEUTRON FLUX BELOW 1.3&ev, AVERAGED OVER THE LATTICE CELL Fig. 4.3. Calculated Thermal Neutron Specitra in the ENERGY (ev) 23 20 3U-Fueled MSRE and in the MSBR. FUEL / FERTILE - FUEL,%«\#}‘ GRAPHIT E’He 53 shown as solid points in Fig. 4.1 (points shown only for the MSRE 233U spectra). The calculated points are connected by straight line segments. As might be expected from the gross similarities be- tween the two reactor systems (temperatures, mod- erating materials, total salt volume fractions), there is a matked similarity between the two spectra. Above energies of about 1 kev, the MSBR flux spectra lie above that of the MSRE, due to the smaller neutron leakage from the MSBR core. Below about 1 kev the spectrum in the MSBR becomes pro- gressively reduced by resonance neutron abscrption in the fertile material, and the two flux spectra tend to approach one another in magnitude. Results of THERMOS themal spectra calculations are shown in Figs. 4.2 and 4.3. Figure 4.2 com- pares the MSRE spectrum for the 237U {uel sait with that for the current 235y fuel loading. The spectra at the center of the fuel channel are chosen as a basis of comparison. In Fig. 4.3, the MSRE thermal spectrum with the ?3°U fuel salt is com- pared with the cortesponding spectra at several points in the MSBR lattice. Note that curve 4 has the same relative position in the MSBR lattice (center of the larger fuel salt channel) as that of the MSRE. For both these reactors the flux spectra of Figs. 4.2 and 4.3 are nomalized to give the same total integrated neutron flux below 1.86 ev, averaged over the total volume of salt and graphite. On this basis, relative comparisons can be made of the enerpy (or lethargy) distributions and position dependence of the spectra in each reactor. One should note, however, that the actual magnitudes of the total thermal flux will be quite different in the systems, depending on the relative fission densities and the fissile materdal concentrations. In Fig. 4.2, the slight “energy hardening,”” or preferential removal of low-energy neutrons for the 2350 fuel loading, reflects the larger uranium con- centration for this case (V0.9 mole % total UF,, 33% eariched in 235U). Again, in Fig. 4.3 the higher concentration and absorptions in fissile uranium in the MSBR and also the absorptions in the fertile salt produce a hardening of the energy spectrum as compared with the MSRE. Just as in the case of the epithermal spectra, however, the similar temperatures and gross material compo- sitions result in a marked similarty in the spectra. An altemate means of comparing the neutron spectra in the two systems is shown in Figs. 4.4 and 4.5. Here the calculated group fluxes have been multiplied by the corresponding group micro- scopic fission cross section for 222U, summed over the energy groups above each energy point, and nommalized to one fission event in 233U occurring in each of the thermal (£ < 1.86 ev) and epithermal (£ > 1.86 ev) energy regions. In the latter case, most of the fission reactions occur between the lower cutoff energy and about 30 ev, where the flux spectra are quite close to one another (Fig. 4.1). As a result, there is a close cortespondence in the normalized distribution of epithermal fission events in the two systems. Larger differences are encountered in the thermal fissions, as shown in Fig. 4.5. Note, however, that the spectra at the center of the fuel channels ORNL--DWG 6714796 TPITHERMAL {>1.86 ev) FISSIONS z' ABCVE ENERGY £ RACTION OF —_ F 1l Ll 100 2 5 10t ENERGY (ev) Fig. 4.4. Normalized Distributions of Epithermal Fission Events in the 233U-Fue|ed MSRE ond in the MSBR. FRACTICN OF THERMAL («188ev) FISSIONS CRNL -DWG 67-1797 e e FLUX SPECTRA TAKEN AT 0B . THE CENTER OF THE | FUEL CHANNEL W ‘ & 06 NN msas & \ L = | o o o4 MSRE -233)) FuEL 239 py +0.0122 “Fuel not circularing, contrel rods withdrawn to upper limits. 3 PBased on 73.2 ft” of fuel salt at 1200°F in circu- lating system aud drain tanks. c . - . . - Increase in uranium concentration required to main- tain criticality with one rod inserted to lower limit of travel, fuel not circulating. c"At initial critical concentration. Where units are shown, coefficients for variable x are of the form 51(/1(8)(; otherwise, coefficients are of the form xak/kfix. ®Isotopic compositions of uranium as given in Table 4.1, These are the results of two-dimensional, four- group diffusion calculations, with the group cross sections averaged over the neutron spectra shown in the preceding section. Both RZ and R geo- metric approximations of the reactor core were con- sidered, in the manner used previously to analyze the physics experiments with the *2°U fuel.* One innovation was the use of the two-dimensional multigroup EXTERMIN ATOR-2 program,® a new version which includes criticality search options and a convenient perturbation and reactivity coef- ficient calculation. The properties listed in Table 4.2 differ in several important respects from the correspending properties of the present ?35U fuel salt. The {issile material concentration is only about one- kalfl that for the present fuel, an effect which re- sults mainly from the absence of a significant quantity of #38U in the new loading. This same effect produces an increase in the thermal diffusion length in the core, thereby increasing the thermal- nentron leakage (as reflected in the larger magni- tudes of the fuel temperature and density coeffi- cients of reactivity). The increased diffusion length also results in an increase in control rod effectiveness (2.75% 8k/k calculated for one rod, as compared with 2.11% calculated and 2.26% measured worth of one rod in the present *°U- fueled reacter).® On the percentape basis given in Table 4.2, the uranium concentration coefficient is higher for the #3%U fuel by a factor of about 1.8, owing to the absence of the 238U and also to the greater neutron productivity of the *?%U. On the basis of 1 g of excess uranium added to the fuel circulating system (70.5 ft3), the reactivity change is +0.00123% S&/k. This is about 3.8 times the corresponding reactivity change when 1 g of 23°U is added to the present fuel salt. Fission Rate and Thermal Flux Spatial Distributions Spatial distributions of the fission density in the fuel salt, obtained from the EXTERMINATOR dif- fusion calculations described in the preceding section, are shown in Figs. 4.6 and 4.7. Figure 4}3. E. Prince ef al., MSRE Zero Power Physics Ex- periments (ORNL report in preparation). >). B. Fowler ot al., EXTERMINATOR-2: A Fortran IV Code for Solving Multigroup Neutron Diffusion Equa- tions in Two Dimensions, ORNL-4078 (April 1967}. 55 4.6 shows the axial distribution of the fission density at a radial position of about 7.3 in. from the core center line, which is the approximate radial position of maximum flux with all control rods withdrawn. The fission density in the salt is nomalized to 10°® fission events occurring in the entire system. Also shown in Fig. 4.6 is the axial distribution of neutron importance for the fast group, which is used in the calculation of the effects of circulation on the delayed precursors (see later section). Figure 4.7 shows the comrresponding radial dis- tributions of fission density in the salt, both with all rods withdrawn and with one and three rods fully inserted. These curves are based on calcu- lations with an R0 model of the core geometry. The approximate geometry used for the control element and sample-holder configuration is indicated schematically in the figure, and the angular position of the flux traverse along the reactor diameter is also shown. In the position where the traverse intersects the control element (Fig. 4.7), the curves for the center portion of the salt-graphite lattice are connected with those for the major part of the lat- tice by dashed straight lines extending through the control region. The cortesponding distributions of themal flux are very similar in shape to the fission density distributions. These are given in Figs. 4.8 and 4.9. The magnitudes of these fluxes are based on 1 Mw of fission energy deposited in the system, using the conversion factor of 3.17 x 10'% fis- sions/Mwsec. The flux magnitudes are also nommalized to the minimum critical uranium loading, s0 that renormalization of these values would be necessary to apply them to the exact operafing concentration. For this purpose the flux can be assumed to be inversely proportional to the 2270 concentration. Note alse that, in these calcula- tions, the thermal neutron {lux is defined as the integral flux below a cutoff energy of 0.876 ev (see the discussion in Sect. 4.32). Reactor-Average Fluxes and Reaction Cross Sections Cross sections for important nuclide reactions in the MSRE 23U spectrum are listed in Table 4.3. These values of the thermal and epithermal aver- ages are based on a cutoff energy of 0.876 ev. To obtain the final column, we have used the average 56 ORNL-DWGC 67-441738 4.8 AXIAL CENTER LINE OF CORE ,,,,,,,,,,,, I v x — 5 T | z 2 - S - e ~{a0 ¢ © ™~ I > O \ o o = < l < a Pl \\ g | é " ; — 5 5 81 """ 132, 2 = = 2 S 2 v Q. o = o g - - = A28 4= N e 24 = & | o Z RADIAL POSITICN, 7.28 in. FROWM E > o ._ @ P wJ O =z Q o o L Lef = 1.6 5 O e o }._ wy < L 8 ] - 08 - N N . * - ol -der L L L . 0 20 30 40 50 80 AXIAL DISTANCE (in.) Fig. 4.6. Axial Distributions of Fission Density and Fast Group tmportance in the MSRE with 233 Fyel Loading. ORNI-DWG §7-14799 T 10 : R {27 FISSION DENSITY s FISSION DENSITY TRAVERSE = fi) TRAVERSE e ~ 3-5 } ------ | i | T T 30 THREE RODS WITHDRAWN 1’77“\ J S CONTROL ROD AND SAMPLE HCLOER CONFIGURATION 25 § ROD NO.2 INSERTED - - \ = -THREE RODS INSERTED - \\ ~ -ROD NO.Z2 INSERTED ! THREE RODS INSERTED FISSION DENSITY IN SALT (fissions/cm> per 08 fissions) }-- SAMPLE HOLDER REGION \ \ ! INORMALIZED TO AXIAL POSITION OF MAXIMUM FISSION DEMSITY) 15 10 5 C 5 10 15 RADIAL DISTANCE (in) Fig. 4.7. Radial Distributions of Fission Density in the MSRE with 2331) Fyel Loading. epithermal-to-thermal flux ratios, determined by magnitude of the thermal fluxes, give the approxi- averaging the EXTERMINATOR group fluxes over mate total reaction rates per atom for neutrons of the volume of the fuel circulating system. These all energies in the spectrum. Table 4.3 also gives eftective cross sections, when multiplied by the the calculated magnitudes of thermal and epithermal « '; A2y —— S - PO = (107 L . o a | . T 0 3 8 | 5 - 3 g W £ oo © Z W = 2 o = z jsed ,,,,,,, O e = G T i o u L 5 ; o = 5 o.l = qal..E 2 I o 'C_) b] z 0 & | ‘r.. = N i | 2330 LOADING > o | i _ < = n- 1l RADIAL POSITION, 7.28 in. FROM AXIAL CENTER LINE OF CORE. FLUX MAGNITUDES NORMALIZED TO MINIMUM CRITICAL QRNI-DWG 67-11800 - ¢r,TH 30 40 AXIAL DISTANCE f{in)) Fig. 4.8. Axial Distribution of Therma! Neutron Flux in the MSRE with 233 Fuel Loading. CRNL -DWG 671301 . T TRy e m e | I T T T tTttTT Tttt (x 40') | | o | ‘ ‘ 9 THREE RODS WITHDRAWN — | ; } | /THHEE RODS WITHDRAWN | z | || r \3/ . | | | FLUX N FLUX Fa & || TRAVERSE (5 | 7g) TRAVERSE § N\ L I l . ! - o NN {1\ 'e T \ Y~ ROD NO. 2 c | I INSERTED CONTROL ROD AND SAMPLE HOLDER = CROD NO. 2. CONFIG N MEERTED | SONFIGURATION = | & | & _ THREE RODS __ | x S INSERTED | -\ i’ . ‘ | é 4 - | | NOTE: FLUX MAGNITUDES \\ i \ | | | NORMALIZED TG POSITION OF W \ \ | MAXIMUM AXIAL FLUX, AND TO _ 3 \ I | MINIMUM CRITICAL #22U LOADING <1 2 | & L = b s CONTRCL | | e SAMPLE HOLOER 5 4 ELEMENT —f o RIEGION ‘ ! | || | o 0 R P R N 20 25 20 15 10 5 0 5 1C 15 20 RADIAL DISTANCE {in) Fig. 4.9. Radial Distributions of Thermal Neutron Flux in the MSRE with 233y Fyel Loading. flux, averaged over the volume of the fuel circu- lating system and normalized to 1 Mw. As dis- cussed in the preceding section, these magnitudes are nommalized to the minimum critical uranium concentration. Note also that these values include the ‘“flux dilution’’ effect of the time the fuel spends in the external piping and heat exchanger. (The corresponding value of the thermal flux, averaged only over the graphite-moderated region of the core, is 3.60 < 1012 neutrons cm ™ ? sec™? Mw™1) 58 Table 4.3. Average Cross Sections for Thermal- and Epithermal-Neutron Reactions in the MSRE Spectium with 233y Fyel Cross Section Averaged over Thermal Spectrum, over Fpithermal Spectrum, Cross Section Averaged Effective Cross Section Nuclide Energy <0.876 ev Energy > 0.876 ev in Thermal Flux (bamns) (barns) (barns) bLi® 442.1 17.7 473.6 L4 0.017 0.0007 0.018 'Be 0.0047 0.0040 0.012 Boron 353.1 14.1 378.2 e 0.0019 <107t 0.0020 " 0.0047 0.0023 0.0088 Zirconium” 0.0865 0.0764 0.222 13%%e 1.32 % 10° 83.3 1.32 x 10° 149%m 3.91 x 10° 92.1 3.93 x 10° 151gm 2900.0 128.7 3127.0 233y (abs) 272.1 45.6 353.2 233y (fission) 249.2 (1 = 2.50) 38.4 317.6 234y 43.5 38.1 111.3 2350 (abs) 290.3 22.3 330.0 235y (fission) 244.2 (1 - 2.43) 13.6 268.5 236y 2.8 18.9 36.4 238y 1.3 16.7 31.0 239py (abs) 1402.6 22.5 1442.7 239y (fission) 836.0 (1 - 2.89) 13.9 860.8 System-average® neutron fluxes (neutrons cm 2 Thermal: 1.48 x 10°° Epithermal; 2.64 x 1012 . . 6_ . “Cross section for the reaction 141(11,(1)3!-1. PNatural isotopic composition, “Rased on 2.00 x 10° ¢m? of circulating fuel. Effect of Circulation on Delayed-Neutron Precursors In previous studies, we described a mathematical model found to be in close correlation with meas- ured values of the delayed-neutron loss caused by circulation of the present 23°U fuel. 46 We have 6B. E. Prince, Period Measurements on the Molten Salt Reactor Experiment During Fuel Circulation: Theory and Experiment, ORNL-TM-1626 (October 1966). applied this same computational model to the ?33U- fueled reactor in order to obtain the reactivity change due to circulation and also to obtain the reactivity increments necessary to produce a given stable period during circulation. In these calcu- lations, temperature feedback effects are neglected. Plutonium-239 is present in sufficiently small amounts to be negligible also. The net effect of the circulation in changing the shape of the distribution of delayed neutron emis- sion within the reactor is shown in Fig. 4.10. This - 59 ORNL—DWG 711802 6 ..... [ [ | | 5l l_ A IR T T TG L | e il | £ 4 @il — | e 0 8% &5l = \ T 5wl a1 zZ5 3 sl ot N _ o7 3 LLI (DI - oL = 0w to- (L w3 Su_ 8 Ol QI — = 2F Sl o] 7 DIRECTION OF FUEL CIRCULATION —-w ol | | a® | ©| 52 | S L)t W | 1.14 0.766 x 10 > 6 2.50 0.088 x 1073 3.01 0.280 x 1073 - Total static 8 0.00264 0.006606 Prompt-neutron generation 4.0 x10* 2.4 x 1074 - time, sec Temperature coefficients of reactivity, (OF)_1 Fuel salt 613 x 107 ° —4.84 + 1077 Graphite ~3.23x10°° —3.70 x 107" 100 50 20 Q5 0.05 Ot 0.2 05 ORNL-DWG &7-118B05 PHASE MARGIN (deg) T 238y FUEL 233y FUEL PERIOD OF OSCILLATION (min) 232 FUEL (e EXPERIMENTAL | | - PERICQDS} U FUEL 233 1 2 5 10 POWER LEVEL (Mw) Fig. 4.13. Comparison of MSRE Fredicted Phase Margins and Natural Periods of Oscillation vs Power Level for 2331. and 2:'Lr’l,l-Beui'itng Fuel Salt Loadings. Part 2. MSBR Design and Development R. B. Briggs The primary objective of the engineering design and development activities of the MSR program is to design a molten-salt breeder experiment (MSBE) and develop the components and systems for that reactor. The MSBE is proposed to be a model of a large power breeder reactor and to operate at a power level of about 150 Mw (thermal). A reference design of a 1000-Mw (electrical) molten-salt breeder reactor power plant is to pro- vide the basis for most of the criteria and for much of the design of the MSBE. We have been spending most of our effort on the reference design. Sev- eral concepts were described in ORNL-4037, our progress report for the period ending August 1966, and in ORNL-3996, which was published in August 1966. We selected the modular concept — four 250- Mw (electrical) reactors per 1000-Mw (electrical) atation — that has fissile and fertile materials in separate streams as being the most promising for irmmediate development and have proceeded with mote detailed study of a plant based on that con- cept. Initial results of those studies were re- ported in ORNL-4119, our progress report for the 5. period ending February 1967. During this report period, we continued to examine the details of the reference plant and its equipment. We also sum- marized the objectives and program for develop- ment of the MSBE in ORNL-TM-1851. To date, the component and systems development activity has been concerned largely with support of the design and with planning. The development program was outlined in ORNL-TM-1855 and in ORNL-TM-1856. Salt circulation loops and other test facilities are being modified for tests of the sodium fluoroborate coolant salt, models of graph- ite fuel cells for the reactor, and molten-salt bear- ings and other crucial components of the fuel cir- culation pump. Essentially no experimental work has heen done yet. We plan to complete essential parts of the ref- erence design during the next report period and to begin to design the MSBE. Experimeatal work will begin also but will be limited to a few of the mote obvious important problems. Design and development on a large scale are planned to start early in FY 1969. Design E. 5. Bettis 5.1 GENERAL E. S. Bettis R. C. Robertson The design program during the past peried has primarily been directed toward refinement and modification of the previously reported concepts for the reactor and other major equipment and re- visions to the cell layouts and piping. 63 The major effort was redesign of the reactor to accommodate the new data on radiation effects in graphite. New approaches wete tried to arrive at a core design which could compensate for dimen- sional changes in the graphite, but it was finally decided to simply double the volume of the core to lower the power density from 40 to 20 kw/liter and thereby assure a stable core life of at least ten years. It was also decided to include pro- visions for replacing the complete reactor vessel ORNL-DWG &£7-3081AR Bl ANKET SALT HEAT EXCHANGLR AND PUMP FUSL SALT HEAT EXCHANGER - AND PUMP e COOLANT - . BLANKET 5401 RZACTOR SALT PP R ' 1 PROCESSING | | 2! | | | ' | | e —— | : Pl ! ‘j | FUEL : / : L 3z L o ! | PROCE SSING : T LT P 4 #¥day lpe 23 1 | 4_—} . R ot el GENCRA ¥ ] o NCRATOR ——— ~ . 4 { z —f1 | H_Lfi FRTE I 5272 Mwte) = =l o e : o it | ] ; i 35 : ;el N 3 1 {u‘v . w2 ' i . ' | ‘ o - - L GENERAIOR = 1 o b : v Y 2 o i ILPT ALPT 35077 Mwle) | o - GASEQUS L 3 P F. = 3FISSICN ‘ ‘ I L o i f gflsorggirs _ ! HSPOSAL For — i o VG TE N Loy ONDENSING I SYSTEN: e i ! \ el ?E, FEED WATER HEF SYSTEMS / { T ! — 5 i - BOILER SUPCRHEAT-R REHEATE RS REHEAT STEAM 62 i ‘ Ve ' PREHCAIFRS BCOSTER MIXING L =5 R o PUVP TEE FUEL SAIT FLUSH- BLANKET SALT COCLAN: SALT DRAIN IANKS — SALT DRAIN TANK IRAIN TANK DRAIN TANK FUEL COOLANT STEAM PERFORMANCE DATA POINT ft/sec psig TEMP (°F) POINT t13/sec psig TEMP (¢F) POINT 10% Ib/hr psta | EMP (°F) 1000 Mwle) GENERATION P_AN] f 25.0 131300 L 375 129 1125 50 257 3800 1000 NET OUTPUT 1000 Mwie) 2 250 9 1300 31375 10 1125 51 10,07 3600 000 GROSS GENERATION 1034.9 Muw (e) 3 250 146 1300 32 375 2680 125 52 715 3515 1000 BOILER FEED PUMPS 29.49 NMw 4 250 110 5% 5.0 208 125 53 297 3515 {000 BODSTER PUMPS 9.2 Mwle) 5 250 50 1000 34 50 212 850 54 292 3500 866 STATION AUXILIARY P57 Mwlel & 250 31 1000 35 841 252 1125 55 514 600 552 REACTOR HIAT INPUT 2225 Mwl(t] 7 250 18 1000 36 81 194 350 56 513 570 650 MET HFAT RATE 7601 Biudkw hr 37375 P03 350 57 128 570 550 NET EFFICIENCY 44979 BLANKET 8 375 198 850 58 128 540 000 LEGHND 20 43 125 1250 39 375 164 1111 59 513 540 1000 FUEL —_ 21 43 60 1250 40 375 38 114 60 500 172 708 BLANKET T : COOLANT —_— 22 43 1110 1250 61 430 15inHg 97 ST r A - 23 a3 200 1150 62 716 3500 551 WATER e 24 43 445 1150 63 1007 3475 695 FREEZE VAl vE i 64 1007 3800 700 JATA POINT X 65 257 3800 700 Fig. 5.3, 250 Mw {Efectrical) Module Heat Flow Diagram. va € * rather than just the core and to provide space to store the spent reactor within the biological shield- ing of the reactor cell. While this new design rep- resents the present status of graphite technology, we believe that an improved graphite will be de- veloped within the next few years and that longer core life or higher power density than used here can be expected in the future. Analysis of the thermal expansion stresses in the various systems indicated that rather extensive modifications were needed. Changes were there- tore made in the plant layouts, the equipment sup- ports, and in the arrangement of the salt piping, with the result that the calculated stresses in the vessel attachments, supports, and piping systems now fall well within the acceptable values. Calculations of the radiation flux levels at the reactor cell walls allowed designs to be prepared for the thermal shielding used to protect the con- crete from excessive temperatures. Drawings were also made for the peneral cell wall design, in- cluding the closure blocks at the top, the cell penetrations, etc. Relatively few changes were required in the MSBR flowsheet during the past period, but for convenience it is presented again in Fig. 5.1. The MSBR design study is expected to be com- plete within the next few months. The description, cost, and performance data for the steam system will probably need little change but will be up- dated as required. The August 1966 design study report on a 1000-Mw (electrical) molten-salt breeder reactor power station (ORNL-3996) will be revised to present firmer cost estimates based on the new and more detailed designs for the reactor plant. 5.2 CELL ARRANGEMENT C. E. Bettis W. Terry H. L. Watts C. W. Collins H. A. Nelms W. K. Crowley J. R. Rose H. M. Poly The reactor is now designed on the basis that the entire vessel will be of all-welded construction and arranged for relatively simple replacement after the useful life of the core has been expended. Replacement of the reactor would of course require provisions for disposal of the spent assembly. Rather than lift this relatively large and radio- active piece of equipment from the reactor plant biological shielding, which would require a large amount of heavily shielded transport equipment, room has been provided in the reactor cell to store the spent unit until the second reactor vessel is replaced. In the intervening period the activity level of the first spent unit would have decayed to a more manageable value. This is the basis of the present design, but, of course, the maintenance procedures will continue to receive careful study and review, Storage of the spent reactor vessel in the cell will require an enlargement of the cell volume by about 40%. Since simplification and reatrangement of the interconnecting salt piping were also re- quired to satisfy stress requitements, the layout of the entire reactor plant was rearranged. 'The new layout is shown in Fig. 5.2. The plant consists of four modules, each having a reactor with a thermal output of about 556 Mw. Steam is generated at 3500 psi—1000°F and re- heated to 1000°F in each of the modules, but the flows are combined to serve a single 1000-Mw electrical turbine-generator unit. As shown in Fig. 5.3, each module consists of a reactor cell, a steam-generating and steam reheat cell, and a drain-tank cell, the last being shared with one other module. A drain-tank cell thus con- tains four fuel drain tanks, two for each reactor. The blanket and coolant salt drain tanks are also contained in the two drain-tank cells, The reactor and steam cell elevations ate shown in Fig. 5.4, The equipment in these cells is mounted on rigid supports, with thermal expansion loops being provided in the connecting piping to maintain the stresses within the allowable limits. The reactor vessel is shown mounted on a single column that is pinned at the bottom to allow some lateral movement of the reactor. This arrangement appatrently presents few major design problems and reduces the stresses within certain elements of the system; but, since the stresses are not prohib- itive even without use of the pinned column sup- port, this arrangement may receive further review. In previously reported MSBR concepts the re- actor was connected to the heat exchangers through concentric pipes in an arrangement designed to minimize the salt volumes and to simplify the ther- mal expansion problems. As will be explained sub- sequently, however, the doubling of the reactor core volume made minimization of the fuel sait volume in the piping less important, and this piping change had relatively little effect on fuel cycle 66 3fr 3t 3 1 | Y i A ! | ,;'_ o }, — - | T | j_ -] 454+ I ] li;Y,, e 1 it ‘ [ e oo | ¥ oy S ! 41 i i 148 ft 56 4 | ; ! i e A - i - ! ; : i L I P ; . \ b ‘ ; [ P : : 40 ft CHEMICAL | 3 Pl ; ; PROCESSING L= = bor e | c — - | T D g PROCESS ROOM L Ll b [ 1 T N = 7311 e 178 ft Fig. 5.2. Reactor Salt System Layout. Plan costs, The concentric piping was also found to present stress problems at certain temperatures of operation. For these reasons the piping was changed to provide separate reactor inlet and out- let lines for the fuel and blanket salt streams. The separate piping will also simplify the main- tenance procedures with remotely operated tooling. The coolant, or secondary, salt piping in the steam cell was also modified to include thermal expansion loops and to include fixed supports for the equipment. The location of the coolant salt circulating pump was changed in order tc gain the necessary flexibility in the piping. Radiation shielding, formed of 3-in.-thick steel plates, lines the reactor cell to protect the con- 23 Gins ‘r-«ZO‘T e 23t Gin, - ORNL-DWG 87-10840A - 3f1 3ft - 4t COOLANT DUMP TANKS COOLANT PUMP - REHEATERS3 -STEAM GENERATORS BLANKET DUMP TANKS FUEL DUMP TANKS BLANKET HEAT EXCHANGER PRIMARY HEAT EXCHANGER - REACTOR FUsH SALT TANK OFF-GAS PRCCESS ROOM INSTRUMENTATION HOT CELLS | o and piping for ™~ 1000 Mw (electrical) unit. crete from excessive temperatures due to absorbed . This thermal shield was designed to attenuate a flux of 1 x 101? 1-Mev gammas cimn ™ sec™ ! and assures a concrete temperature of less than 200°F. As shown in Fig. 5.5, the mounting columns for the components are fitted with double bellows seals where they penetrate the thermal shield. It may also be noted that the columns are supported on shock mounts to provide the requi- site protection against seismic disturbances. The walls of the reactor cell present the most serious design problem. They are exposed to gamma flux heating and to an ambient temperature as high as 1150°F inside the cell, they must be leak-tight to maintain containment integrity, and radiation. 2 67 ORNL-DWG 57-10639A SUPPORT FOR REACTOR STORAGE \ / STEAM CELL i) L& { - k) . v, 2 / - : 5 - s e . . . . . 7 3 IS - ' - . " e e - - - - r R i : i L Ct| e Y ’fl g S ] y ": ‘\‘\\ 95:-'1 3 ' ! ’:_‘/" ':’-,' AR et | : VO« L) T R = - L > S | LY A / 4 . - CNEEE ST : - Ry E Fezrdl A - ey T RN S . STEAN PIPING A e oo L - ? - ) ; o fi)'_ . g SO g s, v X e : | @’flw % A OFF -GAS 2 X ey \ PROCESS ROCM oy SYEYEHEY el 1 L Tt e O / A i e W\,__L” REACTOR CELL / STEAM CELL N CUMP AND STORAGE TANKS CELL Fig. 5.3. Reactor and Steam Generator Cells ~ Plan. 250 Mw (electrical) module. ORNL~-DW(G 67-10645A0 FUEL AND BLANKET PUMP DRIVE MOTORS —r N STEAM L COOLANT SALT PUMP REACTOR GEMERATORS ST TN O 4 g STEAM /e ING - REHEATERS 62 Tt 2in. [0 1) .:-5)}(2—3&1\: e iz REACTOR SUPPORT FQR STORAGE S RERIMARY HEAT EXCHANGER AND PUMP Fig. 5.4. Reactor and Steam Generator Cells — Elevation. 250 Mw (electrical) module, section A-A. { —55ft gin.———- ISOLATOR — T REACTOR | I VIBRATION p 68 ORNL-DWG 67-1064848 — - B&ftOin—- - . R — PRIMARY HEAT EXCHANGER (\fl ! | J (//l L\\W BLANKET - | : HEAT EXCHANGER | l / / Fiy. 5.5. Reactor Cell Thermal Shield ond Component Supports. they must also provide the necessary biological shielding for the reactor. As shown in Figs. 5.6 and 5.7, the reactor cell biological shielding consists of 7-ft-thick rein- forced ordinary concrete around all sides below grade and an 8-ft thickness above grade. The top consists of removable concrete roof plugs having a total thickness of 8 ft. The method of construc- tion would be to first pour the 2-ft-thick reinforced floot pad. A 3-in.-thick carbon steel floor plate would be laid over this, and the 3-in.-steel vertical wall plates would be welded to it, the latter serv- ing as forms for pouring the side walls. A second 3-in.-thick steel plate would be erected inside the COOLING CHANNEL 7‘ ----- HEAT REFLECTOR — DOUBLE CONTAINMENT 69 ORNL~-DWG 67-40644A L R e i S THERMALSHELD/// 0O 24 6 810 Ll L L] INCHES Fig. 5.6. Reactor Cell Construction first with a 3-in. air gap between the two. A sim- ilar second plate and air gap would be provided for the floor. During reactor operation, cooling air would be circulated through the gap at a ve- locity of about 50 fps to remove the heat penet- ated by gamma absorptions in the wall. Cooling air is also used in the removable roof plugs, the air duct connections being flanged to facilitate removal. The total heat losses from the reactor cell are estimated to be a maximum of 800,000 VIBRATION ISOLATOR — Component Support Penetration. Btu/hr. The thermal shield could tolerate loss of cooling air for up to 1 hr without an excessive temperature rise. A 3/16--in.—thick carbon steel membrane is in- stalled inside the inner 3-in.-thick steel wall mentioned above. This membrane is hermetically tight and would satisfy the containment leak-rate requirements. The membrane is continuous around the top plugs and also at the penetrations of the cell, as shown in Fig. 5.7. The space between it EQUIPMENT SUPPORT ORNL-DWG 67-106374 COOLING SUPPLY ,,,,,, oo = REMOVABL.E . ROCF SLABS COOLING CHANNEL" N DOUBLE CONTAINMENT tjaqw-CELL HEATER THIMBLE 0 2 4 6 810 i INCHES Fig. 5.7. Reactor Cell Construction — Maintenance and Heater Access Details. and the 3-in. plate is exhausted by a vacuum pump which discharges into the cell atmosphere; this “pumped-back’’ system thus provides opportunity for continuous monitoring of the double contain- ment system. The inside surface of the sealing membrane is covered with a 6-in.-thick blanket-type thermal insulation protected on the inside by a 1/1 cin. skin of stainless steel. This skin also serves as a radiant heat reflector to reduce heat flow into the wall, The fuel, blanket, and coolant salts will be maintained above their liquidus temperatures by operating the interior of the cells at temperatures up to 1150°F. During normal operation the tem- perature is self-sustaining, and, as mentioned above, the problem is one of heat removal from the wall. For warmup and low-power operation, 4 however, electric heaters are provided in thimbles around the inner walls of the cells, as shown in Fig. 5.7. The electric leads for the heaters are brought out through sealed bushings in the thimble caps. The off-gas cells and the drain-tank cells have double containment but do not need the thermal shield to protect against the radiation flux. The steam cells require only the thermal insulation and cooling air, since the radiation levels will be relatively low and double containment is not re- quired in these spaces. 5.3 REACTOR G. H. Llewellyn W. C. George W. C. Stoddaxt W. Terry H. L. Watts H. M. Poly During the past teport period, the new data dis- cussed in Sect. 6.1 became available on the di- mensional changes that occur in graphite as a result of neutron irradiation. Because of this experimental evidence, we decided to redesign the reactor even though there is optimism that a more stable graphite will be developed within the next few years. The MSBR cost and performance characteristics continue to be attractive even though penalized by designing on the basis of the immediate technology. We also decided that - the reactor should be designed in such a way that major redesipgn or modification would not be re- quired, to take advantage of a more stable graphite when it becomes available. Several new approaches were tried for the core design, one of which was to put the fertile salt in the flow passages through the core graphite and to allow the fuel salt to move through the in- 71 terstices. This so-called “‘inside-out’’ design could probably accommodate the dimensional changes in the graphite, assuming that suitable adjustments were also made in the fuel enrich- ment. A major disadvantage, however, is that the fuel salt would also penetrate into the interstices of the radial blanket, a position in which it is ex- posed to relatively low neutron flux and thus pro- duces relatively litile power. Since the flow in this area would also be somewhat indeterminant, this design of the reactor was not pursued further. Attempts to design a removable graphite core for the reactor led to the conclusion that such an arrangement would probably be impractical. One major problem would be containment of the highly radioactive fission products associated with re- moval of a bare reactor core. There would also be the problem of assuring leak-tightness of a large-diameter flanged opening which must be sealed only by use of remotely operated tooling. As previously mentioned, it wasg decided to re- place the entire reactor vessel. Selection of a ten-year life for the reactor, or about 5 x 1027 nvt (greater than 50 kev) total max- imum neutron dose for any point in the core, meant that the power density would be reduced from the 40 kw/liter used in previous concepts to 20 kw/litex. This involved doubling the core volume from 503 ft® to about 1040 ft3, and also required that the reactor vessel size be increased cor- respondingly. The factors entering into selection of these conditions are given in Table 5.1, which shows the effect of power density on the perform- ance factors for the plant. At 20 kw/liter, it may be noted that the fuel cycle cost is 0.5 mill/kwhr and the yield is 4%/year. This appears to be the most practical design point, although it is without benefit of improved graphite. Tt should be pointed out that the differences in capital costs shown in Table 5.1. Performance Factors of MSBR as Function of Average Core Fower Density Fisaile Inventory Power Densitly Core Size {ft) Yield Fuel Cycle Cost Capital Cost Lide [ j et et et o : . ki for 1000 Mw {kw/liter) . . (7 /year) (mills /kwhr) [$/kw (electrlcal)] (years) T ) Diameter Height : (electrical)l 80 6.3 8 5.6 0.44 117 2.5 880 40 8 10 5.0 0.46 119 5 1040 20 10 13.2 4.1 0.52 125 i0 1260 16 12 i8 2.7 0.62 132 20 1650 72 OHRNL-DWG 67-10643A - e : 14§t 0in. 0D - - 10 ft O in. DIAM CORE et FUEL CELLS 420 AT 5% in. PITCH 53 in. HEX x 13t 3in. LONG BUANKET CELLS 252 AT 53 DIAM (~15 in. TOTAL ) REFLECTOR GRAPHITE {~6in. TOTAL) COOLING SKIRT BLANKET PLENUM PRESSURE VESSEL CONTROL ROD HOLD-DOWN PLATE FUEL SPACER SHEETS BLANKET CELL SPACER TIES Fig. 5.8. 250 Mw (Electrical) Reactor — Plan. 420 SS;é-in. cells. Table 5.1 do not take into account the cost of re- placing the graphite moderator at intervals deter- mined by radiation damage to the graphite. The frequency of replacement and the contribution to the power cost are greater at the higher power densities, and this effect essentially offsets the difference in initial cost. Plan and elevation views of the reactor are shown in Figs. 5.8 and 5.9. The design param- eters are given in Table 5.2, There are 420 two-pass fuel flow channels in the reactor core graphite. As shown in Fig. 5.10, each of these channels is formed from two graphite extrusions. The outer, and longer, of the pieces is hexagonally shaped, 53/3 in. across the flats, and has a 2 23/3 ,-in.-diam hole in the center. The total length of the extrusion is about 14% ft. The inner piece is a cylinder, 2% in. OD x 1% in. 1D, and fits inside the hole of the hexagonal extrusion. The fuel salt enters the reactor through the inlet plenum at the bottom of the core and flows upward through the annulus between the inner and outer graphite extrusions to the top of the reactor, where it turns and flows through six - by 17%-in. slots 73 DRNL-DWG 67-10538A - e {4 Oin DIAM (D_ /CONTROL ROD TS Bt e - s e T A e ,r»-‘c’:j’.j_’jf - / Lo - e TRy | A7 L CORE 10 [t Oin. DIAM x 43 3 in. HIGH \*qu‘ HOLE-DOW, 1 / 420 FUEL CELLS AT 5¥8 HEX N’ - TO PU‘TE\H S : . BLANKET HEAT | — EXCHANGER -~ ]_._ - oy / T 3 19 ft Bin, REACIOR : ’ | ALIOR ‘ 13t 2in. VESSEL - enm| ‘ | CORE D | | i 1 REFLECTOR ke GRAPHITE :}\ niE R - ] BLANKET %< CELLS — {r»_;. 2 ft 2in. FUEL SALT PLENUMS—" -~ SUPPQRT LEGS {8) Fig. 5.9. 250 Mw (Electrical) Reactor — Elevation, 420 5?78-in_ cells. 3in. 3in. DIAM e 3Y5in. DIAM 1ft 3in. e T, e ———— o 13 £ 3in. REACTOR ! 5.3750n. 5 223/321&![) ‘fi \ N 10in. [ 3in. MIN Fig. 5.10. Fue! Cell Eiement. ORNL-DWG 87-10646A Y5 in. 1IDx 2% in. 0D SiX Yo —x1¥-in. SLOTS 2 ¥in. DIAM 1/2“’1. GRAPHITE TO METAL BRAZE 17gin. ODx 43/4in. 1D 1Y in. 0DxtYgin 1D FUEL INLET PLENUM FUEL QUTILLET PLENUM - r . Tahle 5.2. Reactor Specifications Average core power density, kw/liter Power, Mw Number required for 1000 Mw (electrical) Vessel diameter, ft Vessel height, ft Core diameter, ft Core height, ft Core volume, £t 3 Fraction of fuel in core Fraction of blanket in core FFraction of graphite in core Blanket thickness, ft Fraction of salt in blanket Breeding ratio Fuel yield, %/year Fuel cycle cost, mills /kwhr Fissile inventory, kg Fertile inventory, kg Specific power, Mw (thermal)/kg Number of core zlements Velocity of fuel in core, fps Average flux >(0.82 Mev Fuel Volume,‘ft3 Reactor core Plenums Entrance nipples Heat exchangers and piping Processing Total Peak/average flux ratio 20 556 4 14 22 10 13.2 1040 0.134 0.064 0.802 1.25 0.58 1.06 4.1 0.52 314 54,000 1.8 420 4.8 3.33 % 1013 135 37 13 160 6 355 ™2 to the inside of the cylindrical graphite extrusion and returns to the bottom of the reactor and the outlet plenum. The average velocity of the fuel in the core is about 4.8 fps as it is heated from 1000°F to 1300°F. The effective height of the core is approximately 137 ft, and the total length of the two-pass flow channel, plenum to plenum, is about 27 ft. The hexagonal graphite pieces have a cylindrical portion about 12 in. long turned at the bottom to which a Hastelloy N nipple, 17/8 in. OD by 1/16 in. in wall thickness, is brazed. The other ends of these nipples are welded to discharge openings in the upper plate of the inlet fuel plenum. Inner Hastelloy N nipples, 1Y a4 in. OD by ? in. in wall thickness, are welded to the outlet plenum tube sheet and have an enlarged upper end which fits snugly, but is not brazed, into the inner 1 1/2—in.—ID hole in the inner graphite cylinder of the fuel ele- ment. The lower end of the core assembly is thus fixed in place by attachment to the plenums, but the upper end is free to expand or contract in the vertical direction. The graphite fuel pieces extend 15 in. above the end of the fuel flow passages in order to serve as the top axial reflector for the core. The top 3 in. of each fuel element is turned to a smaller diam- eter to establish a shoulder, as shown in Fig. 5.10. Triangular stampings of Hastelloy N sheet are slipped down to this shoulder and engage three of the elements, as shown in Fig. 5.8. These stampings are interleaved to maintain the radial spacing of the fuel channels, yvet eliminate the need for a large-diameter upper diaphragm drilled to close tolerances. The center six fuel channels engage a ring which is attached through six ribs to the vessel itself, thus stabilizing the entire assembly. Immediately outside the core region of the reactor are graphite tubes around and through which the fertile salt of the radial blanket is circulated. These tubes displace the more expensive fertile salt and also, by scattering the neutrons, promote more effective capture by the thorium atoms in the blanket. The ratio of fertile fraction to graphite is about 58% in this region, as determined by the code used for optimization of the reactor design. The graphite tubes are slipped over short nipples extending from a mounting plate at the bottom of the reactor, as shown in Fig. 5.9. The tubes are radially positioned at the top by overlapping con- nectors in much the same manner as the fuel ele- ments, Solid cylinders of graphite, 5 in. in diameter, are arranged on the outer circumference of the re- actor to serve as a reflector. A can of 1/‘#-in. wall thickness surrounds the reflector graphite and serves to direct the entering fertile salt down the inside wall of the vessel and to the bottom of the core. The fertile salt stream then divides; part of it moves upward through the interstices between the fuel elements, while the major portion flows through the graphite tubes in the blanket region. It may be noted that the fuel channels them- selves provide sufficient graphite for moderation of the reactor and for the top reflector without use of any special shapes or pieces, as was required in earlier MSBR concepts. All the graphite con- sists of extrusions which require little in the way of machining or close tolerances. These design improvements were made possible by relaxing the restriction that the wall be thin enough to permit reduction of the permeability to the 10~ 7-cm?/sec range by impregnation. This requirement was eliminated because, as discussed in Part 5, there appears to be more hope for obtaining very low permeability through a surface sealing technique, which would not be limited by wall thickness, than through impregnation, which requires thin sections. The fuel enters and leaves the reactor through separate pipes rather than by the concentric pipe arrangement employed in the previous concepts. The new design eliminates the need for the inner slip joint and also relieves the thermal stress problems that existed in certain temperature ranges. The thermal expansion loops now shown in the salt piping add to the fuel salt volume in the sys- tem, but this is of less significance thau it would have been in the 40-w/cm? reactor, which had only one-half the core volume. 5.4 FUEL HEAT EXCHANGER T. W. Pickel W. Terry The heat exchanger for transfering heat from the fuel salt to the coolant salt remains essentially unchanged since the last report. The exchanger is shown in Fig. 5.11, and the design data are given in Table 5.3. As mentioned previously, there has been some change in the salt headers. Design of the gas sparging system has not been completed. 5.5 BLANKET HEAT EXCHANGER T. W. Pickel W. Terry No major changes have been made in the heat exchanger which transfers heat from the fertile, or blanket, salt to the coolant salt. The exchanger is shown in Fig. 5.12, and the pertinent data are listed in Table 5.4. The blanket heat exchanger has been lowered in the cell so that the pump drive shaft will be the same length as the fuel pump drive shaft and thus allow interchangeability of parts. The coolant salt piping has also been modified to replace the concentric piping formerly used. 76 Table 5.3. Fuel Heat Exchanger Specifications Number required per reactor module Rate of heat transfer, Mw Rate of heat transfer, Btu/hr Shell side Hot fluid or cold fluid Entrance temperature, °F Exit temperature, °F Entrance pressure, psi Ex1it pressure, psi AP across exchanger, psi Mass flow rate, 1b/hr Tube side Hot fluid or cold fluid Entrance temperature, °F Exit temperature, Oy Entrance pressure, psi Exit pressure, psi AP across exchanger, psi Mass flow rate, lb/hr Tube material Tube OD, in. Tube thickness, in. Tube tube sheet, ft Inner annulus length, tube sheet to Outer annulus Shell material Shell thickness, in. Shell ID, in. Tube sheet material Tube shect thickness, in. Top outer annulus Top inner annulus Floating head Number of tubes Inner annulus Outer annulus Ditch of inner ananulus tubes, in. Radial Circumferential Pitch of outer annulus tubes, in. Type of baffle Number of baffles Inner annulus Outer annulus 1 520 1.80 x 10° Cold (coolant salt) 850 1110 198 164 34 1.68 x 107 Hot (fuel salt) 1300 1000 146 50 96 1.09 % 107 Hastelloy N 0.375 0.035 15.3 16.7 Hastelloy N 1 67 Hastelloy N 1.5 2.5 _ 3.5 ] 4347 3794 0.600 0.673 0.625, triangular Doughnut CRNL:-DWG 67 -10642A 77 = — & i Il = = > = L8k w - & ~ = Z 5 o w8 o o W 3 < 4 7z = a4 m G C @ = = £ = 5 O n - . = = = 5 c »ng ..Am - O o = M = 1 M ] = o= = . - _ =z o o D ox g © b o 9L @Ff o W 2 = C Weoow O » m 2 3 o ) 7 £ 2 & 4 29z o4 H o= : = o = & = 8 6w b =z : r < < < 2 b C == L O bl ooy Lo ~ = w = — r I 2 - oo il = Ll = o n © ~ I JoLw Z ¥ < om £« o _ f ' : i : < | ' : | W 1 ,,‘hmi\ll, . " : i e i _ i : ! i[\\ifl(}.n fi ! : ! N i " 0. : Z _ < ¢ = e 35 o ./!l - e / m\.\»w/ : ““““ =TT Vi = L ! ) e 7 a / ” fl Kl ooy .,‘T " ..«\VM. .fl = W m_ : . 7 , b i : / A N | L W v 0 / P ) , al / Lo W\ Z| T //nfis = < g 4 - : : & A ; : (& i : [ j : o : : % ” : & : ] . fi = | _ N < : = = s £ L ; - ” 0 : 5 ) ; = = e ! * — - —— —r — - - ./._| —— fi +— - - - | e : - - - = = | o ” b - 0 o - - 23 1 7Yin ] 29ft 3'%in DRAIN LINE 2-in. Fuel Heat Exchanger and Pump. 250 Mw (electrical) unit. Fig. 5.11, 30ft 6¥4in. | | GAS CONN | i | | | | | 19ft Qin. GAS SKIRT- 4t 8in. OD | COOLANT 20in. NPS Fig. 5.12. Blonket Heat Exchanger and Pump. 250 Mw (electrical) \ | 4f+ Qin. 8in. NPS ' } i | 2ftCin. 14§t 1%in. | 211 0in. D|f-\M--n~- . ] 78 N R . o ""I T — _ . ' 2Tt 9% in. ‘:;1 i I A Lo 0l : I i ] ! 1“\; it - |\ [ . H \ ] il B : { r+fl;1‘ w al 0L i ...... L - ; S ! o e T g oS e s e R by V(7 rain NeS L ! \ S . i ORNL-DWG 67-406294A - MOTOR FLANGE SPCOL CPERATING LEVEL MOLTEN SALT BEARING SLANKET PROCESSING -DRAIN LINE BLANKET TO REACTOR 8in. NPS FLANGE IMPELLER 8ft 4in. -- QUTER TUBES 822 AT ¥gin. OD INNER TUBES 834 AT ¥gin. OD unit. Table 5.4. Blanket Heat Exchanger Specifications Number regquired Rate of heat transfer, Mw Rate of heat transfer, Btu/hr Shell side Hot fluid or cold fluid Entrance temperature, °F Exit temperature, °F Entrance pressure,® psi Exit pressure,? psi AP across exchanger,b psi Mass flow rate, ib/hr Tuhbe =side Hot fluid or cold fluid Entrance temperature, °F Exit temperature, °p Entrance pressure,? psi Exit pressure,?® psi AP across exchanger,b psi Mass flow rate, lb/hr Velocity, fps Tube material Tube OD, in. Tube thickness, in. Tube length, tube sheet to tube sheet, ft Shell material Shell thickness, in. Shell ID, in. Tube sheet material Tube sheet thickness, in. Number of tubes Inner annulus Outer annulus J Pitch of tubes, in. Total heat transfer area, £t ? Basis for area calculation Type of baffle Number of baffles Baffle spacing, in. Disk OD, in. Doughnut ID, in. Cverall heat transfer coef- ficient, U, Btu el a2 4 27.8 9.47 % 107 Cold (coolant salt) 1110 1125 138 129 15 1.68 x 107 Hot (blanket salt) 1250 1150 111 20 91 4.3 % 10° 10.5 Hastelloy N 0.375 0.035 8.3 Hastelloy N 0.50 55 Hastelloy N 1 834 822 0.8125, triangular 1318 Tube OD Disk and doughnut 4 19.8 33.0 31.8 1030 “Includes pressure due to gravity head. PPressure loss due to friction only. 79 5.6 FUEL DRAIN TANKS H. L. Watts T. W. Pickel Two 60-in.-diam by 20-ft-high dump tanks are required for draining the fuel salt from each of the reactor systems. These are shown in Fig. 5.13. One drain tank is connected to the 5-in.-diam overflow connection on the sump tank of the fuel circulating pump and receives all the salt from the reactor when it is drained. The drain time is estimated to be short, probably less than 10 sec, and this method could be used for emergency shutdown if necessary. The other drain tank is connected to the fuel heat exchanger and fills through a siphon action due to the differences in elevation. The maximum heat to be removed in each fuel drain tank is 12 Mw (thermal). The two tanks are cooled by a total of 422,000 1b/hr of steam taken from the high-pressure turbine-generator exhaust at about 550 psia. The steam is heated from 650°F to about 1000°F in vertical thimbles spaced on 3- in. centers in each dump tank. There are 271 of these thimbles per tank, extending to within 1 in. of the bottom. The reentrant-type steam thimbles are inserted in 2-in.-OD INOR-8 thimbles which contact the fuel salt, the 0.025-in. radial clearance between the two thimbles being filled with stag- nant inert fuel salt to setve as a heat transfer medium. The double thimble arrangement provides the necessary double barrier between the steam and the enriched fuel salt. Steam leaves the thimbles at about 72 fps and enters the steam chest at the top of the dump tank. 5.7 BLANKET AND COOLANT SALT DRAIN TANKS H. L. Watts W. C. George The drain tanks for the blanket salt and for the secondary coolant salt are essentially simple storage reservoirs. Cooling systems such as those used in the fuel salt tanks are not required. The details of the designs for these tanks have not been completed. The location of the tanks in the cells is shown in Fig. 5.3. 80 ORNL-DWG 67 - 10644 A .- STEAM {*8-in. PIPE) e 650 °F - 570 psig 271 TUBES 2% A PITCH 4 . -11/2 in. i 44 DISH R TYP ALL HEADS 2 in L ; | ’ ! 1 41 in. _BUFFER "] . GAS 0072 in o _FUEL A15-in, PIPE) ‘ o ey W 24 f1-9'% in. | ‘ Ff_fij,..cvj AN ] TTUPLENUM \}\ ‘z; R T RISER SKIRT N . ~~ HEAT SHIELD 19 -6 - TUBE 6in SUPPORT ¢.042 n. - = - --jo.O?z in A1 'y . , _ e | \5(_’_ | S R - 1 = i | e /, b= / ) ! N A TO FREEZE kfi_fl/ Y | A VALVE [t b \ ; T~ FUEL DRAIN | ; 2 in. 0D x Q035 in. L 1 % in. OD x 0.048in. 1'% 1. OD x 0.025in Fig. 5.13. Fue! Drain Tank with Decay-Heat Removal System Details. 81 58 STEAM GENERATOR-SUPERHEATER The initial concepts have changed little, however, AND REHEATER and the design work has consisted primarily of detailing tube sheets, supports, headers, etc. C. E. Bettis T. W. Pickel Details of this equipment will be covered in sub- Some work was continued on the design of the sequent reports, steam generator-superheater and on the reheater. 6. Reactor Physics A. M. Perry 6.1 MSBR PHYSICS ANALYSIS O. L. Smith Optimization of Reactor Parameters H. T. Kerr The two principal indices by which the perfor- mance of the Molten-Salt Breeder Reactor is cus- tomarily evaluated are the cost of power and the annual fuel yield, that is, the annual fractional in- crease in the inventory of fissionable material. These are used as figures of merit in assessing the influence of various design parameters or the effect of design changes that may be contemplated, and, in fact, we customarily combine them into a com- posite figure of merit, F-y+100(C 1+ X)77, in which y is the annual fuel yield in percent per year, C is that part of the power cost which de- pends on any of the parameters considered, and X is an adjustable constant, having no physical sig- nificance, whose value merely determines the rela- tive sensitivity of F to y and C. Since a large num- ber of reactor parameters are usually involved, we make use of an automatic search procedure, carried out on an IBM 7090 computer, which finds that com- bination of the variable design parameters that max- imizes the figure of merit, F, subject to whatever constraints may be imposed by the fixed values of other design parameters not allowed to vary. This procedure, called OPTIMERC, incorporates a mul- tigroup diffusion calculation (synthesizing a two- space-dimensional description of the flux by al- ternating one-dimensional flux calculations), a determination of the fissile, fertile, and fission product concentrations consistent with the proc- 82 essing rates of the fuel and fertile salt streams, and a method of steepest gradients for optimizing - the values of the variabies. By choosing different values for the constant X in the figure of merit, F, we can generate a curve showing the minimum cost associated with any attainable value of the fuel yield, and by carrying out the optimization proce- dure for different, successive fixed values of se- lected design parameters, we can generate families of such curves of C vs y. (In OPTIMERC any of some 20 parameters may either be assigned fixed values or be allowed to vary within specified limits subject to the optimization procedure.)} One of the design parameters which has a sig- nificant influence on both yield and power cost is the power density in the core. (Actually, the core dimensions for a given total power are the param- eters used.) The performance of the reactor is better at high power densities. At the same time, the useful life of the graphite moderator, which is dependent on the total exposure to fast neutrons, is inversely proportional to the power density (see next section). It is necessary, therefore, to de- termine the effect of power density on performance with considerable care. In Fig. 6.1 the fuel-cycle cost is used because it contains, in fact, most of the variation of power cost with the parameters being varied. It may be seen from Fig. 6.1 that a reduction in power density from 80 to 20 w/cm? involves a fuel-cycle cost penalty of about 0.1 mill/kwhr (electrical) and a reduction in annual fuel yield of perhaps 1.5%. There is, of course, also an increase in capital cost (cf. Chap. 5), but this is essentially offset by a reduction in the cost of replacing the graphite and reactor vessel at intervals determined by ra- diation damage to the graphite. The combined penalty for having to replace the graphite, com- pared with a high-power-density core not requiring replacement, is about 0.2 mill/kwhr (electrical), 1.00 o o [ 20ST [ milis/kwhr(eiec‘rricali] ~ u L CYCLE FUE 075 o 025 [~ 83 ORNL--DWG 67—448B06 0 2.5 50 whether this comprises higher capital cost plus lower teplacement cost at low power densities, or lower capital cost plus higher replacement cost at higher power densities. Figures 6.2 and 6.3 show the variation of other selected parameters both with power density and with the adjustable constant X. It is apparent from these results that the useful life of the graphite is not increased by reducing core power density without some sacrifice in other aspects of reactor performance. The reduction in yield and the increase in cost are quite modest for a reduction of power density from 80 to 40 w/cm?, but they become increasingly more significant for each further factor-of-two reduction in power den- sity. Nonetheless, it appears that with an average power density as low as 20 w/cm>, the MSBR can still achieve an annual fuel yield of 3.5 to 4% and a fuel-cycle cost of less than 0.5 mill/kwhr (elec- trical). Fig. 6.1. MSBR Fuel-Cycle Cost vs Annual Fuel ) 7.5 FUEL YIELD (% per year) Yield. ORNL-DWG 67141807 0.20 } I [ [ SALT VOLUME FRACTIONS 3 | i Q4B Joooe 042 2/ O 0.08 L 10m ke __‘_._EEETLLE 6 = T ee— e 0 21 TE R0 o~ 0 0.04 3 _ SPECIFIC POWER [Mw({thermal}/kg ] o 20 40 60 POWER DENSITY (w/cm3 ) /U RATIO (X107} 0 20 40 50 80 POWER DENSITY (w/cm3) Fig. 6.2. Variation of MSBR Parameters with Average Core Power Density. Numbers attached to curves are values of the adjustable constant X. ORNL-DWG 67-141808 RADIAL BLANKET THICKNESS THICKNESS (ft) 0.08 [ e e e NEUTRON I OSSES TO Li Bef e e o [10 0.06 4 40 0.08 | T r NEUTRON LOSSES TO MODERATOR 0.06 0.04 +— NEUTRONS ABSORBED /7 SCURCE NEUTRONS 0.02 2.25 2.24 w = 2.23 2.22 400 e b “. 200 TR e, ‘—J ERNAL | ™ 20 POWER DENSITY (w/cm3) 40 60 80 Fig. 6.3. Variation of MSBR Parameters with Average Core Power Density. Numbers attached to curves are values of the adjustable constant X. 84 Useful Life of Moderator Graphite A. M. Peny Information currently used in the MSR project re- garding the dependence of graphite dimensional changes on fast neutron dose is derived primarily from experiments carried out in the Dounreay Fast Reactor (DFR). A curve of volume change vs fast neutron dose for a nearly isotropic graphite at temperatures in the range 550 to 600°C is shown in Fig. 6.4, which is taken from the paper of Henson, Perks, and Sim- mons.! (The neutron dose in Fig. 6.4 is expressed R W. Henson, A. J. Perks, and J. H. W. Simmons, Lattice Paramefer and Dzmensronal Changes in Graphite Irradiated Between 300 and 1350°C, AERE-R5489, to be published in the proceedings of the Eighth Carbon Con- ference. ORNL 44 _fiv.,.....................”‘\‘........................,....... T s r ’ 400-440°C B 550-600°C 8 , [ DWG 6? 61809 + 10 0o {Ye) CHANGE ,. - — VOLUM & 3 neuhrons,/cm2 ) 0 1 EQUIVALENT PLUTO DOSE HO‘ZZ Fig. 6.4. Yolume Changes of Near-lsotropic Graphite Resuvlting from Meutron lrradiation. See text for dose in terms of MSBR flux. 85 ORNL OWG 67-11810 ’O Y - L et b jg’wgmrg)dg RAEG = wggm— [ eI dE ED el 2 .‘é S i = Hy,0 SPECTRUM < B 2=0p,0 specTRUM | LT L 1 L v T g,'f 3 = CARBON SPECTRUM // 2 4 = FAST SPECTRUM e / w3 A 7 L .V o ~23% ;/.// " l S / u A
    ‘ %r/“"f 2 1 A @ = et T2 S B | it - e S e e O b gl b o L—-= ® .«-'/ O |-~ T 4 10? 2 5 10° 2 5 10° EO (eV) Fig. 6.5. Fost Flux as ¢ Measure of Radiation Damaoge. in terms of an equivalent Pluto dose; the total DFR dose, that is, frfmqsuz, f) dE dt | 0 0 is 2.16 times the equivalent Pluto dose.) From an inspection of all the available data, we have con- cluded that a dose of about 2.5 x 10?2 neutrons/ cm? (equivalent Pluto dose) could be sustained without any significant deterioration of the physi- cal properties of the graphite, and this has been adopted as an allowable dose, pending further de- tailed consideration of mechanical design problems that might be associated with dimensional changes in the graphite. In order to interpret these experiments to obtain predictions of graphite damage vs time in the MSBR, it is necessary to take into account the dif- ference in neutron spectra in the two reactors. This, in turn, requires assumptions regarding the effectiveness of neuntrons of different energies for producing the observable effects with which one is concerned. At present the best approach avail- able is to base one’s estimates of neutron damage effectiveness on the theoretical calculations of graphite lattice displacements vs carbon recoil energy carried out by Thompson and Wright.? The Thompson and Wright ““damage function’’ is inte- grated over the distribution of carbon recoil en- ergies resulting from the scattering of a neutron of a given energy, and the result is then multiplied by the energy-dependent scattering cross section and integrated over the neutron spectrum in the re- actor. Tests of the model have been made by Thompson and Wright by calculating the rate of electrical resistivity change in graphite relative to the 5®Ni(n,p)°8Co reaction, in different reactor ’M. W. Thompson and S. B. Wright, J. Nucl. Mater. 16, 146.-54 (19635). 86 DRNL-OWG 67-{1814 NEUTRCON ENERGY (MeV} 5108 6 4 3 2 1.5 1.0 ; f \ [ [ [ 1 ’ I ! 1= H,0 SPECTRUM 2=D,0 SPECTRUM 3=CARBON SPECTRUM 4=FAST SPECTRUM ({50% Na, 50% U-METAL, 20% U, 80% ->°U) 1.0 S S ¢ (U} Larbitrary units) 0.8 0.6 G0 0.08 I ] ! \‘ f LETHARGY U Fig. 6.6. Meutron Flux Per Unit Lethargy vs Lethargy. Normalized for equal damage in graphite. spectra, and comparing it with experimental de- terminations of the same quantities. The results indicate that the model is useful at least for pre- dicting relative damage rates in different spectra. A useful simplification arises from the observa- tion that the damage per unit time is closely pro- portional to the total neutron flux above some energy £, where Eo has the same value for widely different reactor spectra. We have reconfirmed this observation, to our own satisfaction, by comparing the (calculated) damage per unit flux above energy E as a function of E for spectra appropriate to three different moderators (HQO, D,0, and C) and for a ““typical’’ fast reactor spectrum. The results plotted in FFig. 6.5, show that the flux above about 50 kev is a reliable indication of the relative dam- age rate in graphite for quite different spectra. 7 Figure 6.0 shows the spectra for which these re- sults were derived. The equivalence between MSBR and DFR experiments is simply found by equating the doses due to neutrons above 50 kev in the two reactors. We have not yet calculated the DFR spectrum explicitly, but we expect it to be similar to the ““fast reactor’’ spectrum of Fig. 6.6, in which 94% of the total flux lies above 50 kev. Since the damage flux in the MSBR is es- sentially proportional to the local power density, we postulate that the useful life of the graphite is governed by the maximum power density rather than - by the average, and thus depends on the degree of power flattening that can be achieved (see next - section). In the MSBR the average flux above 50 - kev is about 0.94 x 10'* neutrons cm™? sec™! at a power density of 20 w/cm?. From the DFR ir- radiations it has been concluded that a dose of 5.1 x 1022 nvt (> 50 kev) can be tolerated. The lifetime of the graphite is then easily computed; this useful life is shown in Table 6.1 for an as- sumed plant factor of 0.8 and for various com- binations of average power density and peak-to- average power-density ratio. It must be acknowledged that in applying the results of DFR experiments to the MSBR, there remain some uncertainties, including the pos- sibility of an appreciable dependence of the dam- age on the rate at which the dose is accumulated 4 Table 6.1. Useful Life of MSBR Graphite Average Power Density r /P ) Life max av (w/cm®) (years) 40 2.0 5.4 40 1.5 7.2 20 2.0 10.8 20 1.5 14.4 as well as on the total dose. The dose rate in the DFR was approximately ten times greater than that expected in the MSBR, and if thete is a significant dose-tate effect, the life of the graphite in an MSBR might be rather longer than shown in Table 6.1. Flux Flattening O. L. Smith H. T. Kerr Because the useful life of the graphite moderator in the MSBR depends on the maximum value of the damage flux rather than on its average value in the core, there is obviously an incentive to reduce the maximum-to-average flux ratio as much as possible, provided that this can be accomplished without se- rious penalty to other aspects of the reactor per- formance. In addition, there is an incentive to make the temperature rise in parallel fuel passages through the core as nearly uniform as possible, or at least to minimize the maximum deviation of fuel outlet temperature from the average value. Since the damage flux (in effect, the total neutron flux above 50 kev) is essentially proportional to the fission density per unit of core volume, the first incentive requires an attempt to flatten the power density per unit core volume throughout the core, that is, in both radial and axial directions in a cylindrical cote. Since the fuel moves through the core only in the axial direction, the second in- centive requires an attempt to flatten, in the radial direction, the power density per unit volume of fuel. Both objectives can be accomplished by maintaining a uniform volume fraction of fuel salt throughout the core and by flattening the power density distribution in both directions to the greatest extent possible. 87 The general approach taken to flattening the power distribution is the classical one of pro- viding a central core zone with k = 1, that is, one which is neither a net producer nor a net ab- sotber of neutrons, surrounded by a ““buckled”’ zone whose surplus neutron production just com- pensates for the neutron leakage through the core boundary. Since the fuel salt volume fraction is to be kept uniform throughout the core, and since the concentrations of both the fuel and the fertile salt streams are uniform throughont their respec- tive circuits, the principal remaining parameter that can he varied with position in the core to achieve the desired effects is the fertile salt volume fraction. The problem then reduces to finding the value of the fertile salt volume frac- tion that gives k, = 1 for the central, flattened zone, with fixed values of the other parameters, and finding the volume fraction of the fertile salt in the buckled zone that makes the reactor critical for different sizes of the flattened zone. As the fraction of the core volume occupied by the flat- tened zone is increased, the fertile salt fraction in the buckled zone must be decreased, and the peak-to-average power density ratio decreases toward unity. The largest flattened zone and the smallest power density ratio are achieved when the fertile material is removed entirely from the outer core zone. Increasing the fuel salt concen- tration or its volume fraction (with an appropriate adjustment of the fertile salt volume fraction in the flattened zone) would permit a still larger flattened zone and smaller Pmax/Pavg, but would be ex- pected to compromise the reactor performance by increasing the fuel inventory, at least if carried too far, There are obviously many possible combina- tions of parameters to consider, It is not a priori obvious, for example, whether the flattened zone should have the same height-to-diameter ratio as the entire core, or whether the axial buckled zones should have the same composition as the radial buckled zone. While we have by no means carried the investigation of these questions as far as we need to, we have gone far enough to recognize several important aspects of the problem. First, by flattening the power to various degrees in the radial direction only and performing fuel- cycle and economic calculations for each of these cases, we find that the tadial power distribution can be markedly flattened with very little effect either on fuel cost or on annual fuel yield, our chief indicators of performance. That is, the radial peak-to-average power density ratio, which is about 2.0 for the uniform core (which is sur- rounded, of course, by a heavily absorbing blanket region and hence behaves essentially as if it were unreflected), can be reduced to 1.25 or less with changes in fuel cost and yield of less than 0.02 mill/kwhr (electrical) and 0.2% per year respec- tively. The enhanced neutron leakage from the core, which results from the power flattening, is taken up by the fertile blanket and does not rep- resent a loss in breeding performance. Second, attempts at power flattening in two dimensions have shown that the power distribu- tion is very sensitive to details of composition and placement of the flattened zone. Small dif- ferences in upper and lower blanket composition, which are of no consequence in the case of the uniform core, produce pronounced axial asymmetry of the power distribution if too much axial flat- tening 1s attempted. radial buckled zones may interact through the In addition, the axial and flattened zone, to some extent, giving a distri- bution that is concave upward in one direction and concave downward in the other, even though the integrated neutron current over the entire boundary of the central zone vanishes. In view of these tendencies, it may be anticipated that a flattened power distribution would be difficult to maintain if graphite dimensional changes, result- ing from exposure to fast neutrons, were allowed to influence the salt volume fractions very strongly. Consequently, some revisions in the details of the core design are under consideration as a means of reducing the sensitivity of the power distribution to graphite dimensional changes. Temperature Coefficients of Reactivity O. L. Smith C. O. Thomas In analyzing power transients in the Molten-Salt Breeder Reactor -~ as indeed for most reactors — one must be able to determine the reactivity ef- fects of temperature changes in the individual com- poneints of the core, for example, the fuel salt, the fertile salt, and the graphite moderator. Since the fuel is also the coolant, and since only small frac- tions of the total heat are generated in the fertile salt and in the moderator, one expects very much smaller temperature changes in the latter compo- 83 ORNL-DOWG 67-11812 8 e ——— . MODERATOR A e i x i & | (o) (b) -—q | Ao 8 | FUEL SALT OVERALL 4 AN f [ - Q X ol N L N _ x 60 - 74 L — u — _ (¢) (¢) ! - 8 e L e s — 800 300 1000 800 300 1000 7 {°K) 7 (°K) Fig. 6.7. MSBR Multiplication Factor vs Temperature. nents than in the fuel during a power transient. Expansion of the fuel salt, which removes fuel from the active core, is thus expected to be the principal inherent mechanism for compensating any reactivity additions to the MSBR. We have accordingly calculated the magnitudes of the temperature coefficients of reactivity sep- arately for the fuel salt, the fertile salt, and the graphite over the range of temperatures from 800 to 1000°K. The results of these calculations are shiown in Fig. 6.7. In Fig. 6.7a we show a curve of change in mul- tiplication factor vs moderator temperature (with 6k arbitrarily set equal to zero at 900°K). Similar curves of &k vs temperature for fuel and fertile salts are shown in Figs. 6.7b and 6.7c, and the combined effects are shown in Fig. 6.7d. These curves are all nearly linear, the slopes being the ternperature coefficients of reactivity. The mag- nitudes of the coefficients at 900°K are shown in Table 6.2. The moderator coefficient comes almost entirely from changes in the spectrum-averaged cross sec- Table 6.2, Temperature Coefficients of Reactivity Coefficient 1 dk Component I (OK)——I k dT Moderator +1.66 X 1077 Fertile salt +2.05 x 1073 Fue! salt —~8.05 x 1077 Overall —~4.34 x 1077 tions. It is particularly worthy of note that the moderator coefficient appears to be quite insen- sitive to uncertaiaties in the energy dependence of the 233U cross sections in the energy range below 1 ev; that is, reasonable choices of cross 389 sections based on available experimental data yield essentially the same coefficient. The fertile-salt reactivity coefficient comprises a strong positive component due to salt expansion (and hence reduction in the number of fertile atoms per unit core volume) and an appreciable negative component due to temperature dependence of the effective resonance-absorption cross sections, so that the overall coefficient, though positive, is less than half as large as that due to salt ex- pansion alone. The fuel salt coefficient comes mainly from ex- pansion of the salt, which of course reduces the average density of fuel in the core. Even if all core components should undergo equal temperature changes, the fuel-salt coefficient dominates; and in transients in which the fuel temperature change is far larger than that of the other components, the fuel coefficient is even more controlling. 7. Systems and Components Development Dunlap Scott Work related to the Molten-Salt Breeder Reactor was initiated during this period. Studies were made of the problems related to the removal of the noble gases from the circulating salt to help identify the equipment and systems required to keep the '?3Xe poison fraction and the fission product afterheat to an acceptable level. Prepara- tions were begun for operation of a small out-of- pile loop in which a molten salt will be circulated through a graphite fuel cell and for operation of an isothermal MSRE-scale loop with sodium fluoro- borate. The major effort of the program at this time is to help establish the feasibility of improved concepts and to define problem areas. Since the production of suitable and reliable salt pumps is one of the longest lead-time items for molten-salt reactors, a major emphasis is being placed on this program, Some of the work related to problems of the MSBR but actually performed on the MSRE is discussed in the MSRE section of this report. The other work is described below. 7.1 NOBLE-GAS BEHAVIOR IN THE MSBR R. J. Ked! In the MSBR conceptual designs, the graphite in the reactor core is unclad and in intimate contact with fuel salt. Noble gases generated by fission and any gaseous compounds can diffuse from the salt into the porous structure of the graphite, where they will serve as heat sources and nuclear poisons. A steady-state analytical model was developed to compute the migration of noble gases to the 90 graphite and other sinks in the MSBR. The sink terms considered are: 1. Decay. 2. Burnup — takes place in the graphite and in the fuel salt passing through the core. 3. Migration to graphite — these gases ultimately decay or are burned up. 4. Migration to circulating bubbles -- these gases are stripped from the fuel loop to go to the off-gas system. Two source termms are considered: genertation directly from fission, which is assumed to occur only in the core, and generation from decay of the precursor, which occurs throughout the fuel loop. The model is based on conventional mass transfer concepts. Some degree of success has been ex- perienced with similar models developed for the MSRE, for example: 1. Xenon-135 poison fraction calculations. The steady-state model is developed in ORNI.-4069. Results of the time-dependent form of the model are summarized in ORNIL-TM-1796. 2. A model was developed to compute the con- centrations of daughters of very short-lived noble gases in graphite (ORNL-TM-1810). Computed concentrations check very well with measured values. With this model, steady-state '?5Xe poisoning calculations have been made for the MSBR [556 Mw (thermal) fueled with 233U and moderated with unclad graphite] to show the influence of various design parameters involved. The reactor con- sidered here is essentially that described in ORNL-3996 [P. R. Kasten e/ al., Design Studies of 1000-Mw (e) Molten-Salt Breeder Reactors], with specific design parameters as listed in Table 7.1. re 91 Table 7.1. Design Parameters Reactor power, Mw (thermal) 556 Fuel 233y Fuel salt flow rate, ft3/-sec 25.0 Core diameter, ft 2 Core height, ft 10 Volume fuel salt in core, £t3 83 Volume fuel salt in heat exchanger, £t 2 83 Volume fuel salt in piping between core and heat exchanger, £t3 64 Fuel cell cross section DOWNFLOW CHANNEL 3—7/3 in. HOLES UPFLOW CHANNEL Total graphite surface area exposed to salt, £t? 3630 Mass transfer coefficient to graphite — upflow, ft/hr 0.72 Mass transfer coefficient to graphite — downstream, ft/hr 0.66 Mean thermal flux, neutrons see ¥ em ™2 5.0 x 1014 Mean fast fiux, neutrons sec_1 cm ™2 7.6 x 1014 Thermal neutron cross section for 233U, bhams 253 Fast neutron cross section for 233U, bams 36.5 Total core volume — graphite and salt, it 503 233U concentration in core — homogenized, atoms barn ~ 1 em ™! 1.11 x 1073 Graphite void available to xenon, % 10 135%¢ parameters Decay constant, 1/hr 7.53 % 1072 Generation direct from fission, % 0,32° Generation from iodine decay, % 6.38° Cross section for MSBR neutron spectrum, bams 9.94 x 16° “The values for the yield of 135 Xe from the fission of 2330 are from an old source and were used in the screening calculations. Recent values of 1.11% for generation direct from fission and 6.16% for generation from iodine decay as reported by C. B. Bigham et al. in Trans. Am. Nucl. Soc. 8(1), June 1665, will be used in the future. The xenon stripping mechanism consists in cir- culating helium bubbles with the fuel salt and then removing them from the system. These bubbles are injected at the inlet to the heat exchanger in the region of the pump. Xenon-135 migrates to the bubbles by conventional mass transfer, and the mass transfer coefficient controls the rate of migration, The circulating bubbles are then stripped from the salt by a pipeline gas separator at the heat exchanger outlet. The heat exchanger, then, is the xenon stripper region of the fuel loop. The principal parameters to be discussed here will be: 1. Diffusion coefficient of xenon in graphite. 2. Parameters associated with circulating bubbles. (2) Mass transfer coefficient to the bubbles. (b) The surface area of the bubbles. 3. The surface area of graphite exposed to salt in the core. In the plots that follow, the diffusion coefficient of xenon in graphite at 1200°F with units of ft2/hr 92 ORNL-DOWG 6711813 G e : e - : i L ‘ L = L — o o= - - i - = z O 8 8 < g Q g5 x z2 o S 5 S o & 7 L _ v E Ll & - W | o Wl a - Wl o o € g E° of b E 6 — s — . _ B — 0 62 £ %z ou ; wn T Lj @ ” wl v B m a e , . e 2 W @ 3 [ ‘ < = - 0 NO CIRCUL ATING BUBBLES D T 0w c o C:D) w :LD S - : S - = @ SO a 2l 4 L _ 4 . £ &5 33 ' n et [ 2 (:)) < L) < o L& x = . s L C 5 - L 2o g - o S o = L S 2 L. O s o L le 82 o%Z | oIy, <, » WA w7 z @= T = i Z aZ 5 ot Ll o W o —— e = & 8 O L|1_J ol e . . T%““"““'- LW E o E =z n oo r S < 20 20 b e e = " O [ ] — 1.0 3000 0 VL - 10 3000 300 e 20 3000 300 40 3000 300 0oL e e SO, _ 1072 1073 1074 105 10776 107 DIFFUSION COEFFICIENT OF Xe IN GRAPHITE AT 4200°F {Ha/hr) ~PERMEABILITY OF He IN GRAPHITE AT ROOM TEMPERATURE (cm2/sec) - Fig. 7.1. Effect of Diffusion Coefficient in Graphite on 135%¢ Poison Fraction. is used as a parameter. Numerically, this is ap- proximately equa!l to the permeability of helium in graphite at room temperature with units of cm?/sec, if Knudsen flow prevails. Knudsen flow should be the dominant flow character for perme- abilities less than 107° cm?/sec. For perme- abilities greater than 10~° cm?/sec, viscous flow becomes important and this direct relationship does not exist. The gas circulated through the system is handled in these calculations as two groups of bubbles. The first group, refetred to as the bubbles, is injected at the bubble generator and removed with 100% efficiency by the gas separator. once-through”’ “recirculated’’ The second group, referred to as the bubbles, is injected at the bubble generator, com- pletely bypasses the gas separator, and recircu- lates through the system until the bubbles are removed with 100% efficiency on their second pass through the gas separator. In the accompanying plots the bubble surface aiea is the quoted param- eter. For proper orientation, note that 3000 ft? of bubble surface area corresponds to an average void fraction of 1% in the stripper region of the fuel loop with bubbles 0.020 in. in diameter, and also cor- responds to a gas flow rate of about 40 scim. Figure 7.1 shows the '*%Xe poison fraction as a function of the diffusion coefficient in graphite. ORNL-DWG 6711814 9 ................ e —— e pameamsemeemmmsem————— % & B feees s s s s esesi A w — CIFFUSION COFFFICIENT OF Xe IN GRAPHITE=10"5 ffz/hr éuj O o= 7 —_— Z = Z g2 & 1o 2 L o &% ; e 6l - S — g EE o n 9 o 3 x Q4 o 0 . §‘ a. {'E - = Q: v w O o zu s ——— = T X © o~ e z ‘& ia B w - i z E D = o o wr - o 1Y m 2 3 _t D a > m m H- m = S L0 o z __‘ = }— o 2 iz o I o - :13 o .l wl = U3 o 8 i = o £ e L. L [ o 7 < /min PUMP SHAF T PURGE o TOTAL PRESSURE, 25 psig HEL I PUMP TANK { BF, PARTIAL PRESSURE, 18 psi T STAGK U FLOW CONTROL —FO0 / { 100 Cm3/m|n T , - Tfiéw A ™~ HELIUM AND BF BF g T - o ‘} 1456 em>/min 3 SUPPLY - SU FLOW CONTROL e — 1 e FILTER 700 cm¥/min T " NaBFg4 - 850 gpm 1200°F i VARIABLE ORIFICE ~ e . i _,.:—_;—:‘:_’_:i’,‘\\ j: T FLOW ELEMENT FREEZE VALVE . Fig. 7.4. Cover Gas Control System for the Fluoroborate Test. about 500 mm Hg at the test operating temperature (1200°F), it will be necessary to maintain an over- pressure of BF . in the pump bowl vapor space to avoid a loss of BE | from the salt with the resultant increase in salt liquidus temperature. The cover gas system is being revised to include the neces- sary equipment for handling and controlling the BF ,. This system, shown schematically in Fig. 7.4, includes some of the features of the cover gas system used with the MSRE coolant salt, and the information developed in the test will be useful in planning the revisions which will be necessary to prepare for testing of NaBF | in the MSRE coolant system. The operating conditions for the loop are: Temperature 1200°F Flow rate 800 gpm Pump head 120 ft Pump speed 1800 rpm The tendency of BF | to induce polymerization of organic materials could cause problems in the pump lubrication system and in the off-gas line, The helium purge to the shaft annulus will be adjusted to minimize diffusion of BF | into the pump bearing chamber, and filters are provided in the off-gas line to protect the pump tank pres- sure control valve., The BF | flow required will be dictated by the total pump bowl pressure, the desired BF3 partial pressure, and the required helium purge flow. It is planned to operate the loop isothermally for a period of about six months. The objective will be to uncover any problems associated with the circulation of NaBF , and to devise and test suitable solutions or corrective measures. 7.4 MSBR PUMPS A. G. Grindell P. ;. Smith L. V. Wilson Survey of Pump Experience Circulating Liquid Metals and Molten Salts A survey of experience with pumps for liquid metals and molten salt was made, and a report? Ip. . Smith, Expetrience with High-Temperature Centrifugal Pumps in Nuclear Reactors and Their Application to Molten-Salt Thermal Breeder Reaciors, ORNL.-TM-1993 (September 1967). 96 was issued relating pump descriptions, operating hours, and the problems encountered during oper- ation. Introduction of MSBR Pump Program The objectives of the salt pump program for the MSBR include the production of suitable and reli- able pumps for the fuel, blanket, and coolant salt circuits of the Molten-Salt Breeder Experiment (MSBE) and its noanuclear prototype, the Engi- neering Test Unit (ETU). Table 7.2 presents the pump requirements as they are presently envi- sioned. A single conditional objective requires that the pumps developed for the MSHE should be capable of flow capacity scale-up by a factor of - approximately 4 to the 550-Mw (thermal) MSBR with little or no additional development work. Our approach is to invite the strong participa- - tion of U.S. pump industry in the design, develop- ment, and production of these pumps. We will prepare pump specifications along with a prelimi- nary pump assembly drawing, pertinent rotor- dynamic and heat transfer analyses, and the re- sults of a survey of fabrication methods for sub- mission to pump manufacturers. The pump manu- facturers would be asked to make an independent analysis of the pump specifications and support- ing material and to define all the changes and im- provements they helieve necessary. Parentheti- cally, it may prove necessary to pay for several independent analyses. The pump manufacturers would then be asked to bid on the production of the detailed pump design and shop drawings, the fabrication and assembly for shop inspection of - the required quantities of salt pumps, and the shipment of disassembled pumps to ORNL for further testing. ) Two important implicit requirements are pro- ) vided in this approach. The individual pump configurations are matched to the varions MSBE systems requirements by and at ORNL, and the responsibility for approval of the final pump de- sign and the detailed drawings rests with ORNL. The principal pump components requiring de- velopment effort are the molten-salt bearing, if used, the shaft seal, and a full-scale rotor-dynamic simulator, if supercritical operation of a salt pump is required, that is, operation of the pump at speeds above the first critical shaft speed. The pump manufacturers would be invited to partici- pate in this and other development work they may Table 7.2. Pumps for Breeder Reactors Fuel Blanket Coolant 2225 Mw (thermal) MSBR Number required 42 42 47 Design temperature, °F 1300 1300 1300 Capacity, gpm 11,000 2000 16,000 Head, ft 150 26 150 Speed, rpm 1160 1160 1160 Specific speed, Ns 2830 2150 3400 Net positive suction head required, ft 25 8 32 Impeller input power, hp 990 250 1440 150 Mw (thermal) MSBE Number required 1 1 1 Design temperature, °F 1300 1300 13060 Capacity, gpm 4500 540 4300 Head, ft 150 80 150 Speed, rpm 1750 1750 1750 Specific speed, Ns 2730 1520 2670 Net positive suction head required, ft 27 5 26 Impeller input power, hp reference design or modular design. deem necessary. However, it would appear more economical to perform the molten-salt bearing de- velopment work at ORNL, where the fuel produc- tion facilities and the handling techniques are already available. Proof testing of completed pumps in molten salt prior to operation in either the ETU or the MSBE will be conducted at ORNL. Endurance testing of prototype pumps in molten salt will also be conducted at ORNL in the proof- testing facilities. Because experience indicates that production of suitable and reliable salt pumps is one of the longest lead-time items for molten-salt reactor ex- periments, it is important to get an early start in the pump program. If study indicates that the MSBE salt pumps can be operated subcritically but that the MSBR pumps must be operated supercritically, then the con- ditional objective may require operation of a salt pump at supercritical speeds during the course of the MSBE program to build confidence in the reliability of a pump with such a long shaft. Fuel and Blanket Salt Pumps Preliminary layouts have been made for the fuel salt pump, blanket salt pump, and coolant salt pump. The concepts of the fuel salt pumps, shown 410 61 350 in Fig. 7.5, and the blanket salt pump are similar, having a shaft of the order of 34 ft long, supported by an oil-lubricated radial and thrust bearing at the upper end and a molten-salt-lubricated journal bearing near the impeller at the lower end. The main differences in the two pumps lie in (1) the size of the fluid flow passages to, through, and from the impeller, (2) the absence of a pump tank on the blanket salt pump, and (3} the sizes of the drive motors. The similarity of the pumps which is derived from their common environment and placement within the cell results in common analytical and developmental efforts in the areas of bearings, seals, rotor dynamics, motor contain- ment, general layout, ditect and remote mainte- nance, ancillary systems, fabrication and assembly, and nuclear heating. The fuel salt and blanket salt pumps are de- signed so that the rotary element which contains all the moving parts can be replaced by remote maintenance without having to cut any of the salt lines to or from the pump. Direct maintenance can be performed on the drive motor and the bearing- seal assembly at the upper end of the pump. A static seal can be brought into play to separate and protect the maintenance area from the radio- activity in the pump when the bearing-seal assem- bly is to be removed. 98 ORNL-DWG S7-11817 MCTOR - COUPLING MBLY - UPPER BEARING AND SEAJL ASSE CITRIVE UL HE BELLOWS -- CONCRETE SHICLDING PUMP TANK "MOLTEN SALT BEARING -IMPELLER o Ha - L) ME2 e UIBLE HiE B - - HEAT EXCHANGER Fig. 7.5. Preliminary Layout of the MSBR Fuel Salt Pump. Nuclear heating of the pump tank and the sup- port structure within the pump tank is removed by circulating a portion of the fuel salt from the main salt stream over the heated surfaces. To remove the nuclear heat from the shaft, a small amount of salt is bled up the center of the shaft and fed into an annulus between the shaft and a cooling tube that extends the length of the pump tank. A [abyrinth seal at the lower end of the tube forces most of the salt to flow to the upper end of the tube, where it spills over into the pump tank. An added benefit is the increased damping and stiffness provided to the shaft by the salt in the annulus. Analyses are being made of the nuclear heating in that portion of the pump casings and shaft for which no cooling is provided. If a problem is found, we can provide cooling or shielding and thermal insulation where needed to reduce the heat generation in the pump structure to an ac- ceptable level. The seal arrangement at the upper end of the shaft is similar to that used in the MSRE salt pumps. It consists of a face-type seal (Graphitar against tool steel) with the 1ubricati_ng oil on one side and the helium in the shaft annulus on the other. Helium is brought into the annulus to serve as a split purge between the salt and gaseous fission products at the lower end of the shaft and any lubricating oil that leaks through the face seal Part of the helium passes down the shaft through a close-fitting labyrinth, where the increased gas velocity reduces the upward diffusion of molten-salt vapor and gaseous fission products. Concurrently, that portion of the helium passing upward through the labyrinth seal into a leak-off line. prevents the downward movement of lubricating oil vapors and also serves to scavenge oil leakage and vapors overboard from the pump. Coolant Salt Pumps Two preliminary layouts of the MSBR coolant salt pump have been prepared. One layout utilizes a pump with a short overhung shaft mounted on two oil-lubricated rolling element beatings, and the other is a long shaft with an oil-lubricated bearing at the top end of the shaft and a molten-salt bearing located just above the impeller. One criterion for the pump requires variable-speed op- eration over the range 300 to 1200 rpm. The dif- 99 ficulty with the short-shaft pump is that to have the pump operate below the first critical speed, the shaft diameter would have to be greater than 8 in., which would present a formidable seal de- sign and development problem. If it were designed to operate above the first critical and below the second critical speed, the shaft diameter would be approximately 3 in., which is inadequate to trans- mit the torque. For the long-shaft pump configu- ration, however, a shaft with a diameter selected on the basis of torque requirements would have a first critical speed well above the maximum oper- ating speed. The long-shaft pump would also use the same upper bearing and seal configuration planned for the fuel and blanket salt pumps. Hence the long-shaft concept appears to be preferable for the coolant pumps. The coolant salt pump will have the impeller and volute mounted in a pump tank of sufficient volume to accommodate the thermal expansion of the cool- ant salt for the most adverse thermal condition that might arise during reactor operation. A double vo- lute pump casing has been selected to reduce ra- dial loads on the impeller and the resultant loads on the molten-salt bearing, particularly when op-~ erating at off-design conditions, and to reduce the diameter of the bridge tube, which provides a flexible connection from the volute to the pump tank nozzle. We believe that the coolant pump drive motor, although having a greater horsepower, can be de- signed to fit the same containment vessel as that for the fuel pump drive motor. Water Pump Test Facility Preliminary layouts have been prepared of a facility for testing the fuel pump with water. The configuration does not incorporate the long shaft of the high-temperature pump but only mocks up those portions which affect the fluid flow. The layout also includes a mockup of the inlet to the heat exchanger tube sheet with sufficient instru- mentation to monitor the flow distribution in the heat exchanger tubes. The distribution of the gas injected to remove the xenon will be monitored also. The configuration has been designed to permit water testing of the blanket pump in the same facility. The purpose of the water test facility in the pump development program is (1) to determine head and flow characteristics of the impeller- diffuser design, (2) to measure radial hydraulic forces acting on the impeller (needed for designing the molten-salt bearing), (3) to measure and re- duce to an acceptable level the axial forces acting on the impeller, and to determine the relationship between the axial clearance at the bottom end of the impeller and the axial force, (4) to observe the fluid behavior in the pump tank, and to make the necessary changes to reduce gas entrainment to an acceptable level, (5) to assure that the mocked-up molten-salt bearing will run submerged under all operating conditions, and (6) to check the point of cavitation inception and the required net positive suction head of the impeller. Molten-Salt Bearing Tests The present layouts of the MSBE and MSBR salt pumps require a molten-salt journal bearing near the impeller. A molten-salt bearing presents three important considerations: (1) the hydro- dynamic design of the bearing to provide the requisite lubricating film, (2) the selection of the kind and form of the bearing materials, and (3) the design of a bearing mounting arrangement which will preserve the lubricating film despite thermal distortions between pump shaft and casings. We are studying the use of hard, wear-resistant coatings on the journal and bearing surfaces. Such coatings present advantages over the sintered, solid-body journal and bearing inserts most often used in high-temperature process fluid lubrication. The hard coatings are convenient to apply and hopefully eliminate the differential thermal expansion problems. Mechanical Tech- nology, Inc., of L.atham, New York, has been engaged to produce Hastelloy N specimens with each of four different hard coatings: (1) cobalt (6 to 8%) bonded tungsten carbide, (2) nickel (7%) bonded tungsten carbide and mixed tungsten- chromium carbides, (3) nickel-chromium (15%) bonded chromium carbide, and (4) molybdenum (7%) bonded tungsten carbide. These coatings will be subjected to corrosion and thermal cycling tests in molten salt at ORNL. A test in molten salt will be made with a 3 x 3 in. bearing using one of these coatings, if one should prove satisfactory. A layout is being made of a tester to accom- modate a full-scale molten-salt bearing for the 100 MSBE fuel salt pump. The tester will be capable of subjecting the hearing and its mounting arrange- ment to start-stop wear tests and thermal cycling and endurance tests in molten salt. Rotor-Dynamics Feasibility lnvestigation Mechanical Technology, Inc., is performing an analysis (Reactor Division subcontract No. 2942) of the rotor dynamics of the preliminary layout of the MSBR fuel salt pump to determine its flexural and torsional critical speeds and flexural response to a dynamic unbalance. Interim results? of the analysis show that the pump will operate between the fourth and fifth flexural critical speeds of the pump system, which includes the pump shaft, inner and outer pump casings, and the drive motor. The third and fifth system criticals are essentially the first and second simply supported beam criti- cals of the shaft. The critical-speed results also show that the pump-system criticals are relatively independent of the bearing stiffness over a range representative of practical bearing designs. The stiffness characteristics of the drive motor coupling also have little effect on system ciiti- cals. The synchronous response amplitudes resulting from a ‘“‘bowed-shaft’’ been calculated over the complete range of pump speeds. The response results show only one sys- tem critical to be significant from a bearing load standpoint -- namely, the ‘‘first shaft critical®’ unbalance condition have which occurs at about 700 rpm. In addition to passing through one shaft critical speed, three additional system criticals must be traversed as the pump accelerates to design speed. Thesec three criticals are basically cantilever resonances of the outer casing. The first two cantilever modes occur at quite low speeds and hence should not be a problem from a steady-state standpoint. However, if the pump system should be transiently excited during normal operation, these cantilever beam modes would be the primary contributors to the resulting transient vibration response of the pump system. The third cantilever beam mode of the outer casing also excites a simply supported resonance of the inner casing. This mode occurs between 2P. W. Curwen, Rotor-Dynamic Feasibility Study of Molten Salt Pumps for MSBR Power Plants, MTI-67TR48, Mechanical Technology, Inc., August 6, 19867, 800 and 930 rpm and makes it appear advisable to separate the inner and outer casing frequencies by suitable changes in the wall thicknesses and diameters of the two casings. A preliminary undamped torsional critical-speed analysis has been made for the pump system, and the results indicate that the two torsional critical speeds that might affect pump operation can be strongly dependent upon the electromagnetic torsional stiffness of the drive motor. We believe that by changing some of the component dimensions and accounting for inherent system damping, the pump will operate satisfactorily between the first and second torsional critical speeds. Fabrication Methods Survey. — Based on the preliminary layouf of the MSBR {uel salt pump, the shaft and inner and outer casings were detailed, and a survey is being made of potential fabrication methods to identify fabrication problems. For the shaft, it is important to determine the straightness and concentricity tolerances that can be supplied. These tolerances have considerable effect on the provisions that must be made for the dynamic balancing of the shaft, which must be done rather precisely when the pump is to operate above the first shaft critical speed. A manufacturer was found who could fabricate shafts in the range 7 to 10 in. outside diameter with a wall thickness as small as 1/2 in., and who would guarantee the straightness from end to end to 0.005 total indi- cator reading (TIR) and the outside diameter— inside diameter concentricity to 0.005 in., but at considerable expense. As these two tolerances are relaxed, more manufacturing capability is available, and the fabrication costs are reduced; however, dynamic balancing of the shaft becomes a more important portion of the fabrication process. 101 An investigation is under way to determine the relationship between shaft precision and total shaft cost (fabrication plus dynamic balancing) as well as its effect on pump design. Several manufacturers have been found who are capable of fabricating the inner and outer casings to the tolerances shown on the preliminary layout, but also at considerable expense. The effects on pump design of relaxing the preliminary values of the tolerances are being studied. Other Molten-Salt Pumps Fuel Pump High-Temperature Endurance Test Facility. — The facility,? incloding the salt pump, gas systems, instrumentation, and handling equip- ment, is being prepared for operation with sodium fluotoborate (NaBF 4) . The new drive motor, rated 200 hp at 1800 rpm, was delivered, and the existing motor support was modified to suit the new motot. Also, the pump rotary element was removed from the facility, disassembled, cleaned, and reassembled. Molten-Salt Bearing Pump Endurance Test Facility. — A new salt bearing constructed of Hastelloy N was installed on this pump.® The gimbals support for the bearing was modified to reduce the possibility of the support becoming disassembled during operation. The bearing and gimbals arrangement was satisfactorily tested with oil as the pumped fluid at room temperature. SMSR Program Semiann. Proge. Rept. Feb., 28, 1967, ORNL-4119, p. 66. *MSR Program Semiann. Progr. Rept. Aug. 31, 1966, ORNIL.-4037, p. 82, Part 3. Chemistry W. R. Giimes The chemical research and development effort in close support of the MSBR program includes, as described in this chapter, a variety of studies. A major share of this effort 1s still devoted to the immediate and anticipated problems of the operat- ing Molten-Salt Reactor Experiment. Sampling of the MSRE fuel and coolant salts and interpretation of the analyses for major and minor constituents of the melt, and examination of metal and graphite surveillance specimens from the core and of specimens exposed to the pump bowl gases, continue as routine, though obviously necessary, portions of the total effort. Minor fractivns of several fission products continue to appear in the pump-bowl gas space. The possibility that these may occur as volatile compounds has prompted examination of the chemistry and the vaporization behavior of the litile-known intemmediate valence fluorides of molybdenum. Oxide-fluoride equilibria in the LiF-BeF , sys- tem and its more complex counterparts with added UF , and ThF , are under study, since such equilibria may well lead to separation processes of value and seem to have shown us container materials that will greatly aid our experimental program. The plans to substitute a lower melting and more economical coolant for the LiF-BelF , mixture in the MSRE have required examination of phase be- havior among the alkali fluoroborates and of ancillary questions of the decomposition pressuze of these materials and of possible undesirable interactions of BF ; with metals and lubricants to which it would be exposed. While recovery of uranium by fluorination (both from fuel and blanket) and recovery of fuel salt by vacuum distillation remain as the design re- processing methods, the recovery of protactinium from the blanket and the removal of fission prod- ucts from the fuel by reductive extraction iato molten metals continue to show promise and are actively pursued. Development studies in analytical chemistry have been directed primarily to improvement in analysis of radioactive samples of fuel for oxide and uranium trifluoride and for impurities in the helium gas from the MSRE. 8. Chemistry of the MSRE R. 8.1 MSRE SALT COMPOSITION AND FURITY Molten fuel, flush, and coolanat salts have been intemmittently circulated and stored in the MSRE for approximately two years. In use, these salts have been subjected to chemical analysis on a regular basis.! The results of these analyses 102 E. Thoma signify that generalized corrosion in the fuel and coolant circuits is practically absent, and that lChemical analyses performed under the supervision of C. E. Lamb, Analytical Chemistry Division; spec- trochemical data were obtained by W. R, Musick, Analytical Chemistry Division. the salts are currently in essentially as pure condition as when charged into the reactor. Fuel Salt MSRE runs 11 and 12 were completed within the current report period. During this period, small amounts of beryllium metal were dissolved into the fuel salt to adjust the oxidation-reduction potential of the salt. The total concentration of uranium in the fuel was also increased by addition of "LiF-?3°UF, to the circulating salt. Curmently, the uranium concentration of the fuel salt is approximately 4.590 wt %, of which the U?" frac- tion of the total uranium is 1.5%. The effects of the beryllium and 7LiF—”‘qUF4 additions are evident in the results of the chemical analyses of the fuel salt shown in Table 8.1. A refinement of analytical procedure was introduced during run 11; preliminary values for the detemination of ura- nium concentrations were confirmed by a second group of analysts before final values were re- ported. Continuous control methods were em- ployed by both groups. This innovation in pro- cedure resulted in a significant improvement in precision. Average scatter was reduced from 10.5% to +0.4% of the value, corresponding to +0.03 and 10.02 wt % uranium. MSRE fuel salt is analyzed by HF-H , purge methods for evidence of oxide contamination. The results of analyses obtained during runs 11 and 12 indicated that the fuel salt does not contain more than 50 to 60 ppm of oxide; that is, it is currently as free of oxide as when it was originally charged into the MSRE. Coolant Salt When run 12 was terminated in August 1967, the coolant salt had circulated in the MSRE for a perod of 12,047 hr. Coolant salt specimens were submitted at one-month intervals during 1967. Results of those analyses show the composition and purity of the salt to be Li Be F Fe Cr Ni 0 {wt %) (ppm) 13.04 9.47 76.30 63 27 12 ™200 103 as compared with the composition and purity it was known to possess a year ago: Li Be F Fe Cr Ni 0 (wt %) (ppm) 13.03 9.46 76.16 58 36 16 ™200 The two compositions are not differentiable with- in the precision of the analytic methods. The constancy of the trace concentrations of the impurities attests to the fact that the cover gas, which is supplied to both the fuel and coolant circuits from a common source, has been main- tained in a state of high purity throughout the entire operational period. Flush Sait Whenever the MSRE fuel circuit is flushed with flush salt, there is cross mixing of fuel and flush salts by residues which remain in the reactor after each is drained. We need to know the amounts of material transferred between the fuel and flush salts because they enter into the calcu- lation of the book-value concentration of uranium in the fuel salt. Sufficient analytical data are now available to enable us to deduce the average mass of these residues. The concentration of uranium in the flush salt appears to change in nearly equal increments duting each flush operation, as shown in Table 8.2. These data indicate that fuel salt which remains in the fuel circuit after drainage of the fuel increases the uranium concentration of the flush salt by 200 ppm each time the drained re- actor is cleaned with flush salt. An increase of 200 ppm of uranium corresponds to the addition of approximately 850 g of uranium to the flush salt. This is the amount of uranium in 19.20 + 0. 10 kg of fuel salt in which the uranium concentration is between 4.570 to 4.622 wt % U, the range for the MSRE during the period considered. On filling the MSRE with fuel salt, "LiF-BeF, (66-34 mole %) flush salt residue is incorporated in the fuel salt, diluting its concentration of UF | and ZrF | slightly. This dilution is reflected in the uranium and zirconium analyses shown in Fig. 8.1. The decrease in zirconinm concentration of the fuel salt from a mean value of 11.33 wt % to 10.85 wt % corresponds to dilution of the fuel by 12.7 kg of salt on each drain-flush-fill cycle. Table 8.1, Summary of MSRE Fyel Sait Analyses, Runs 11 and 12 Sample Li Be Zr :na;yt Ubook F Fe Cr Ni Total No. (wt %) {(wt %) wt %) {wt %) {wt %) (wt %) {ppm) (ppm) (ppm) (wt %) ¥FDP11-01 11.18 6.28 10.90 4,603 4,576 67.46 131 66 54 100.44 FP11-02 11,10 6.27 10,65 4,599 4.576 69,16 112 75 63 101.81 FP11-03 10.42 6.21 11.08 4,604 4,575 565,72 150 61 64 99,06 FP11-04 11,190 6.33 11.19 4,562 4.575 67.87 143 62 22 101.01 FP11-05 wt/SU - 0.37% FP11-06 11.25 6.31 10,97 4,555 4,574 67.44 131 67 33 100.54 FP11-07 11.38 6.70 11.27 4.558 4,574 69.92 172 62 50 103.86 FP11-08 i1.50 6.62 10,81 4.569 4,573 68.89 312 54 107 102.43 FPi1-09 Gas samplie ¥Pii-10 Beo addition, 11.66 g Frii-11 10.57 6.57 10.87 4.551 4,572 69.94 165 73 43 102.62 ¥FP11-12 10.93 6.27 10.88 4.567 4,572 69.96 76 75 34 102.62 FP11-13 U T2U = 0.42% FP11-14 10.98 6.35 11.26 4.539 4.571 67.83 98 78 75 100.98 FP11-15 1.11 6.49 18,74 4,552 4,571 68.89 i 62 37 101,80 FP11-i96 Gas sample FP1i-i7 11,33 6.47 11.21 4,553 4.571 70.25 67 58 47 193.85 FP11-18 10.73 6.67 11.17 4.579 4,570 68.68 120 56 43 101,85 FP11-i9 10.47 6.57 11,11 4.589 4,570 67,12 117 68 42 90.38 FP11-20 10,51 6.72 10.98 4,561 4,570 67.60 122 59 49 100.39 FP11i-21 10.50 6,36 10.89 4.576 4,570 66,823 168 63 46 99.19 FP1:-22 10.55 6.57 10,92 4.572 4.559 66.05 104 62 63 98.68 FP11i-235 10.53 5,49 10.95 4.583 4,569 69,70 136 63 55 102.28 FP1i-24 1G.55 6.51 10,99 4,547 4.5569 58.50 i73 67 63 101,04 FP11i-25 Sampie for oxide analysis; analysis unsuccessful FP1i-29 10.43 .49 10,91 4,57C 4.568 67.17 118 52 53 99.59 FP11-27 10,52 .85 10.85 4,577 4.568 66.04 72 63 &80 90,84 FP11-28 Sample for oxide analysis; oxide concentration, 58 ppm FP11-29 10.48 45 10.92 4.584 4.567 64.52 126 64 56 97.08 FPil-30 10.48 .39 11,02 4.597 4.597 64.51 i15 56 73 97.01 FP11-31 1G.53 58 11.06 4,559 4,566 67.06 80 64 50 95,81 FP11-32 U320 - 0.34% FP11.33 10.53 65.43 11.14 4,567 4,564 66,38 142 72 72 99,07 FP11-34 10.55 5.33 11.37 4.582 4.5646 68.99 146 64 64 101,85 FPii-35 10.53 6.35 1i.12 4.566 4.565 67.24 194 73 64 99,84 1 Table 8.1. (continued) Sample Li Be Zr Zn‘alyt Ubook F Fe Cr Ni Total No. (wt %) (wt %) {(wt %) (wt %) (wt %) (wt %) . {ppm) {ppm) {ppm) {wt 7o) FP11-35 Gas sample FP11-37 10.55 6.33 10.75 4,541 4,565 65.93 79 80 49 96.12 FP-11-38 Sample for us +/S.U analysis; analysis unsuccessful FP11-39 11.57 6.44 10.92 4.536 4,565 66.55 182 69 52 100,05 FP11-40 Be? addition, 8.40 g FP11.4% 10.42 6,37 10.77 4.579 4,564 68.5 135 56 58 100.74 FP11-42 Gas sample FP11-43 ud +/):U — no analysis performed FP11-44 10.50 6.60 11.01 4.561 4,564 69,88 140 59 44 102,57 FP11-45 10.58 6.50 10.65 4,548 4,563 67.13 88 54 41 09,43 Average 10.80 * 0.35 6.46 £ 0.15 10.97 £0.18 4,570 £ (.018 67.81 £1.46 131 +48 64 L6 54 £ 6 FP11-46 Gas sample FP11-47 10.95 6.48 10,96 4,604 4.582 66,65 169 71 75 949.67 FP11-48 10.45 6.52 10.85 4,578 4.581 67.23 210 49 58 99,66 FP11-49 U +/EU — no analysis performed FP11-50 Gas sample FP11.51 11.20 6.45 10.97 4,571 4,580 69.61 114 61 61 102,83 FP11-52 11.33 6.45 10.79 4,566 4,580 69,27 80 60 25 102.43 FP11-53 Gas samnmple FP11-54 10.93 6.63 10.95 4,551 4.579 £69.91 158 61 63 103.00 FP11-35 Special 50-g sample FP11-56 Sample for oxide analysis; oxide concentration, 50 to 100 ppm FP11-57 No sample obtained FP11-58 10.48 6.60 11.27 4.607 4.578 70.95 131 81 42 104.00 Average 10.82 £0.35 6.47 +0.15 10.97 £0.17 4,571 £0.019 57.09 £1.33 133 £47 64 17 54 T 16 FP11-59 13.43 7.57 0.345 0.0292 76.18 222 68 40 97.51 FP11-60 13.40 9.36 8,260 0.0268 79.23 110 74 34 102,30 FP12-01 13.40 8.64 <0.20 0.6778 80.22 108 66 26 102.56 FP12-02 13.70 0,590 <0.20 8.07%3 75.40 104 7 23 88,90 FP12-03 13.60 9,38 <0.20 0.0826 77.4Q a3 63 24 100,68 FP12-04 Sample for oxide analysis FP12-05 11.20 6.74 10.94 4.550 4.548 66.32 123 52 60 99.77 FP12-06 udt/Zu = 0.37% FP12-07 Gas sample FP12-08 Be addition, 7,93 g FP12-09 Be additien, 9.840 g SOT Table 8.1, (continued) Sample Li Be Zr :naiyt Ubook F Fe Cr Ni Total Ne. (wt 70) {wt %) (Wt %) (wt %) {wt %) (wt 7o) {ppm) {ppm) {ppm) (wt %) FP12-10 11.60 6.91 10.57 4.525 4,547 67.27 i34 71 72 141.90 FPi2-11 U2 = 1% FP12-12 11.60 6.54 10.91 4,545 4.547 66.66 113 64 62 100,78 FPI2-13 Be addition, 8.33 g FP12-14 11.50 6.50 11.22 4.557 4.546 67.95 145 82 47 101.76 FP12.15 Be addition, 11,68 g FPi12-16 11.40 6.40 10,62 4.567 4.546 68.27 269 110 08 101,31 FP12-17 11.30 6.40 10,656 4.532 4,545 66.92 215 144 53 99.85 ¥P12-18 Sample for oxide analysis; oxide concentration, 57 ppm FP12-19 11.50 6.19 11.00 4,522 4.545 65.05 100 102 62 98.29 FP12.20 10.60 6.36 10,76 4,557 4.545 65.76 &1 654 44 G8.06 FP12-21 U3 /ZU - 0.5% FP12.22 10,50 6.52 10.43 4.566 4.544 86.18 247 ad 50 GR.22 FP12-23 10.60 5,68 10,58 4,526 4,544 H5.26 154 78 76 a7.67 FP12-24 11.38 5.36 10.67 4.567 4.544 66.46 176 av 56 99.47 ¥FP12-25 10.70 5.46 1G.53 4.496 4.544 66,10 208 68 62 98.32 FP12-26 Gas sample FP12-27 10.70 6.61 10.78 4,550 4.544 67.50 195 75 72 100.18 FP12-28 10,70 6.44 10.656 4.569 4,544 69,00 150 68 52 101.40 FP12.29 10.70 6.41 10.87 4,520 4.543 67.11 177 &4 78 99.65 FP12-30-35 "LiF-?37UF , additions FP12-36 10.50 6.58 10.96 4,554 4,555 66.40 156 84 359 99.15 FP12-37-40 "LiF-?*°UF additions FP12-41 10.60 6.52 10.68 4.562 4,565 68.00 110 64 192 100,40 FP12-42-46 "LiF-??UF, additions FPi12-47 10.60 6.50 10.50 4,586 4.576 65.34 i20 72 70 98.56 FP12-468-50 "LiF-??PUF | additions FP12-51 i1,22 6.60 10.32 4.588 4,576 66,32 94 72 39 99,07 FP12-52 10.70 6.47 10.71 4.5084 4.576 65.98 119 72 932 98.48 FP12-53 10.50 6.39 10.66 4,503 4.575 65.48 132 72 60 97.56 FP12-54 Sample for isotopic dilution analysis FP12-55 10.75 6,44 11.12 4.577 4,574 65.72 156 72 300 98.66 10.80 6.42 10.95 4,575 4.574 54,57 136 58 720 97.42 901 Table 8.1. (continved) Sample Li Be Zr :nalyt Ubook F Fe Cr Ni Total No. (wt %) (st %) (wt %) (Wt %) (wt Te) {(wt %o) (ppm) (ppm) {(ppm) (wt To} FP12-56 Be addition, 9.72 g FP12-37 11.33 6.59 11.08 4.587 4,574 65,14 156 34 170 98.77 11.30 6,74 11.28 4,549 4,574 66.53 160 654 424 100.46 FP12-58 11.00 6.58 10.77 4,600 4,574 66.62 138 74 606 99.60 FP12-56 Sample ladle remained in pump bowl Average 10.93 6.25 10.78 66,50 0,40 +3.15 10.24 10,03 a - . . fak Corrected to compensate for isotopic composition, LOT 108 Table 8.2, Chemical Analyses of MSRE Flush Salt Specimens Average Uranium Average Increase in Number of Samples Run No. Found Uranium for Flush (ppm) Analyzed (ppm) FP-3 (final) 195 1 195 P-4 (initial) 218 6 218 FP-8 (initial) 460 3 230 FP«8 (final) 616 1 205 FP-9 (final) 840 1 210 ¥P-11 (final) 930 1 186 FP-12 (initial 799 3 160 FP-13 (initial) 1186 3 197 e Overall average = 200 ppm Implications of Current Experience in Future Operations Currently, the MSRE is entering its final period of operation with ?3°U fuel salt. It is planned that the MSRE will operate with ?33U fuel in 1968, % as will the MSBE later, and that the con- centration of uranium tetrafluoride in those fuels will be only one-fourth that employed in the MSRE. Several inferences may be drawn from the ex- perience developed during previous "operation which have significant implications regarding operation of the MSR[Z when it is charged with 2337 fuel, as well as for larger molten-salt re- actors. In general, we must conclude that if chemical analyses are to function as operational controls, appreciably greater precision than is now available must characterize the methods for determining the concentration of uranium as well as the U®" fraction in the total uranium. The overall composition of the present fuel salt may be determined in routine chemical analysis with a precision of 0.2 to 0.3 wt % (Fig. 8.1). Precision in the determination of the uranium con- centration is considerably better, £0.02 wt % on a statistical basis (Fig. 8 2). Such precision in the detemination of the uranium concentration is, however, only one-tenth that which is obtained in outine computations of reactivity balance. The Zp. N, Haubenreich et al. to R. B. Briggs, private communication, Dec. 19, 1966, high sensitivity in the reactivity balance to varia- tions in uranium concentration vitiates applica- tion of periodic batch analysis of fuel as a sig- nificant control parameter in reactor operations; such analyses have come to function primarily as an independent basis for cross-checking burnup and inventory computations. It is anticipated that when the reactor is fueled with 233U, the pre- cision of the reactivity balance will be improved by a factor of at least 4,° while the precision of the chemical assay of uranium will fall in pro- portion to the uranium concentration as it is re- duced from 0.83 to 0.20 mole %. The limitations on the present methods of analyzing the MSRE fuel indicate, therefore, that it will be necessary to develop improved methods for detemining fuel - composition. In reactor systems in which fre- quent or nearly continuous chemical reprocessing is carried out, composition of the fuel and blanket - systems will undergo constant change. Un- questionably, composition determination will then be necessary by way of on-line techniques sup- plemented by methods which have high intrinsic accuracy. The intiinsic corrosion potential of the fuel salt is proportional to the UF | concentration, which, to date, has been determined directly only by an intricate and difficult method which is probably near its limit of capability with salt of 3J. R. Engel to R. E. Thoma, private coiununica- tion, April 28, 1967. LITHIUM (wt %) BERYLLIUM {(wi %) ZIRCONIUM (wi 9% (W‘.L (yo} URANIUM CHROMIUM (oprn) -~ ~ IRON {ppm) NICKEL {ppr} 12.00 14,50 - 14,00 b §0.50 ot 10.00 7.50 - 7.00 4.600 - 4,550 +— 4.500 100 80 &0 40 |- 20 200 150 160 ORNL.-DWG 67 -14818 50 |- Fig. TrTY fu T e ,4}"’____ {I L L | T T LI ________ Ldoe topord g1 T TTTTT T { 4>}{17§ } \}i ,,,,,,, * e } 1 L . LH}E %4} ,,,,,,, } i AmE ”M_;w “““ - [ T I ] . T L g o ® { - fl _______ S __d B """ T T T Tl e b 1T 1 | - T e " T T 7T ] | L |L i *""“filf T 109 ORNL-OWG 67T-11353 QB0 (e ?{CHEM{CAL ASSAY 2600 | % - R — A5G0 & 4540 | URENILM (wt %) 4520 4500 |1 J RUN A1---— 2 24 ™} 30 VEGAWATT HOURS (x10%) &8 a2 Fig. 8.2. Urenium Conceniration in MSRE Fuel Seit. 4 While this method has been used with moderate success with the MSRE fuel salt, the low total concentration of uranium which is anticipated in future fuel salts makes it improbable that this method can have continued application. It will be of considerable importance in the near future to employ direct spectrophotometric metheds for the determination of U?" concentration in the fuel salt. Results of recent laboratory experiments indicate that this approach is feasible. 5 In the future, the MSRE fuel salt as well as the fuel salts in the large reactor plants will be sub- jected to fluorination and to HF-H , purge streams during chemical reprocessing. Salt streams in those reactors may be expected to contain even lower concentrations of contaminant oxides than currently exist in the MSRE and should therefore not require oxide analysis. the present uranium concentration. Results of the chemical analysis for chromium have shown sufficient precision (::10%) that the method has come to serve as an excellent and reliable measure of generalized corrosion within the MSRE. The utility of this analysis as an indicator results from the fact that at present the total concentration of chromium in the fuel salt is low (~70 ppm). Relatively minor changes in cotrosion are, therefore, reflected in significant changes in chromium concentration. In future operation it is possible that the total concentra- tion of chromium in the fuel circuit will increase 4A. 8. Meyer, Jr., to R. E. Thoma, private commumni- cation, May 12, 1967. SJ. P. Young, MSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4119, p. 163. tenfold or more as a consequence of chemical reprocessing. Unless either the precision of chromium analysis is improved or low concentra- tion of chromium is maintained in the salt which is returned to the reactor fuel circuit, much of the capability for immediate detection of corrosion will have been lost. It is anticipated that gas chromatographic methods for analysis of gas streams will be tested soon at the MSRE. If application of such methods to cover gas analysis succeeds in providing a sensitive means forthe quantitative detemmination of volatile fluoride, hydrogen, and oxygen-bearing phases, a major advance toward on-line analysis of salt purity will have been achieved. 8.2 MSRE FUEL CIRCUIT CORROSION CHEMISTRY Corrosion on salt-metal interfaces in the MSRE is signaled by an increase of chromium concen- tration in salt specimens. An increase of 10 ppm corresponds to the removal of approximately 40 g of chromium from the Hastelloy N surfaces. Cur- rently, the chromium concentration of the fuel salt is 72 =7 ppm; this concentration represents an increase of only 34 ppm and removal of about 170 g of chromium from the Hastelloy N container since operation of the MSRE began in 1965. If the total amount of chromium represented by this increase were leached uniformly from the fuel circuit, it would correspond to removal of chromium from a depth of 0.22 mil. Recent evidence indi- cates, however, that only half the chromium in- crease observed in the fuel salt may be attributed to corrosion in the fuel circuit. On temnination of run 7, graphite and metal surveillance specimens were removed from the core of the reactor and were replaced with speci- mens contained in a new perforated metal basket. Fuel specimens taken throughout the next run, No. 8, were found to contain a chromium concen- tration of 62 ppm, rather than 48 ppm, the average concentration which had persisted almost from the beginning of power operations. Since the only known environmental alteration was the installa- tion of the new surveillance specimen assemblage, we speculated that the container basket and the Hastelloy N specimens had sustained most of the corrosion responsible for the observed increase in chromium, and that corrosion might be evident 110 to a depth of 10 mils. However, recent inspection of specimens from the basket (see Part 5, this report) did not disclose that the anticipated cor- rosion of the metal had occurred. On several previous occasions, salt was re- turned to the fuel circuit after storage in the drain tanks without developing evidence of an increase. Nevertheless, we are forced to conclude that the increase of chromium in the fuel salt took place while it was stored in the drain tank during the ten-week interval between runs 7 and 8. Although it is not evident how the drain tank may have become contaminated, its surface seems to be the source of the additional chromium in the fuel salt. If all the chromium was leached uni- formly, corrosion in the drain tank will have reached a depth of 0.7 mil. If the increase of 48 to 62 ppm is attributed to the drain tank, the total increase of chromium resulting from fuel- circuit corrosion is only 20 ppm throughout the entire operation of the MSRE, and corresponds to ~ 100 g of chromium, or 0.13 mil of generalized corrosion in the fuel circuit. 8.3 ADJUSTMENT OF THE UF, CONCEN. TRATION OF THE FUEL SALT The fuel salt, free of moisture and HF, should remove chromium from Hastelloy N only by the equilibrium reaction L 0 . o ,'2CI' + UF4 —_— /2ch 2 +UF3 (d) (d} {d) When the above corrosion equilibrium was first established in MSRE power operations, the UF, . produced in this reaction, together with that - originally added to the fuel concentrate, should have totaled 1500 g, with the result that as much as 0.65% of the uranium of the system could have been trivalent soon after the beginning of power operation. The UF, content of the MSRE fuel was determined after approximately 11,000 Mwhr of operation to be no greater than 0.05%. The fuel salt was considered to be far more oxidizing than was necessary and certain to become more so as additional power was produced unless ad- justment was made in the UF, concentration. A program was initiated early in 1967 to reduce 1 to 1.5% of the uranium inventory to the trivalent state. The U®"* concentration has been increased 111 since that time by addition of 84 g of beryllium metal. The method of addition was described previously. ® The low corrosion sustained by the MSRE fuel circuit, which is in general accord with the re- sults from a wide variety of out-of-pile corrosion tests, might have been expected to be greater during the first 10,000 Mwhr of operations be- cause the UF, concentration of the fuel was markedly less than was intended. According to Grimes,’ ““The lack of corrosion in the MSRE melts which appear to be more oxidizing than intended can be rationalized by the assumption (1) that the Hastelloy N has been depleted in Cr (and Fe) at the surface so that only Mo and Ni are exposed to attack, with Cr (and Fe) reacting only at the slow rate at which it is fumished to the surface by diffusion, or (2) that the noble- metal fission products are forming an adherent and protective plate on the reactor metal.”’ If it is assumed that corrosion of the fuel pro- duced 3.3 equivalents of UF , and that fission has resulted in the oxidation of 0.8 equivalent of UF, ®R. E. Thoma and W. R. Grimes, MSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL.-4119, p. 123, 7W. R. Grimes, Chemical Research and Develop- ment for Molten-Salt Breeder Reactors, ORNL-1853, p- 70 (June 6, 1967). per gram-atom of fissioned uranium,® the addi- tions of beryllium metal to the fuel have been fol- lowed by the concentrations listed in Table 8.3. The calculated values ate for the most part higher than the measured values. The cause of this disparity is not currently understood, but is under investigation. All dissolutions of beryllium metal into the fuel sall proceeded smoothly; the bar stock which was withdrawn after exposure to the fuel was observed to be smooth and of symmetrically reduced shape. No significant effects on reactivity were ob- served during or following the beryllium additions, nor were chemical analyses indicative that such additions were made until after run 12 was begun. The additions preceding run 12 had increased the U3" concentration in the total uranium by ap- proximately 0.6%. Four exposures of beryllium were made at close intervals during the early part of that run. Samples taken shortly after the last of these four exposures (FP12-16 ef seq., Table 8.1) began to show an unprecedented in- crease in the concentration of chromium in the specimens, followed by a similar decrease during the subsequent sampling period. Previous laburatory experience has not dis- closed comparable behavior, and no well-defined 81bid., p. 65. Table 8.3, Concentration of UF3 in the MSRE Fuel Salt? Uranium Uranium yl * Net ul +/EU u? +/EU Sample Mwhr Burned Burned Oxidized Be Added Be Added Equivalents Calculated Analysis No. (@) (moles) (moles) ) (equivalents) poguced %) () 9-4 10,978 554 2.34 1.87 0 0 3.13 0.31 0.1 10-25 16,450 829 3.50 2.80 16.28 3.61 5.81 0.58 0.5 11-5 17,743 953 4.02 3.20 16.28 3.61 5.21 0,53 0.37 11-13 20,386 1029 4.34 3.46 27.94 6.20 8.74 0.88 0.42 11-32 25,510 1287 5.43 4.34 27.94 6.20 6.86 0.69 0.34 11-38 27,065 1365 5,76 4.60 27.94 6.20 6.60 Q.66 11-49 30,000 1514 6.39 5.10 36.34 8.06 7.96 0.80 12-6 32,450 1637 6.91 5.50 36.34 8.06 7.76 0.77 0.37 12-11 33,095 1670 7.05 5.60 54.11 12.01 11.40 1.14 1.2 12-21 35,649 1798 7.59 6.10 74.12 16.45 15.35 1.5 0.5 “These numbers assume that the salt originally was 0.16% reduced; that the increase in Cr from 38 to 65 ppm was real, occurred before 11-14-66, and resulted in reduction of U + atom of U3+; and that there have been no other losses of U3 + + . . 4 to U3 ; that each fission results in oxidation of 0.8 mechanism is available which satisfactorily A possible cause, which is partially supported by experi- accounts for the observed behavior. mental data, is described as follows. On the two occasions when the most rapid rates of dissolution of the beryllium rods were observed, chromium values for the next several fuel samples, FP11-10 et seq., and FP12-16 et seq., rose tempo- rarily above the lo level and subsequently re- turned to normal. That the increase in chromium levels in samples FP12-16 to -19 was temporary indicates that the high chromium concentration of fuel samples removed from the pump bowl was atypical of the salt in the fuel circuit and implies that surface-active solids were in suspension at the salt-gas interfaces in the pump bowl. That atypical distribution of species in this location does indeed take place was demonstrated earlier by the analysis of sample capsule support wires that were (1) submerged below the pump- bowl salt surface, (2) exposed to the salt-gas interface, and (3) exposed to the pump-bowl cover gas. The results showed that the noble-metal fission products, Mo, Nb, and Ru, were deposited in abnormally high concentrations at the salt-gas interface. Such behavior suggests that the high chromium concentrations in the fuel specimens were caused by the occurrence of chromium in the Fig. 8.3. Surface Appearances of Fuel Salt Specimens Taken Before and After Beryllium Addition. (b) FP12-57. 112 R-39267 pump bowl in nonwetting, surface-active phases in which its activity was low. A possible mecha- nism which would cause such a phenomenon is the reduction of Cr?" by Be® with the concurrent reaction of Cr? with graphite present on the salt surface to form one or more of the chromium car- bides, for example, Cr3C2 (AHOf: —21 keal at 298°K). Such phases possess relatively low stability and could be expected to decompose, once dispersed in the fuel-circuit salt. The possibility that surface-active solids were formed as a consequence of the Be? additions was tested late in run 12 by obtaining salt specimens at the salt-gas interface as well as below the sur- face. First, specimens were obtained in a three- compartment sample capsule that was immersed so that the center hole was expected to be at the interface. Next, a beryllium metal rod was ex- posed to the fuel salt for 8 hr with the result that 9.71 g of beryllium metal was introduced into the fuel salt. Twelve hours later a second three- compartment capsule was immersed in the pump bowl. Chemical analyses of the fuel salt speci- mens FP12-55 and -57 (Table 8.3) do not show significant differences in chromium; however, the salt-gas interface in FP12-57 is blackened as as compared with FP12-55 (see Fig. 8.3). R-39266 (a) FP12-55, 113 Fig. 8.4. Appearance of Upper Part of Three-Compartment Sample Capsule FP12-57 After Use. An additional purpose of sampling with the three-compartment capsule was to detemine whether foamlike material was present in the sampler area and would be collected in the upper compartment. However, analysis of the upper compartment has not yet been performed. Glob- ules were noted on the upper part of FP12-57 (Fig. 8.4), indicating that conditions in the pump bowl were different when samples FP12-57 and -55 were obtained. Examination of the metal basket which con- tained the beryllium rod while it was exposed to the fuel showed the presence of dendritic crystals along with a small amount of salt residue (Fig. 8.5). Spectrochemical analysis of material re- moved from the basket (Fig. 8.6) indicated that the material contained 7.8 wt % chromium and less than 10 ppm of iron and nickel. The evidence obtained to date does not permit inference as to the identity of the phases which have formed within the pump bowl as a conse- quence of the beryllium additions. It does strong- ly imply that nonwetted flotsam can be formed and accumulated temporarily in the MSRE pump bowl. R-39264 114 The overflow tank has not been mentioned as a possible factor in the behavior of chromium fol- lowing a beryllium addition. It could, however, bear some responsibility for the persistence of chromium for several days following an exposure of the salt to beryllium. Fuel salt accumulates in the overflow tank steadily during operation and remains there in relative isolation from the fuel stream. At intervals of about one day, part of the salt (60 1b) is retumed to the fuel stream. Recog- nizing that chromium might be injected into the pump bowl as the salt retums, we performed an experiment in which salt samples were obtained from the pump bow! within an hour after fuel was retumed from the overflow tank to the pump bowl. The purpose of the experiment was to determine whether material from the overflow tank con- tributed appreciably to the perturbations in the chromium concentration. The results were nega- tive, possibly, in part, because sampling and salt transfer operations were not performed con- currently for safety reasons. Fig. 8.5. Dendritic Crystals and Salt on Basket of Beryllium Addition Capsule FP12-56. 115 R-39265 Fig. 8.6. Residue from Beryllium Addition Capsule FP12-56. 9. Fission Product Behavior in the MSRE S. S. Kirslis Results of previous tests in-pile and in the MSRE have been very reassuring regarding most aspects of the chemical compatibility of graphite and Hastelloy N with molten fissioning sait. The aspect of chemical behavior currently causing some practical concem is the observed tendency of noble-metal fission products (Mo, Ru, T¢, Te, and Nb) to deposit on graphite surfaces exposed to fissioning fuel in the MSRE. Some of the iso- topes of these elements have neutron cross sec- tions high enough to affect significantly the neutron economy of an MSBR after several years of operation if a large fraction of these isotopes deposited in the graphite core. Recent work in the MSRE has been directed mainly toward eluci- dating the behavior of these noble-metal fission products. Information derived from several kinds of tests is reported in some detail in the following sec- tions. These include (1) quantitative measure- ments of the concentration of fission product species in samples of the MSRE pump bow] cover gas captured during a number of diverse reactor operating conditions, (2) analysis of fuel samples for fission products, (3) examinations and analyses of graphite and Hastelloy N specimens exposed to molten fissioning fuel salt in the MSRE for 24,000 Mwhr of power operation, and (4) qualitative meas- urements of fission product deposition on metal and on one set of graphite samples exposed to the cover gas and fuel phases in the MSRE pump bowl under a variety of reactor operating conditions. 9.1 FISSION PRODUCTS IN MSRE COVER GAS Sampling of the cover gas in the MSRE pump bowl has been continued in an effort to define the nature and the quantity of the gas-borne species. F. 116 F. Blankenship Five gas samples were obtained and analyzed during the period covered by this report. In these studies we attempted to determine how the nature and amount of gas-borne species are affected by (1) addition of beryllium to the fuel, (2) stopping . the MSRE fuel pump, (3) a long reactor shutdown, and (4) increasing the volume of helium bubbles in the fuel. All samples were taken in evacuated 20-cc capsules sealed by fusible plugs of 2LiF - BeF ; these plugs melted and permitted the capsules to fill with the cover gas upon insertion into the heated pump bowl. Samplers, sampling procedures, and analytical operations were, in each case, very similar to those previously described.! Since it was not possible to instrument the sampling assemblies, no definitive information concerning the temperature of the gas at time of sampling is available. Temperatures wete cer- tainly higher than the melting point of the fusible plug (~450°C). It seems likely that the tempera- ture is near 600°C and that the capsule actually drew 20 cc of gas at this temperature and 5 psig; the sampled volume of gas at STP was, therefore, about 8.5 cc. The results obtained from these five samples are shown in Table 9.1. All values (except those for uranium) are given in disintegrations per minute for the total sample, and all have been corrected to correspond to the time of sampling if the reactor was operating, or to time of shut- down for the two cases where the reactor was not at power. The uranium values are in micro- grams of uranium per sample. Conditions of re- actor operation and special features of each test are indicated in Table 9.1. 1s. s, Kirslis, MSR Program Semiann. Progr. Rept. Aug. 31, 1966, ORNL-4037, p. 165. 117 Table 9.1. Fission Products in MSRE Pump Bow! Gas as Determined from Freeze Valve Capsules Experiment No. FP11-42 FP11-46 FP11-53 FP12-7 FP12-26 Sampling date 4/11/67, 02:49 4/18/67, 02:28 5/2/67, 10:43 6/2/67, 06:50 7/17/67, 06:03 Operating time, days On 65, off 1.5 hr Off 14, on 72 Off 14, on 86 On 92.3, off 42.5 Off 46, on 23 Nominal power, Mw 0 7.2 7.2 0 7.2 Be addition After 8.40 ¢ No No No After 37.8 g Features Pump off 1.2 hr Regular Helium bubbles Power off 42.5 days Regular Accumulated Mwhr 27,000 29,100 31,700 32,650 36,500 Isotope Yield Disintegrations per Minute in Total Sample %10 6.06 1.05 x 10! 2.31 x 10! 1.57 x 1011 2.74 % 1011 1030, 3.0 2.51 % 107 4.64 % 10° 1.12 x 10" 4 % 10°? 106p, 0.38 8.2 % 107 9.49 x% 107 4.03 % 108 1.7 x 10°® 132 ~4.7 1.15 x 10! 3.35 % 101! 1.88 % 101 4.17 % 107 3.16 x 10'7? 12%pg 0.35 7.98 x 10% 3.51 x 107 2.17 x 10 6.6 x 103 5NL 6.2 6.45 % 108 1.3 x 107 1.05 % 161° 2.26 % 1092 3.52 x 10° 957, 6.2 <4.4 =% 107 ~2 % 107 1.8 % 108 8.64 % 107 2.98 % 10”7 1405, 6.35 6.16 x 10° 1317 ~3.1 5.7 % 10° 9.81 x 108 8.63 % 10° 1.67 x 101'° 895y 4.79 2.13 x 108 4.08 % 107 3.72 % 10° 8.35 x 105 3.71 x 10° 111 8 . 8 Ag 0.019 5.5 x 10 1.33 < 10 1410, ~6.0 4.79 x 10° 14404 ~6.0 4.17 x 107 2350 (ue 59 9.2 23 25 The considerable scatter in the data for ?°Zr probably reflects scatter in the quantity of salt mist in the sampling region and trapped in the gas samples. The largest value (1.8 x 103 dis/ min of ?3Zr) was found under conditions in which helium bubbles in the fuel were at a maximum. If this ?°Zr were present as fuel mist, it would represent about 1.3 x 1072 g of salt per sample or about 1.6 x 10™* g of salt per cubic centimeter of helium in the sampling region. In all other cases studied the ?*Zr, and presumably the mist, was two- to tenfold more dilute, The values for 50.5-day ®°Sr, which is born from 3.2-min 8%Kr, afford an opportunity to check whether the mist shield which encloses the sam- pling station permits equilibrium mixing with the rest of the gas in the pump bowl. The values for 8951 are nearly an order of magnitude higher in every case than are those for *3Z1. A small fraction of the ®°Sr probably arises from salt mist; it seems vittually certain that most of it arises in the vapor phase through decay of the 89Kr. In the three gas samples taken while MSRE was at power, the 8%Sr activity in the samplers averaged 3.8 x 10° dis/min. The total number of #7Sr atoms per standard cubic centimeter of gas deduced from calculating the 29Sr counting rate back to the time of sam- pling was, therefore, 4,7 x 1013, If the sample as taken is assumed to represent a uniform mix- ture of the pump bowl gas containiag **Kr and its daughters ®°Rb and #°Sr formed since the B9Kr left the salt, then the ®9Kr was being stripped at 2 x 1017 atoms/min or at some 31% of its production rate. If, on the other hand, it is assumed that all the ®*°Rb and ®°Sr are washed back to the salt phase by the spray from the spray ring, then the quantity of ®*°Kr that must have been stripped is more nearly 50% of the production rate, With the simplifying assumptions that (1) no 89Kr is lost to the moderator graphite, (2) no 89Kr is lost through burnup, and (3) the fuel which flows through the pump bowl is stripped of 89Kr with 100% efficiency, it may be simply shown that 89Kt stripped/min F 89Kr produced/min_ A+ F where F is that fraction of the fuel volume which passes the stripper (pump bowl) per minute and A is the decay constant (0.693/3.2 min) for 8?Kr. For MSRE, with the pump bowl flow rate at 4% of total flow, this expression yields the value 29% for the 89Kr lost to the stripper gas. It is un- likely that the stripping efficiency in the pump bow! is 100%; moreover, it is certain that some fraction of the 8?Kr is lost by penetration into the graphite. This rate of loss of 8°Kr to the moderator — which is probably less than 30% of the production rate — will in effect introduce a third tern to the denominator of the equation and lower the fraction lost to the off-gas. From these data, therefore, it seems possible that the gas at the sampling station may be more concentrated in 29Sr than is the average gas in the pump bowl, but the concentration factor is not large. The absolute amounts of ?°Mo, corrected back to sampling time or time of previous shutdown, show rather minor variations with reactor op- erating conditions. The lowest value is for run F'P11-42, in which the sample was taken 1.5 hr after shutdown and 1.2 hr after stopping the fuel pump. The normal runs showed the higher ?°Mo concentrations, although the highest value might have been expected for FP11-53, in which the off- gas pressure was suddenly lowered just before sampling to ensure a release of helium bubbles from the pressurized graphite bars into the cir- culating fuel. From the amount of ?°Mo found in these samples (mean value 1.9 x 10'! dis/min or 2.2 x 1019 dis min~! ce~! or 1.3 x 10'* atoms ?9Mo/cc), the effective partial pressure of the volatile species is calculated to be 4 x 10~ ° atm. The total ??Mo lost at 4.2 liters/min of helium flow is 7.8 x 1029 atoms/day. This is ahout 18% of the total inventory of 4.4 x 102! atoms, or about 70% of the daily production rate. If it is assumed that the material is lost as MoFfi, 118 this corresponds to a loss of 4.6 x 102! fluorine atoms/day, or an equivalent of F~ per 150 days. The very considerable quantities of ®?Mo found on the graphite and the Hastelloy surveillance specimens as well as in the fuel stream do not seem consistent with such large losses to the gas system. The 193Ru, 1%°Ru, 129Te, and '3?Te concen- trations generally showed parallel behavior over all the runs, with particularly good correspondence between the 23Ry and '9%Ry values. The ratios for the two isotopic pairs agreed satisfactorily with the ratios of fission yields divided by half- lives. stoppage was minor on this set of isotopes. The highest values for '93Ruy, '9%Ru, and '?%Te were obtained during the pressure release run, FP11-53. However, 132Te, like ??Mo, did not show a high value during this particular test. The effect of short shutdown and pump The appreciable concentrations of #°Sr found in samples FP11-42 and FP12-7 are puzzling. In each case the reactor had been shut down long enough to permit the complete decay of 89Kr be- fore the sample was taken. The ?°Zr analyses indicated fuel mist concentrations too low to account for the 3°Sr values. It may be that other activities interfered with the beta counting re- quired for 3%Sr analysis. If these values are accepted as real (and further sampling must be done to confirm them) it would appear that, at least when the pump is off or the reactor is not at power, the sampling volume is poorly flushed by the cover gas system. The ?°Nb values showed a fivefold increase of magnitude from the first run to the fifth, with a marked peak at the pressure release run, FP11- 53. Not all of the overall increase could be as- cribed to the increase of °Nb inventory in the reactor system with continued reactor operation (cf. moderate increases in !°3Ru and '?°Ru). The analyses for ?3°U from four of these runs yielded values higher by at least a factor of 10 than those observed in earlier tests.! These relatively high values for 233U are, at best, dif- ficult to explain. If, as noted above, the values for °°Zr are taken to indicate the existence, and quantity, of salt mist in the sample, then sample FP11-53 might have been expected to have about 20 pg of 35U in such mist. However, in no other case of those shown in Table 9.1 can such an ex- planation account for more than 20% of the observed uranium. The losses indicated by these samples are ap- preciable. The mean of four samples shows nearly 30 pg/sample or about 3.5 pg per cubic centimeter of helium. This would correspond to somewhat less than 20 g/day of 275U or nearly 60 g/day of the uranium in the fuel. (The figure remains at nearly one-half this level if the highest figure is rejected,) It seems absolutely certain that the reactivity balances on the reactor through 250 days of full-power operation would have speedily disclosed losses of far smallet magnitude than these. It is difficult to conceive of a mechanism other than volatile UF to account for losses of uranium to the vapor. It is, however, extremely difficult to see how losses of this sort could be reconciled with the nuclear behavior (and the chemical analyses) of MSRE. IFurther samples of the gas system will be made to help resolve this apparent dilemma. The effect of short shutdown and pump stoppage (FP11-42) on noble-metal volatilization was in- significant. This suggests that the volatilization process does not depend on fuel nor bubble cir- culation nor on the fissioning process itself, but is probably a local phenomenon taking place at the fuel-salt—gas interface. Beryllium additions to the fuel had no discernible effect on the volatilization of noble-metal fission products. This fact is rein- forced in some detai! (see subsequent sections of this chapter) by data obtained by insertion of getler wires into the pump bow! vapor space. The effect of reactor drain and long shutdown (FP12-7) was consistently to lower observed noble-metal volatilization (observed activities were, of course, calculated back to the time of previous shutdown). This observation is con- sistent with all previous data on the effect of long shutdowns. Presumably an appreciable fraction of the noble metals is deposited on the walls of the drain tank and reactor system during a long shutdown. The minor effect of short shut- down and the incomplete deposition during long shutdown seem to indicate a slow deposition process, The pressure release experiment (FP11-53) re- sulted in an appreciable increase of volatilization of most fission products but not of %Mo or 32 Te, Niobium-95 showed the most distinct increase. It is particularly puzzling that the '?°Te concentra- tion increased markedly while the '3?Te concen- tration fell slightly. Clearly, further experiments 119 of this kind would be required to permit definite interpretations. It cannot be claimed that the mechanism for volatilization of many of these wmaterials is known. As has been shown elsewhere! ? the volatile flu- orides of many of these materials (notably Mo, Ru, Te, and, probably, Nb) are too unstable to exist as equilibrium components of a fuel system con- taining appreciable quantities of UF . It is true that, for example, Mo formed at about 5 x 10%? atoms/day probably exceeds greatly the solubility limit for this material in the salt. Since it is born in the salt in an atomic dispersion, it may, in ef- fect, behave as a supersaturated solution; that is, the activity of molybdenum may, in fact, be mark- edly greater than unity. It seems most unlikely that it can have activities greater than 1019, which would be required if MoF . were to be stable. Such findings as the presence of 111Ag, for which it is difficult to imagine volatile fluorides, seem to render the volatile {luoride mechanism more un- likely. It still seems more likely that an ex- planation based upon the finely divided metallic state will prove true, though such explanations also have their drawbacks. 9.2 FISSION PRODUCTS IN MSRE FUEL The program of sampling and analysis of samples of MSRE tuel for fission product species has been continued by use of apparatus and techniques de- scribed previously.' =3 Table 9.2 shows data for 16 fission product isotopes and for 23°Np obtained from a series of four samples from a long stable run of MSRE, from a sample obtained shortly after shutdown in this run, a sample after the salt had been cooled for 42.5 days, and a sample from a relatively early stage in the subsequent power run. Captions in this table show the operating history of MSRE and the special conditions prevailing in the reactor at the time the sample was taken. Table 9.2 also includes estimates of the fraction of the isotope represented by these analyses in the 4860 kg of circulating fuel, 23, 8. Kirslis and F. F, Blankenship, Reactor Chem. Div. Ann. Progr. Rept. Dec. 31, 1966, ORNL-4078, p. 48, 33. 8. Kirslis and F. F. Blankenship, MSR Program Semiann. Progr. Rept. Feh. 28, 1967, ORNL-4[19, p. 124, Table 9.2. Fission Products in Fuel Samples Sample No. FP11-45 FP11-51 FP11-52 FP11-34 FP11-58 FP12-6 FP12-27 Accumuiated Mwhr 29,000 30,800 31,250 32,000 32,650 32,650 36,550 Mominal power, Mw 7.2 7.2 72 7.2 0 0 7.2 Operating time, days® Oif 14, an 71 Off 14, on 82.5 Off 14, on 85 Off 14, on 89 On 92.3, off 9.1 0n 92.3, off 42.5 Off 46, on 23.5 Exposure time, min 1 1 1 2 i P 6 Features No purge,” after 8.4 g Minimum leve! Regular Regular Pump off 2 hr, Before startup. After 38 g of of Be added in pump bowi power off 2.3 hr after shutdown Be udded Fission o Percent S Pe;;rI{ Percent Percent Percent Per.cent Percent Isotope Yield dis min~' g~ ! of dis min™! g7 ! of dis min~! g ! of dis min—!' g ! of dis min=! g~ ! of dis min' ! g7! of dis min~! ¢! of (%) Total® Total® Total€© Total® Total® Total® Total® 67-hr *Mo 6.06 1.53 - 194! 108 7.22 . 1010 5% 8.17 . 1910 58 3.50 < 1010 25 3.19 . 10'° 23 Decaved 1.24 . 104! 88 39,7-day '%?Ru 3.0 1.40 » 1010 23 3.8 .10° 7.1 8.86 . 1010 165 1.82 . 10° 3 9.5 . 10° i7 6.9 . 10° 12 5.55 . 10° 23 1.02-year '?°Ru 0.38 ~4 . 108 ~3 . 108 2.38 . 108 7.5t . 107 3.6 « 10® 3.9 . 108 2.43 . 10% 78-hr 132Te ~ 4,7 3.50 . 10!° 32 1.68 « 101° 15 1.56 - 10° 14 1.07 . 10!° 10 8.39 . 10¢ 8 Decayed 1.27 . 1910 12 37-day '¥7Te 0.35 5.20 . 10% 3 5,00 - 10° 8 5.45 - 10% 8 2.06 -« 10° 3.1 .108 i0 35-day *SNb 6.2 2.61 . 101° 4.0 . 10° 1.91 ~ 10° 311 1010 2.15 . 1010 8.04 . 101° 1.91 . 10'° 65-day *%Zr 6.2 2.97 . 10192 1.20 . 10%! 1.34 . 101 .34 . 101! 1.54 . 101! 1.46 . 101! 1.00 . 107¢ 12.8-day '*°Ba 5.35 1.67 . 10! 115 1.82 . 10" 125 2,17 .0t 149 1.74 . 101! 119 50.5-day *°Sr 4.79 8.6 . 1010 123 8.02 . 19'° 106 8.11 . 101° 106 1.02 101! 128 8.89 . 10%° 111 6.07 - 101 116 9.7-hr °'Sc 5.81 1.29 . 10"! 95 33-hr '*2Ce 5.7 1.14 . 101! 86 7.67 . 10'° 53 1.78 . 101! i34 33-day *!Ce ~6,0 1.48 « 10'! 129 1.4 . 101! 142 1.71 « 10! 143 1.37 . 101! 115 260-day '*4Ce ~6.0 6.36 « 10'° 6.05 « 101? 6.24 « 10'° 6.01 . 10'0 7.6-day '''Ag 0.019 5.1 x 107 12 6.5 - 107 15 4.9 . 107 11 1.63 - 107 4 3hr P 2Ag 0,019 8.0 « 107 34 8.9 . 107 38 8.05-cay '?'1 -3 9.17 - 1017 128 8.20 - 10'° 114 8,22 . 12 ° 115 6.97 - 10'° 97 1.15 « 10%° 16 7.12 . 10t° 114 5.6 - 107 ! 3.29 « 1017 2.33-day **°Np “Duration of previous shutdown and of continuous operating time just before sampie was taken; vice versa 5.36 - bUsually a 200-cc/min gas purge was passed down the sample line, ¢Percent of calculated total amount in reactor based on fission yields and power history. for samples taker after shutdown. 0c1 121 As in previous analyses of the reactor fuel, it is clear that a very high fraction of isotopes (such as '4%Ba, 89Sr, 1%1Ce, 1*%Ce, and °!Sr) that form very stable and soluble fluorides is present in the circulating fuel. As judged by the behavior of 13171 the isotopes of iodine also seem to be present in a high fraction. On the other hand the more noble elements, such as Mo, Nb, Te, and Ru, are generally present in much lower fractions. It is worthy of note that these materials seem to show a much wider scatter than do the ‘‘salt seekers’’; this may indicate that they are not present in true solution. It must also be noted that these noble metals are present in considerable quantity in the gas stream and that they adhere tenaciously to metal specimens ex- posed to them. It is possible, therefore, that a considerable fraction of the material found in the “‘salt’’ is due to the fact that the sampler is low- ered and then raised through this gas phase. It is possible that the relatively high values obtained for %Mo, 19%Ru, and 132Te in FP11-45, where no gas was used, may be due to this phenomenon. This possibility will be verified, if possible, ex- perimentally in the near future. 6.3 EXAMINATION OF MSRE SURVEILLANCE SPECIMENS AFTER 24,000 Mwhr A second set of long-term surveillance speci- mens of MSRE graphite and Hastelloy N was ex- amined after exposure to citculating molten-salt fuel in an axial core position of the MSRE for 24,000 Mwhr of power operation. During the last 92 days of this exposure, the MSRE operated at a steady power of 7.2 Mw essentially without in- terruption, whereas power operation was frequently interrupted during the previous 7800-Mwhr expo- sure. Examinations of the specimens gave results very much like those previously reported after a 7800-Mwhr exposure.! ™3 Examination of Graphite As before, three rectangular graphite bars were available for examination: a 9.5 x 0.47 x 0.66 in. bar from the middle of the core, and 4.5 x 0,47 « (.66 in. bars from the bottom (inlet} and top (out- let) of the core. Visually, the graphite appeared undamaged except for occasional bruises incurred during the dismantling. The two 4.5-in.-long bars gained 0.002 t 0.002 in. in length and 13 mg (0.03%) in weight. Metallographic examination showed no radiation or chemical alteration of the graphite structure and no evidence of surface films. X radiography of thin transverse slices showed occasional salt penetration into previously exist- ing cracks that happened to extend to the surface of the sample. Similar penetration was observed in control samples that were exposed to molten salt but were not irradiated. X-ray diffraction by the graphite surface exposed to fuel gave the normal graphite lines except for a very slightly expanded lattice spacing. A few weak foreign lines that were probably due to fuel salt were observed. Autoradiography of the samples showed a high concentration of activity within 10 mils of the surface, with diffuse irregular penetration to the center of the cross section. These observa- tions confirmed previous demonstration of the satisfactory compatibility of graphite with fis- sioning molten salt as far as damage and per- meation by fuel are concerned. Concentration profiles for fission products in the graphite were determined by milling off layers, which were dissolved and analyzed radiochemically. Near the surface, layets as thin as 1 mil were ob- tained; farther in, layers as thick as 10 mils were collected until a total depth of about 50 mils was reached. The predominant activities found were molybdenum, tellurium, ruthenium, and nicobium. These elements can be classed as noble ele- ments since their fluorides (which are generally volatile) are relatively unstable, The practical concern, because of long-term neutron economy, is with Mo, *"Mo, °9Tc, and '°'Ru, but these particular isotopes are either stable or long-lived and thus were not amenable to direct analysis. However, we considered as sound the assumption that their deposition behavior was indicated by either that of a radicactive isotope of the same element or that of a radioactive noble-metal pre- cursor with a suitable haif-life. Over 99% of the noble-metal activities were en- countered within the first 2 or 3 mils of the graphite surface; the same was true for °Zr. This is illustrated in Figs. 9.1-9.4. The values for blanks shown in the figures were obtained from samples of about 0.2-g (1 mil) size that were milled from fresh unexposed graphite in the hot cell and with the equipment used for the irradiated specimens. The blanks show that the apparent tails on the curves for Mo, ?3Nb, and '°°Ru are 1044 109 4 - Q - n disintegrafions per minute per grum 2 1010 5 2 10° L 0 10 20 30 40 50 DISTANCE FROM SURFACE (mils) Fig. 9.1. Distribution Profile of %0 in Grephite from MSRE Core. meaningless. Figure 9.4 indicates that there probably was measurable penetration of ?5Zr to a depth of 50 mils. Figure 9.5 shows that 132Te concentrates, as do the noble metals, at the surface; however, a measurable concentration of this element pene- trates to a depth of 50 mils. The behavior of fission products with rare-gas precursors is shown in Figs. 9.6 (1*%Ba) and 9.7 (3°Sr). Because of the concentration at the surface, the amounts present in the several layers near the surface were representative of the total 122 '3 ORNL--DWE 67—12232 10 e — } - . T Lo ] 5 | 2 ] ol ] 5 5 2 o 2 o't _; o e 5 R c E °r '"f,'j @ _ o E 2 5 £10° — £ 5 ) 5 ....... 2 10° 5 2 L 108 L ] 0 10 20 20 40 50 DISTANCE FROM SURFACE (mils) Fig. 9.2. Distribution Profile of 95Nb in Grophite from MSKRE Core. amount of many isotopes in the graphite. Table 9.3 shows the quantity of several materials (in disintegrations per minute per square centimeter) on graphite specimens as functions of position and of exposed face. In each case, face No. 3 was in contact with other graphite and was not exposed to salt. In general, this face contained about as much activity as those exposed to salt; this finding supports the view that the deposited activities were gas boine. In Table 9.4 the values (averaged for all faces of the graphites) obtained in the specimens ex- posed for 24,000 Mwhr are compared with the 123 " ORNL—DWG §7--12233 1010 H— o — * BLANK 5 4~ —_— —_ . o w N w n o ® disintegrotions per mingie per gram 0 10 20 30 40 50 DISTANCE FROM SURFACE (mils) Fig. 9.3, Distribution Profile of 106Ru in Graphite from MSRE Core. values previously obtained after 7800 Mwhr. It is clear from this table that the data for °?Mo, 132Te, and '°?Ru are generally similar, while the values obtained for °*Nb and °°Zr are ap- preciably higher for the longer exposures. It should be noted that the data for 3Nb are ob- tained from gamma scans without chemical sep- arations and are probably less certain than the others. The graphite samples accumulated by milling these bars were analyzed for uranium by neutron activation. Typical data from one of these bars disintegrations per minute per gram 10 ——— -y 12 ORNIL--DWG 6712234 0 10 20 20 40 50 DISTANCE FROM SURFACE (mils) Fig. 9.4. Distribution Profiie of 951r in Graphite from MSRE Core, are shown as Table 9.5, It is clear from this table that the 23°U is present in all faces of the graphite and that the concentration diminishes with the depth of milling. The small concentra- tion in the blank (as in blanks from other speci- mens) indicates strongly that these uranium con- centrations are real. The absolute quantity of 233U or of uranium in the graphite moderator is apparently very small. If these analyses are, in fact, typical of those in the 2 x 2 in. moderator bars, then the average concentration of ?3°U in and through the bars 124 ORNL—-DOWG ©67-12235 1013 disintegrations per minute per gram 20 30 40 DISTANCE FROM SURFACE (mils) 50 Fig. 9.5. Distribution Profile of 132Te in Graphite from MSRE Core. is much less than 1 ppm. Since the MSRE con- tains about 70 ft® or slightly less than 4000 kg of graphite, the total 2?°U in the moderator stack is almost certainly less than 4 g. Since the MSRE fuel is slightly more than 30% enriched in 23°U, the total uranium in the moderator should be, at most, 10 to 15 g. Examination of Hastelloy N The graphite specimens described above were, as was the case for the specimens examined after 7800 Mwhr, contained in a cylindrical, perforated holder of Hastelloy N. Samples were cut from this cylinder as before! =3 and analyzed by standard techniques for several fission products and for 2357y, 10"2 _ N O‘ o w disintzgrations per minute per grom n 102 |- DISTANCE FROM SURFACE (mils) Fig. 9.6. Distribution Profile of 1404, in Graphite from MSRE Ceore. Data are shown in Table 9.6 for seven fission product isotopes and for uranium at five positions from the top to the bottom of this Hastelloy N as- sembly, Data are also shown for the comparable samples examined after 7800 Mwhr of operation in mid-1966. It is clear from these data that the behavior of ?9Mo was similar in the two sets of specimens, with the 24,000-hr exposures generally showing somewhat more deposited activity. Behavior of the !*2Te seems generally similar in the two ex- posures. Somewhat more '°3Ru was found on the 24,000-Mwhr specimens than on those previously examined, while ?°Zr (present in small amounts, in any case) seems less concentrated on the specimens exposed to the higher dosage. lodine- 131, which is present in appreciable concentra- tions, shows some scatter with location of the 1o" N O-h- o disintegrations per minute per gram DISTANCE FROM SURFACE (mils) ORNL—DWG &7--12237 Fig. 9.7. Diswribution Profile of B¢ in Graphite from MSRE Core. 125 specimen but about the same deposited activity in the two series of tests. The uranium values correspond in every case to negligible quantities. It appears that, if these analyses are typical of all metal surfaces in the system, about 1 g of 233U may be deposited on or in the Hastelloy of the MSRE. Fission Product Distribution in MSRE The computer code for calculation of total in- ventory of many fission product nuclides in the operating MSRE has not yet been completed. Ac- cordingly, no detailed information as to expected quantities of several isotopes is available. How- ever, from the data given in previous sections of this chapter approximate values can be given for the distribution of several important fisston prod- ucts among the salt, graphite, and metal phases of the reactor system. These are presented in Table 9.7. Such approximate values must, of course, assume that the fission products found on the relatively small samples from the surveil- lance specimens in midcore are those to be found on all graphite and metal in MSRE, Since there is, as yet, no test of this assumption we regard the values in Table 9.7 as tentative, Table 9.3. Fission Product Deposition on MSRE Graphite Faces After 24,000 Mwhr Fission Product (dis min~! cm"g) Graphite Location Face 66-hr 78-hr 37.6-day 1.02-year 35:day 65-day 99M0 132Te IDSRH IOGRU QaNb gSZr x 1010 x 1010 x 10° x 108 x 10'° x 108 Top 1 2.16 1.45 4.54 2.26 5.00 2.26 2 3.59 1.75 8.86 3.58 6.15 3.58 3 2.76 1.74 6.97 2.77 4.77 2.77 4 6.30 3.52 15.9 3.08 g9.16 3.08 Middle 1 2.28 2.51 6.92 3.51 2.00 3.51 2 5.32 3.47 12.9 6.33 17.8 6.33 3 6.16 3.31 7.94 7.94 4 5.41 4.68 15.9 6.98 15.4 6.98 Bottom 1 3.24 1.79 6.95 3.11 11.6 3.11 2 8.05 4.50 18.6 8.04 30.6 8.04 3 2.85 2.22 9,72 3.64 14.2 3.64 4 3.43 2.14 8.65 3.74 13.3 3.74 126 Table 9.4. Fission Product Deposition on MSRE Graphite After 7800 and 24,000 Mwhr Fission Product (dis min " ? cm“'z)a Graphite 1.02-year . 67-hr Mo 78-hr **?Te 39.7-day '?3Ry 106 35-day °°Nb 65-day 252Zr Location _ _—~ 7 7 o T ) Rn ™™™ = =" b ] A B A B A B A B A R A B x 101 x10'% x10'% x10'? x10!'0 %1010 « 1019 w10'% x10'% x10'% x10!'® Top 3.97 3.69 3.22 2.12 0.83 0.91 0.003 0.46 6.2 0.0004 0.001 Middle 5.14 4.79 3.26 3.82 0.75 1.19 0.006 2.28 1.17 0.003 0.007 Bottom 3.42 4.39 2.78 2.66 0.48 1.10 0.005 2.40 17.4 0.002 0.02 IMean values for all four faces. b A: value obtained after 7800 Mwhr, B: value obtained after 24,000 Mwhr. Tahle 9.5. Uranium in Millings from Top Graphite Bar in MSRE After 24,000 Mwhr Cumulative Specimen 235y 235y Samplle Sample Depth Weight Analysis in Specimen IL.ocation Number (mils) () (ppm) (1) Side 1 1 1.56 0.145 9 1.3 (facing salt) 5 4.04 0.224 2.1 0.47 7 5.51 0.132 2.4 0.32 9 7.50 0.172 0.80 0.13 11 10.75 £.2065 0.03° 0.0099 13 13.60 0.256 .72 0.19 17 17.40 0.340 (.48 0.16 21 21.75 .390 0.29 0.11 25 25.64 0.350 0.49 0.17 26 30.57 0.445 0.39 0.017 - 27 36.58 .540 .28 0.15 89 48.18 1.05 0.15 0.15 A Side 2 5.58 0.360 8. 3.2 : (facing salt) 8.60 0.100 3.0 0.57 11.84 0.210 3.6 0.76 10 14.03 0.140 2.6 0.36 12 16.62 0.165 1.4 0.23 14 24.44 0.500 1.6 0.80 Blank 0.28 18 31.90 0.480 0.99 0.48 22 40.94 0.580 1.33 0.77 Side 3 3 5.50 9.495 6. 3.3 (facing 15 8.79 0.300 1.2 0.36 graphite) 19 12.98 0.470 0.84 0.39 23 18.56 0.505 0.78 0.39 “Questionable value. Table 9.6, Fission Product Depo Isotope Extreme Top Bb All values except 235¢7 in disintegrations per B sition on Hastelloy N in MSRE Core A minute per sguare centim Hastelloy N Location® A Middle B fter 7800 and After 24,000 Mwhr eter of Hastelloy Extreme Bottom B 67-hr Mo .84 x 101 3.12 (10t 3.94xtott 2.76 x 1017 2.86 x 10! 2.04 x 10%? 4.42 x 1017 0.74 % 10%? 78-hr 102 Te 6.66 x 10%} 5.08 x 101 5.66 x 101* 3.41 x 10 2.06 x 10%} 4.07 x 10*1 402 % 10!t 1.94 x 101! 37.6-day ‘" Ru 0.09 x 1010 3.55X% 1010 10.6 x 1017 9.55 x 100 1.56 X 1010 2.32x 100 602X 101? 35-day J°Nb 1.04 x 101? 2.32 x 101 .32 x 10** 1.66 x 10%* 0.16 x 10*} 65-day °Zr 2.32 x 10° 18.5 x 10° 1.56 x 10° 18.4 x 10° 8.62 x 10° 25.8 x 103 4.10 x 108 0.32 x 103 50.5-day > Sr 1.69 % 16° 10.3 x 108 46.1 x 10° 5.93 x 10° 0.834 x 108 8.05-day ‘"1 1.60 x 10° 8.24 x 10° 3,93 x 10° 3.97 x 107 3,18 x 10° 5.24 % 10° 7.67 x 10° 235y (ug/cm® $1.3 0.92 0.61 0.61 LTt 1.51 aa11 surveillance specimens in center axis of core. ba. after 7800 Mwhr, B: after 24,000 Mwhr. 128 Table 9.7. Approximate Fission Product Distributions? that had been exposed to the liquid phase, the in MSRE After 24,000 Mwhr gas phase, or the interface region. In all cases, e — e the nickel-plated key exposed at the very top of Fuel Graphite Hastelloy N the gas phase was analyzed, but these results are Isotope {(total (total (total not included here. Data from seven separate tests dis/min)} dis/min) dig/min) are shown as Table 9.8. In addition, the behavior ..... i of '°®Ru is shown in Fig. 9.8. In virtually every case the specimens were 5 leached first with an aqueous chelating agent and 1320a 5.3 % 1018 5.9 x 10%° 4.1 x 1017 then with an oxidizing acid solution. Some iodine and ruthenium may have been lost from the acid b %Mo 1.7 x 1017 8.6 x 1016 3.2 x 10%7 b %Ry 2.9x 10 2.2« 101° 5% 1018 leach solutions, and it is not unlikely that some of these values are low. Table 9.8 reports, in all Wz, 7.3 x 10'7 1.8 x 1014 3.6 x 10t? cases, the sum of values obtained with both types - b 95Nb 9.6 % 1010 1.6 x 1017 1.5 x 1017 of leach solution. A nuinber of conclusions from the data, of which those in Table 9.8 are typical, follow. T T ' The deposition of noble metals on the pump bowl 8%, 4.9 % 1017 1.3 x 1013 131y 3.5 x 1047 4% 101° Selected mean value from fuel analyses; assume con- specimens under normal reactor operating condi- centrations on graphite and fuel specimens are typical of tions was quantitatively quite similar to that pre- all such surface in core. ) 3 BMean of considerably scattered values. viously reported.® Added purge flow down the ORNL-0OWG 67-12238 9.4 DEPOSITION OF FIS3ION PRODUCTS tinuing series of studies of deposition of gas- SSI bomne activities on metal specimens. The tests carried out since the previous report have attempted to observe the effect of duration of previous reactor operation, addition of metallic beryllium, level of liquid in the pump bowl, exposure time of speci- mens, stopping of the reactor pump, length of shut- down time of the reactor, and draining of fuel from the reactor. Exposure conditions were generally similar to those used in previous tests of this kind,! ~3 but in these studies the stainless steel cables which hold the sampler were used without additional getter materials. To reduce the increasing fis- sion product contamination of the sampler-enricher assembly, a downward flow of some 200 cc of helium was used with all insertions after FP11-45. The cables (in all cases except FP11-50, de- scribed in some detail in the following section) attached to standard samplers for obtaining spec- = imens of fuel were exposed under conditions and FP NUMBER for times shown in Table 9.2, The cables were then segmented into approximately equal lengths Fig. 9.8. Deposition of 106y in Pump Bowl Tests. FROM MSRE GAS STREAM ON ol e e METAL SPECIMENS o sl dpm/g =+ ,,J T e SSL l Access to the pump bow! provided by the ~ o SS6 R sampler-enricher tube has been used in a con- g Y ( toia! disintegrations per minute 107 12-27 ,[ -2z | 11-45 | 1450 5 #4-52 - 1-54 12-6 - 129 Table 9.8. Deposition of Yarious Fission Products on Metal Specimens lnserted in MSRE Pump Bowl 235 Exposure 0 a 09 131 103 132 95 Run No. Condition Mo I Ru Te Zr (lug/sarnple) = 1011 < 1010 x 1019 x 1011 x 107 FP11-45 Liquid 0.65 0.46 0.2 0.9 0.035 21 Interface 1.8 1.0 1.4 1.3 10 20 Gas 1.1 1.1 0.2 0.35 0.80 5.2 FP11-50" Liquid 3.5 12 0.7 14 250 35 Interface 26 30 2.8 100 320 190 Gas 0.2 9.2 23 17 60 FP11-51 Liquid 0.3 0.1 0.13 0.14 1 3.5 Interface 0.85 0.85 0.5 0.52 1.5 19 (Gas 0.4 0.6 0.25 0.3 3 6.0 FP1i-54 L.iguid 0.42 0,22 0.25 0.17 b 6.4 Interface Q.72 0.70 1.0 0.61 3.5 5.6 Gas 0.42 0.55 0.40 G.28 .5 2.0 FP11-58 Liquid 2.05 0.11 0.13 0.17 0.3 8.4 Interface 0.15 0.28 0.15 0.07 0.6 3.3 Gas 0.18 0.60 0.23 0.04 0.7 2.5 FPr12-6 Liquid 0.11 0.82 22 Interface 0.17 0.60 46 Gas 0.38 0.30 a0 FPri2.27 Liguid 0.06 0.14 0.05 0.09 0.9 8.2 Interface 0.72 0.24 0.04 0.18 1. 5.6 Gas 0.23 0.37 0.22 0.40 1.3 6.6 ASampling conditions are those shown in Table 9.2. b . . o - . . . Special sample connecting two sets of graphite immersion samples immersed for 8 hr; experimental details are in the following section. sampler-enricher tube during exposures had no significant effect on deposition on fresh metal specimens. The coosiderable additional reactor operating time also apparently had little effect, except for possibly a slight decrease in °°Mo and 132T¢ deposition for the later runs. The deposition of noble metals was usually heavier on the interface sections of the stainless steel cable than on either the gas-phase or fuel- immersed sections. This suggests the presence of a sticky scum or froth containing noble metals on the surface of the fuel in the pump bowl. As previously observed?® and as verified by the results from sampling the MSRE gas stream, the reduction of the MSRE fuel with beryilium caused no discernible change in deposition behavior. This observation is difficult to reconcile with the hy- pothesis that deposition behavior is related to the presence of volatile noble-metal fluorides. The deposition of noble metals on the sections of stainless steel cable immersed in the fuel usually paralleled their concentrations in the {uel (see Table 9.2), The B-hr exposure (FP11-30) revealed curious differences in the deposition behavior of the various fission products. The deposition of 1??Te and **Zr was about 100 times that observed in 1- or 10-min exposures. The cotresponding factor was 30 or less for 14%Ba, 10 or less for ?°Mo, 3 or less for 1°?Ru and '°*Ru, and less than 1 for ¥5Nb. Usually, comparisons between runs have shown similar factors for all the noble-metal fis- sion products. This unusual observation indi- cates that the noble metals do not travel together {e.g., in metallic colloidal aggregates) but deposit individually by separate mechanisms. In all but one of these tests, the exposure times varied from 1 to 6 min. The amounts of noble metals deposited during these intervals showed no increase with exposure time (if anything a de- crease) and were not significantly different from the results of previous 10-min exposures. Even in the case of 132Te, which deposited heavily in the 8-hr exposure, no significant differences were ob- served for exposure times between 1 and 10 mia. This and related information will be discussed in more detail later. The deposition behaviors of 40-day !Y3Ru and 1.0-year '9%Ru were fairly closely parallel. The incomplete data for 33-day '2°Te also followed well the variations among runs of the 77-hr 132 Te deposition data. At production-decay equilibrium, the number of disintegrations per minute of each fission product should be proportional to its fis- sion yield, For the tellurium isotopes the ratio of activities deposited was about 20, while the ratio of fission yields is 13.4. The ratio of ac- tivities of the two isotopes in the fuel was also about 20. The discrepancy between 20 and 13.4 may be partly due to the fact that the 12°Te may not have reached equilibrium activity. In the case of the ruthenium isotopes the much larger discrep- ancy between the ratio of activities and the ratio of fission yields is very probably due to the fact that the one-year '°°Ru had not nearly reached equilibrium activity. Experiment FP11-51 was carried out with the pump bowl level as low as possible, with the ob- jective of introducing more than the usual amount of circulating helium bubbles into the fuel. Nommal deposition of noble metals was observed. There- fore, the fraction of circulating helinm bubbles has no effect on fission product deposition. Experiment FP11-58 was carried out after 92.3 days of virtually uninterrupted power operation, 2.3 hr after shutdown, and 2 hr after stopping the fuel pump. Under these conditions, the deposition of noble metals decreased by a factor of 2 to 9, most sharply for tellurium and least for ruthenium. The concentrations in the salt (see Table 9.2) re- mained nearly constant except for ruthenium, for which the data show a fivefold increase. r I'he rather moderate decreases in noble-metal deposition with the reactor shut down for 2.3 hr and particularly with the fuel pump off are dif- ficult to explain by the previously mentioned metal colloid theory of deposition. Any suspended species in the pump bowl cover gas should have been swept out by the 4-liter/min helium flow through the pump bowl gas space. The 3-in.-diam by 6-in. volume inside the mist shield was addi- tionally swept by the 200-cc/min purge flow prior to and during the exposure. The result is one of the most conclusive indications that tiuly gaseous species originating from the fuel suriace are responsible for the observed concentrations of noble-metal fission products in the pump bowl gas space. Experiment FP12-6 was carried out after a re- actor drain and refill after 42.5 days of reactor shutdown and prior to resumption of power op- eration. Only slightly lower than normal depo- sition of noble metals was observed. After long shutdowns in the past, sharp decreases in depo- sition occurred. Also contrary to past experience, the concentrations of noble metals in the fuel either remained constant (ruthenium) or increased (tellurium and niobium). An additional experiment (FP14-1, not shown in the table) was carried out after the MSRE had been shut down for 38.1 days, then at full power for 2.5 days, then shut down and drained for 2.0 days. The fuel pump continued to circulate helium in the drained reactor with the temperature of the pump bowl top at 700°F. Specimens were exposed for 11 min. The only specimens available for analysis were the gas-exposed stainless steel cable and the key. The analyses are not complete, but the avail- able results are remarkable. Compared with the previous ““normal’’ run (FP12-27), the activity (calculated back to the end of the 2.5-day op- eration) deposited on the cable was slightly - higher for °°Mo, '9°Ru, and ?°Nb, distinctly higher for '3!'I and 235U, slightly lower for 132Te and '°3Ru, and distinctly lower for 957r. Most of the activities on the key dropped by a factor of about 10 from the previous run. These high depositions are all the more re- markable when it is considered that there was a two-day shutdown before sampling and that even the short-lived fission products could not have reached equilibrium activities in the 2.5- day period of power operation following the 38.1- day shutdown. It is not clear whether the observed deposits originated from material previously deposited on the graphite or metal walls of the reactor system or whether it came from the 40 1b of fuel salt which remains in the reactor system after a fuel drain. Some of this fuel salt is in the form of shallow puddles in the intemals of the fuel pump where contact with the circulating helium should be good. Another possibly pertinent fact is that gas circulation through the reactor and pump bowl with the pump operating in the drained reactor is much faster than whea the reactor is filled with fuel, Thus the pump bowl specimens may con- tact a larger volume of gas per unit time. The fuel may also act as a scrubbing fluid to remove gaseous or suspended activities from the pump bowl gas, even while it is the source of these activities at a different time and place. Sorbed gaseous activities or loosely deposited solids from the reactor surfaces may conceivably move to and stay in the gas phase more readily in the absence of fuel salt. A few analytical results are available from test FP14.-2, which was run one day after test FP14-1 and after the reactor had been refilled with fuel but before power operation was resumed. For the gas-phase stainless steel cable specimen, the deposition of %Mo, 132Te, and 31 was lower than for FP14-1 by an order of magnitude. The ruthenium and niobium activities increased slightly, and that of ?°Zr decreased slightly. The amount of 233U deposition increased by a factor of 4. The decreases in activity are dif- ficult to explain. The curious fact appears to stand that deposition of noble metals in the drained reactor is the same within an order of magnitude as with fuel in the reactor, Occasional radiochemical analyses were car- ried cut on the leaches of the metal specimens for two additional noble-metal fission products, 111A0 and '12Pd., These behaved like the other noble metals, with heavy deposition on the gas- phase and submerged specimens. This behavior is of chemical interest since no highly volatile fluorides of silver and palladium are known to exist. This observation favors the gaseous metal suspension theory of noble-metal volatili- zation. Occasional analyses were also made for other alkaline-earth and rare-earth species such as 91Qy, 141Ce, 143Ce, 144Ce, and '*'Nd, These species remained in the fuel melt and showed light deposition on the pump bowl specimens. Most of the pump bowl specimens were analyzed for 235U deposition by delayed-neutron counting 131 of their leaches. Unusually high values were ob- tained for the FP11-50 (8-hr exposure) specimens, averaging 70 pg of 2*°U on each specimen. An- other set of high values averaging 35 ug of 235U per specimen was obtained on run FP12-6 (42.5- day shutdown). The deposition of ?5Nb was also unusually high for this run. The values for run FP11-45 (normal) were also high, averaging about 30 pg of 23°U per specimen. The results for 2357 did not parallel those for any other fission product, and the reasons for the wide variations between runs are not known. 9.5 DEPOSITION OF FISSION PRODUCTS ON GRAPHITES [N MSRE PUMP BOWL Graphite specimens have been exposed in the MSRE pump bowl to observe fisgion product depo- sition under a known set of short-term conditions. Such an exposure has been carried out for 8 hr during steady 7.2-Mw operation of MSRE.. Since this exposure, with the sampler-enricher tube opened to its separate containment for 8 hr, posed some problems in reactor operation, the test was designed to yield a considerable amount of infor- mation. Three graphite samples, two of CGB grade and one of pyrolytic graphite, were exposed in the gas phase, while similar specimens were exposed to the liquid. A comparison sample of Hastelloy N was exposed in each phase. Assemblies used in this experiment consisted of a pair of the holders shown in Fig. 9.9. These holders, for safety reasons, eiclosed the praphite and Hastelloy gpecimens in perforated metal cylinders. They were connected by a loop of stainless steel cable of a length such that the bottom holder was completely immersed in the molten fuel while the top holder was exposed only to the gas phase. This assembly was exposed in the pump bowl for 8 hr with 200 ce/min of helium purge except for the last 10 min. After the exposure, tiny droplets (~0.2 mm in diameter) of greenish white fuel salt were seen adhering loosely to the gas-phase graphite gpecimens. The mechanical disturbance involved in removing the specimens from the holder was sufficient to dislodge the droplets from the graphite surfaces. Nevertheless, two of the graphite specimens were wiped with small squares of cloth, which were analyzed radiochemically. The remaining graphite and metal specimens were then leached in a neutral mixture of sodium ver- senate, boric acid, and citric acid, which dis- SAMPLES, 3/gin. DIAM X tY%gin. —— CGB GRAPHITE —- 132 ORNL-~DWG 67 —12239 MATERIAL: Ni OR NICKEL PLATED MILD STEEL | I/35-in. HOLES FOR CABLE (SPACE BETWEEN SAMPLES) ———GET SCREW =———Ni WIRE RING THROUGH HOLE IN GENTRAL ROD. MAY BE OMITTED IF SET SCREW CONSIQERED SAFE AGAINST LOOSENING INOR ~8 -4-— CGB GRAPHITE PYROLYTIC GRAPHITE Fig. 9.9. Graphite Sample Assembly. solves fuel salt but which should not attack metal. Finally, the graphite specimens were dissolved in H2804—HNO3, and the metal specimens were leached free of activity with 8 N nitric acid. All leaches were analyzed separately. Deposition of noble metals on the sections of stainless steel cable (shown in Table 9.8 as FP11-50) was as much as two orders of magni- tude higher than that on the graphite and Hastel- loy N specimens of similar surface area. Depo- sition was heaviest on the interface section of the cable and least on the vapor-phase section. It is likely that the perforated sample holders (which surrounded the specimens and which could not be analyzed) intercepted much of the activities which would otherwise have deposited on the graphite and Hastelloy N specimens. The fraction of activity removed by the aqueous chelating leach, which preceded the acid leach or dissolution, varied markedly from one isotope to the next. As Table 9.9 indicates for several typical leaches, a sizable fraction of the °Mo and °*Nb (and often the predominant amount) was found in the chelating leach. By contrast, this leach usually removed only about 10% of the '¥2?Te and ruthenium from both graphite and metal samples, but it generally removed more than half of the 1311, 95Zr, 14%8a and 3°Sr. 133 Toble 9.9. Comparison of Aqueous Chelating Agent with HN03 Leach for Removing Leposited Fission Products from Graphite and Hastelloy Digintegrations per Minute in Total Sample E,x§loied Sample® 55 1 Mo 327 1030y 952y x 101° % 101° x 10° x 167 Gas CGB V 4.6 2.99 0.15 4.3 A 1.75 32.1 1.19 0.89 T 6.35 35.1 1.34 5.23 Liquid CGB V 4.60 2.99 0.154 A 3.24 2.25 3.00 0.13 T 7.84 5.24 3.15 4.47 Gas Hastelloy N V 4.05 5.89 0.12 1.49 A 0.21 70.2 <5.3 T 4,87 76.1 <6.79 Liquid Hastelloy N V 2.24 1.99 0.07 0.35 A 1.51 15.5 0.66 <2.3 T 3.75 17.5 0.73 ~2.7 ?V represents leaching in aqueous chelating solution; A represents leaching in acid; T represents total. Since this solvent dissolves salts but not rnetals, these observations may possibly indi- cate that Mo and ?°Nb are present as oxidized species (fluorides?), while tellurium and ru- thenium are present as metals. Wiping with a cloth removed larger fractions of °°Mo and °°Nb than of activities such as 132Te, 131, 957, 14084, and #9Sr which penetrated the graphite. The deposition of fission products on pyrolytic graphite (see Table 92.10) in the gas phase was similar to that on CGB graphite, although '32Te deposited to a smaller extent on the pyrolytic graphite. In the liquid phase, ®°Mo and **2Te deposited slightly more heavily on pyrolytic graphite than on the CGB. The deposition of ?5Nb was markedly heavier on Hastelloy N in the liquid phase than in the gas phase. Tellurium-132 and 1*%Ba deposited more heavily on the metal exposed in the gas phase. Other isotopes deposited similarly on metal in the two environments. When metal and graphite were exposed to the gas phase, somewhat more 132Te, 1311 and 95Zr deposited on the metal. When both were exposed to the liquid phase, less °°Mo, 14°Ba, and 2%Sr and more '32Te and '°'] deposited on the Hastelloy. The deposition per square centi- meter of a given isotope varied by less than a factor of 10 between any Hastelloy N or graphite samples. When the activity, in disintegrations per minute per square centimeter, of fission products on the graphite samples from the 24,000-Mwhr core ex- posute is compared with that on submerged graphite specimens from the 8-hr exposure (Tables 9.4 and 9.10) the ratios were 0.9 for °Mo, 1.7 for 132Te, 3.5 for '?3Ru, 14.0 for 19%Ru, 95 for ?5Nb, and 56 for *3Zr. A similar comparison of the activities on the Hastelloy N specimens from the two exposures (Tables 9.6 and 9.10) gives ratios of 20 for ®°Mo, 6 for 132Te, 170 for 193Ruy, 400 for 19°Ru, 75 for 5Nb, and 30 for 93Zr. The ratio of full-power exposure times is 417; the ratio of total exposure times is, of course, considerably greater., Both sets of activity ratios may be bi- ased high since the sample holder in the 8-hr ex- posute may have intercepted a sizable fraction of the depositing activities. Examination of the ratios given above shows that only deposition of '°*Ru on Hastelloy N proceeds at approximately the same rate for 8 as for 3340 full-power hours. However, the measured deposit of a given radioactive nu- clide actually reflects the deposition rate only over the last two or three half-lives of exposure time for that nuclide. Thus the measured ac- Table 9.10. Deposition of Fission Products During B-hr Exposure in MSRE Pum p Bowl Exposed Sampte Area . Disintegrations per minute in Total Sample” 235y in — (sz) 99M0 132Te 1311 E}szr 8981‘ 140-Ba 103Ru 106RU QSNb 143Ce (ng) w1010 x1o!l c10° %107 x 108 x 107 x 107 % 107 x10% x10® Gas CBG No. 1 3.68 9.35 3.51 3.42 5.23 8 1.62 1,34 4.0 4,84 5.14 Gas CBG No. 2G 3.68 5.56 4,44 1.65 1.4 0.47 2.16 6.0 8.8 1.31 Gas PG 2.76 6.13 0.371 1.25 0.91 7.52 1.09 2.3 6.0 5.3 0.71 Liquid CBG No. 1L 3.68 7.8 0.52 3.14 4.47 13.0 1.58 3.15 9.27 4.0 4,76 fLiquid CBG No. 2L 3.68 11.2 0.464 2.34 0.58 0.33 5.29 15.9 4.4 .27 1.55 Liquid PG 2.76 33.4 0.7¢ 2.81 2.4 158 0.59 25.2 8.06 5.4 15.5 1.37 Gas Hastelloy N 2.46 4.87 7.61 15.6 4.8 10.9 1.09 2.3 6.55 Ligquid Hastelloy N 2.46 3.75 1.75 6.77 2.7 1.97 0.063 0.73 2.0 63.5 13.8 8p11 activities calculated for time of sampling. veT tivities of ?°Mo and !'32Te deposited in a long exposure are a measure of deposition rate over only the last week or so of exposure. If we assume a constant deposition rate, the amount deposited in time t is given by the familiar equation: k —At N=—(10 -, )\( where k is the constant deposition rate and A is the decay constant. Using this equation a theoretical ratio can be calculated for each iso- tope for exposure times of 3340 and 8 hr: . M. .o 1 — e—AX3340 L AXE Ns 1 -¢ These ratios turn out to be 12.6 for °?Mo, 14.5 for 132Te, 157 for 123 Ru, 365 for 19%Ry, 142 for ?5Nb, and 218 for ?3Zr. Whether fortuitous or not, these calculated ratios agree with the ob- served ratios for deposition on Hastelloy N very well for the ruthenium isotopes and within a factor of 2 for molybdenum, tellurium, and niobium. This agreement suggests that the deposition of noble metals on Hastelloy N does indeed proceed at approximately the same con- stant rate for a 3340-hr exposure as for an 8-ht exposure, By contrast, the calculated ratios are much higher (usually by teafold or more) than the ob- served ratios for deposition on graphite except for >5Nb and ?°Zr, whose calculated ratios are high by factors of only 1.5 and 4 respectively. This indicates that the deposition rate of noble metals (except *°Nh) on graphite decreases with exposure time. Neutron economy in MSRE would benefit if the deposition rate of noble-metal fis- sion products on graphite decreases with expo- sure time while that on Hastelloy N remains constant., As the data in Table 9.10 indicate, the 8-hr deposition rates are not greatly different 135 on graphite and metal, and the activities of %Mo, 1327Te, '93Ry, and 19°Ru were similar for the 7800- and 24,000-Mwhr exposures. This suggests no change of rate between 7300 and 24,000 Mwhr for ?*Mo, 32Te, and '°*Ru and a decreased rate for the long-lived '%8Ru. The discussion above gives a generally satis- factory picture of noble-metal deposition for ex- posures of 8 hr or longer. The picture is greatly complicated by the inclusion of results from the short (1 to 10 min) exposures of the stainless steel cable specimens in the pump bowl fuel (see Table 9.8). For most isotopes, the depositions were only slightly greater (a factor less than 5) for the 8-hr exposure than for 1- to 10-min ex- posures. Tellurium-132 and ?°Zy deposits in 8 hr were about 100 times larger than for the short eXpOsSUures. The calculated ratios from the formula given above would be very close to the time ratios for these short exposures, Comparing the 8-hr run with a 1-min run, the time ratio is 480. The short-exposure data thus indicate that (except for 1*?Te and ?°Zr) the original deposition rate is very fast compared with the rate after 8 hr. It may be that when fresh samples of metal are exposed to fuel salt two deposition mechanisms come into play. The first process rapidly de- posits noble metals on the stainless steel in less than a minute and then stops (or reaches equilibrium). The second (slow) process then continues to deposit noble metals at a fairly constant rate for exposure times up to 3340 hr. It may be speculated that the first process is a rapid pickup of colloidal noble-metal particles from the fuel on the metal surface; the second may be a slow plating of noble metals in highet valence states present at very low concentra- tions in the fuel. It will be interesting to expose graphite samples in the fuel for short times to see whether a short-term mechanism also exists for deposition on graphite. 10. Studies with LiF-BeF, Melts 10.1 OXIDE CHEMISTRY OF ThF¥ ,.UF, MELTS B. F. Hitch C. E. L.. Bamberger C. F. Baes, Jr. The precipitation of protactinium and uranium from LiF-BeF -ThF , mixtures by addition either of BeO or ThO, was demonstrated several years ago by Shaffer et al.?~? as a possible method for processing an MSBR blanket salt. The purpose of the present investigation is (1) to verify that the oxide solid phase formed at equilibrium with UF -ThF ,-containing melts is the expected solid solution of U0, and ThO, and (2) to determine the distribution behavior of Th** and U*™ between the oxide and the fluoride solutions: U*H(E) + Th* %) = U*¥o) + Th* XD (1) (where f and o denote the fluoride and the oxide phases). The experimental technique is similar to that used in a UO,-Z10, phase study.*® The ThO, and UO, were contacted with 2LiF - BeF , containing UF , and ThF , within nickel capsules under a hydrogen atmosphere in a rocking furnace. The equilibrated oxide solids were allowed to settle before the samples were frozen. Provided suf- ficient fluoride phase had been added originally, good phase separation was thus obtained. A (U—Th)O2 solid solution has been found in every sample examined thus far, and, moreover, 'y, H. Shaffer et al., Nucl. Sci. Eng. 18(2), 177 (1964). 2_]. H. Shaffer, G. M. Watson, and W. R. Grimes, Ke- actor Chem. Div. Ann. Progr. Rept. Jan. 31, 1960, ORNL-2931, pp. 84—90. 3_]. H. Shaffer et al., Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1961, ORNL-3127, pp. 8--11. 4K. A. Romberger, C. ¥. Baes, Jr., and H. H. Stone, J. Inorg. Nucl. Chem. 29, 1619 (1967). 136 the lattice parameter determined by x-ray diffrac- tion® was consistent with the composition calcu- lated for such an oxide phase. The equilibrium quotient for the above metathetic reaction was determined by analysis of the fluoride phase for the small amount of uranium which it contained. The results obtained thus far give Xuy X X Th{f) 1000 to 2000 U (f) X’[‘h(o) Q- ) at 600°C. The mole fraction of uranium in the oxide phase [XU(O)] was varied from 0.2 to 0.9, while the mole fraction of thorium in the fluoride phase was in the range 0.01 to 0.07. It thus appears that relative to Th**, the U*" present is strongly extracted from the fluoride phase by the oxide solid solution formed at equi- librium. There is every reason to expect that Pa*" will distribute even more strongly to the oxide phase. This is in agreement with the ef- fective precipitation of U*" and Pa* ™ by oxide ~ first reported by Shaffer ef al. The formation of a single oxide solution phase to which Th*7, U*" and Pa®™ all could distribute offers a much more flexible and effective oxide separation method for a single-region breeder reactor fuel than would be the case if only the separate oxides (ThO, and UO,) were formed. For example, a possible processing scheme for an MSBR might involve some sort of a countercurrent contactor (Fig. 10.1) in which a (U-'Th)02 solid solution of 5These x-ray examinations were performed by G. D. Brunton and . R. Sears of the Reactor Chemistry Di- vision. The mole fraction of ThO;, in the oxide solid solutions was calculated from the lattice parameter, using the equation given by L. O. Gilpatrick and C. H. Secoy, Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNIL.-3789, p. 241. ORNL-DWG &7-14820 HF, Hs TO REACTOR l ["‘“‘”“HF’ Hy0, Hp —mr—— FROM REACTOR Li+, Belt £ Tht UM et COUNTERCURRENT CONTACTOR . FLUORIDE SOLUTICN OXIDE SOLID SOLUTION [=——(Th, U) 02 ret—ee {Th, U, P0)O3 =t HOLDUP f——se] MAKEWP =L HOLDUR e MAKELE r e HYDROFLUORINATION Lit Be2t - ThAt ya+ PRODUCT MAKELP Fig. 10.1. Suggested Flow Diagroms for Protactinium Removal by Oxide Extraction. controlled initial composition is equilibrated with the blanket salt. It is expected that 233Pa will distribute strongly to the oxide phase, while the amount of ??3U removed will be controlled by the uranium contenf of the influent oxide phase. Sub- sequently, the oxide phase might be stored while the ?%°*Pa decays and then largely recycled to the contactor with some adjustment of the composition and with removal of a portion of the (Th-U)0, solid solution as product. Before returning the salt to the reactor, the dissolved oxide could be removed by HF-H, sparging. This processing concept might be applicable to a one-region breeder fuel. It should only be necessary to adjust the uraninm content of the oxide phase upward in order to be compatible with the higher uranium content nec- essary for a fuel salt. It is planned to continue the present measure- ment of the disttibution of U??, and perhaps Pa*®, between fluoride and oxide phases as a function of composition and temperature. In these meas- urements, vigorous agitation will be used in the hope of shortening the equilibration times. If the results should warrant, a more detailed study of the rate-controlling factors in the distribution of U** and Pa** will then be made. 10.2 CONTAINMENT OF MOLTEN FLUORIDES IN SILICA Chemistry C. E. L. Bamberger J. P. Young R. B. Allen C. F. Baes, Jr. The advantages of silica as a container material for molten salts in spectrophotometric, emf, and other measurements include high thermal shock resistance, good ultraviolet transmittance, high electric resistivity, low price, and ease of fabri- cation. For the containment of molten LiF-BeF mixtures, an obvious disadvantage is possible chemical attack as a result of the reaction 2BeF (d) + Si0,(s) & 8iF (g) + 2BeO(s) . (3) However, when the equilibrium partial pressure of SiF , was calculated for this reaction from avail- able formation free energies for crystalline §i0,,° crystalline BeO, ® gaseous SiF ,® and dissolved GJANAF Thermochemical Tables, Clearing House for Federal Scientific and Technical Information, U.3. Dept. of Commerce, August 1965, 138 BeF , in 2LiF - BeF ,, 7 this partial pressure was found to be surprisingly low, that is, 0.03 mm at 700°K to 21 mm at 1000°K. Furthermore, pre- liminary direct measurements now in piogress con- firm the low range of these partial pressures. This suggested that an overpressure of SiF , might pre- vent, or at least reduce, the attack of the silica container and prevent the precipitation of BeO. The compatibility of Si0, with these LiF-Belk', melts is primarily associated with the relative stability of BeF, compared with that of BeO; the standard free energies of formation at 1000°K are ~206 and 120 kcal/mole respectively. Since many other metallic fluorides show even greater stability compared with their corresponding oxides, it seems that S10, should not be ruled out as a container material for molten fluorides without first estimating the equilibrinm position of such reactions as X ZSioz(S) + MF (s or H= MOX/Q(S) LSPGO, @ which involve the pure metal fluoride and metal oxide, and S10,(s) + 4F ~ = 207~ + SiF (g) . (5) The latter reaction is the one of most interest, since it gives the level of oxide contamination of the fluoride melt. be judged if activity coefficients or solubilities can be assigned to MF_and MO, ,, in the melt under consideration. In the present case the ac- tivity of BeF , in 2LiF - BeF , (~0.03) and the solubility of BeO (~0.01 mole of O? per kg of salt at 600°C) have been measured.’ From these and the avaijlable thermochemical data the follow- ing equilibrium quotient may be estimated for the above reaction [Eq. (5)] in 2LiF - BeF , at 600°C: Its equilibrium position may Q=[0%"]? PSiF4 =4 %10~ 7 atm mole */kg? . 6) 7C. F. Baes, Jr., *"The Chemistry and Thermodynamics of Molten Salt Reactor Fluoride Solutions,?’ SM-66,/60 in Thermodynamics, vol, 1, JAEA, Vienna. From this, it can in turn be estimated that in the presence of 1 atm of SiF , the oxide ion concen- tration at equilibrium with Si0, should be only 6 x 10™? mole/kg. Another factor to be considered when using sil- ica is the possible formation of silicates; for exaniple, 2MF ,s or 1) + 2510 (s) = M,Si0,(s) + SiF (&) . (7) In the present case, for example, Be ,510, (phen- acite) should be formed as the stable reaction product rather than BeO at low S, parttial pres- sures; however, phenacite is not very stable rel- ative to 2BeO + SiO, and should not alter the above conclusions about the compatibility of sil- ica with LiF-BeF , melts. In other fluoride sys- tems, the formation of metallic silicates may be the controlling factor. The solubility of SiF', in the molten fluoride must be considered because possible reactions such as SiF (&) + 2F~ = SiF *~ (&) could produce high solubilities which not only would alter the salt phase, but also would cause the reaction with Si0, to proceed farther than otherwise expected. Accordingly, the solubility of SiF , in 2LiF - BelF , was estimated by means of transpiration measurements, The prepurified salt was saturated with SiF , by bubbling a mix- ture of 0.1 atm of SiF , in He throngh the salt until the effluent and influent compositions were the same. The dissolved gas was then stripped with helium and trapped in an aqueous NaOH bub- bler, the amount of SiF , being equivalent to the amount of NaOH consumed by its hydrolysis: SiF ,(g) + 40H™ = 4F~ + 2H,0 1 $i0(s) . (9) The solubility was calculated by means of [SiF,] K= 5 ; (10) SiF4 [SiF 1 =[2ng, . H(PSiF4V/RTm)]/w, (11) }‘:‘nsm == {otal number of moles of SiF, 4 titrated, PSiF4 = SiF, pressure at equilibrium with the melt, E(PS. . V/RT ) = correction term which in- iF 4 m cludes the various dead vol- umes at different tempera- tures, T, w = weight of the melt in kg. Using Egs. (10) and (11) the solubility was found to be K ;= 0.032 £ 0.005 mole kg™ Latm—1! at 499°C, KH = 0.035 +0.005 mole kg~ ' atm— ! at 540°C. These solubilities are of the same order of magni- tude as the solubility of HF in 2LiF - BeF , (~0.01 mole kg~ ! atm ™ 1).® At higher temperatures, in the range 600 to 700°C; the solubility was so low that by the rather insensitive method of measure- ment only an upper limit could be set: K _ <0.01 mole kg~ ! atm ™. The data obtained indicate that at 1000°K, using an SiF , pressute close to the equilibrium value over 2LiF - BeF , the solu- bility of SiF | is ~0.003 mole %. Another undesirable side reaction that should be considered in these systems is the formation of silicon oxyfluorides; for example, SiF ,(g) + Si0 (s) = 2Si0F ,(g) (12) 3SiF ,(8) + Si0,(s) & 251 ,0F (&) . 13) Such silicon oxyfluorides have been identified by Novoselova et al.? during the synthesis of phen- acite from 5i0,, BeO, and Na ,BeF , at tempera- tures in the range 700 to 800”C. In one meas- urement they report equilibrium partial pressures of 0.162 atm of SiF , and 0.07 atm of SiOF , over the reaction mixture at 813°C. In other experi- ments wherein equilibrium was not reached, they were able to identify Si OF .. We performed tests in which He after passage through molten 2LiF - Bel , containing powdered BeO and SiO, at tem- peratures up to 650°C was analyzed by means of a mass spectrometer. No silicon oxyfluorides were found to be present. 8P, E. Field and J. H. Shaffer, J. Phys. Chem., in press. 139 Spectrophotometric Measurements with Silica Cells C. E. L. Bamberger C. F. Baes, ]Jr. J. P. Young Behavior of U4™, which an S1F, overpressure would prevent reaction (5), UF , a colored solute which would produce an even less soluble oxide than does BeF ,, was added to LiF-BeF , melts in SiO, containers. Tet- ravalent uranium thus should act as a sensitive ““internal indicator’ of the reaction of F~ with Si0,. It has a known absorption spectrum of suit- able inteusity, *? and hence its concentration can be followed spectrophotometrically. For the re- action - To estimate the extent to UF4(d) + SiOz(s) = SiF 4(g) + UOZ(S) , (14) equilibrium partial pressures of SiF , were pre- dicted to be reasonably low; for example, with 0.002 mole fraction UF‘5 in 2IAF - BeF ,, calcn- lated SiF , pressures varied from 3 mm at 700°K to 115 mm at 1000°K. Spectrophotometric mon- itoring of the concentration of gt provides a very good means for studying the stability of the sys- tem. Melts of 2LAF - BeF , containing variable con- centrations of UF, (0.0053 to 0.13 mole % or 0.003 to 0.008 mole/liter) were held under SiF (400 mm) at tempetatures ranging from 490 to 700°C in sealed silica tubes. The tubes were placed di- rectly in a special furnace'! located in the light path of a Cary spectrophotometer, model 14M, or in a heated nickel metal block and later transferred to the furnace inside the instrument. Temperatures were controlled to +1°C by means of a Capacitrol 471 controller. Constancy of the U*" absorbance peaks at 1090 and 640 nm showed the uranium concentration in the melt to be constant within the precision of the spectral measurement (about 1%) for at least 48 hr. It was, therefore, evident that no significant attack of the silica container or significant contamination of the melt occurred. During longer periods of time A. V. Novaselova et al., Proc, Acad. Sci. USSR, Chem. Sect. {(English Transl.), 159, 1370 (1964); A. V. Novoselova and Yu. V. Azhikina, [norg. Mater. USSR (English Transl)), I{9), 1375 (1966). loj. P. Young, Incrg. Chem. 6, 1486 (1967). 17, P, Young, Anal. Chem. 31, 1892 (1959). 140 (up to one week) the transparency of the containers was noticeably affected, and the base line of the spectra rose. The net absorbance of the U** peaks, however, remained reproducible even though in some experiments the temperature was cycled from 500 to 700°C. No spectral experiment was con- tinued for longer than a week. However, in other studies, melts of 2LLiF * BeF | were maintained as long as one month at S00°C without noticeable damage except loss of transparency {(presumably due to some process of devitrification) of the fused silica. The precision of these quantitative spectral meas- wrements of U*" was better than that possible with windowless cells.}? The molar absorptivity (Am) of U** had been observed!® to be 14 + 10% and 7 + 10% at 1090 and 640 nm, respectively, in molten LiF-Be}l , at 550°C contained in windowless cells. With the melts contained in the present silica cells the A_of U*'is 14.4 + 0.5 and 7.3 + 0.3 at 1090 and 640 nm, respectively, in molten 2LiF - BeF , at 560°C. The effect of temperature on the 4 _ of several absorption peaks of U*™ in 2LiF - BeF ) was studied. The results are summarized in Table 10.1. Behavior of Cr® . — The solubility of CrF | in Li.F-BeF2 is being investigated spectrophoto- metrically with the melts contained in SiO, cells. A minimum solubility of 0.43 mole % Cr®* was ob- tained at a temperature of 550°C. This is not the solubility limit, but the solutions are too opaque to observe spectrophotometrically in a 1-cm path length. Further studies will be carried out with 510, cells of a shorter path length. The spectrum of Cr3'in LiFF-BeF , at 500°C consists of three peaks at 302, 442, and 690 nm, all arising from transitions within the 3d level. sorptivity of the 690-nm peak was calculated to The molar ab- be 6.6. This value compares favorably with an Table 10.1. Effect of Temperature on the Molar Absorptivity of Several Peaks of U4+ in Molten 2LiF - BeF, Molar Absorptivity Temperature (°C) 1050 nm 640 nm 498 16.4 3.0 560 14.4 7.3 650 13.8 6.2 698 12.9 5.7 approximate value of 7 reported from a cursory study of the system in windowless cells.!® The molar absorptivity of the two other peaks is some- what higher, about 10, but fucther work will be necessary to establish a more precise value. 10.3 ELECTRICAL CONDUCTIVITY OF MOLTEN FLUORIDES AND FLUORGBORATES J. Braunstein K. A. Romberger Studies have been initiated to measure the elec- trical conductivity of molten fluorides and fluoro- borates of interest as fuel, blanket, and coolant salts for the Molten-Salt Breeder Reactor. Pre- liminary measurements were made to provide esti- mates of the electrical conductivities for engi- neering use in the design of electrical circuits for instrumentation in the MSBE. Cell desiga has been investigated, including the use of silica in view of recent indications that dry fluorides may be stable to silica in the presence of low concen- trations of SiF4 (see Sect. 10.2). Preliminary measurements were made of the con- ductivity of 2LiF-BeF ,, of LiF-ThF , eutectic (29 mole % ThF ), and of NaBF , by observing current-voltage curves at 100 hertz and using plat- inized platinum wire electrodes in a silica cell at 4000 hertz. !* Potassium nitrate at 390°C was used to determine the effective cell constant. Temperature coefficients were estimated from data on mixtures of alkaline-earth fluorides with alkali fluorides or with cryolite !® and from the data of Greene ' ® on FLINAK and NaF-Z:F -UF , mixtures. The cell used for the current-voltage measure- ments was similar to that used by Greene!® and consisted of a pair of platinum wire electrodes of 0.07-cm diameter separated by about 2 cm and im- mersed to a measured depth in the melt contained 12}. P. Young, Anal. Chem. 36, 390 (1964). Bpsr Program Semiann. Progr. Rept. Aug. 31, 1966, ORNL-4037, p. 194. 1 . . . . 4J. Braunstein to R. I.. Moore, private communication, Aug. 14, 1967. 1%, w. Yim and M. Feialeib, J. Electrochem. Soc. 104, 626 (1957); J. D. Edwards et al., J. Electrochem. Soc. 99, 527 (1952); M. de Kay Thompson and A. L. Kaye, Trans. Electrochem. Soc. 67, 169 (1931). 16N. D. Greene, Measurements of the Electrical Con- ductivity of Molten Fluorides, ORNIL-CF-54-8-64 (Aug. 16, 1954). 141 in a silica tube. (The standard, KNO,, was con- creasing the current path length through the elec- tained in a Pyrex beaker.) The voltage drop across trolyte. In the new cell, the electrodes will enter a resistor in seties with the electrodes and elec- the melt through silica tubes of about 0.3 cm ID trolyte was measured with an oscilloscope whose and extending 3 cm below the surface of the melt. vertical amplifier was provided with a calibrated The dependence of the measured conductivity on voltage offset. Bridge measurements also were the temperature and composition of the melts and made with the same cell, using a General Radio on frequency is under investigation. rmodel 1650A impedance bridge. In order to im- prove the sensitivity and reproducibility of the measurements, a bridge was constructed with a 100-ohm Helipot to provide the ratio arms and with Table 10.2, Electrical Conductivity of Molten Fluorides and Fluorobsrates a decade resistance box and variable parallel ca- pacitance in the balancing arm. An oscilloscope Conductivity Temperature 1cmperature served as the phase balance and null detector. (ohm ™ em™1) (") Coiffidelm The preliminary results listed in Table 10.2 are [ probably in error by 20 to 50%, the limiting factor . 3 being the small cell constant in the initial ex- 2LAF Bel 1.5 650 1x10 periments. LiF -ThF 2.0 650 3x107° A new dipping cell has been constructed to pro- NaBF 0.8 420 2% 1073 vide a cell constant of the order of 100 by in- 11 The appearance in appreciable concentrations of several fission product species in the gas space of the MSRE pump bowl remains the biggest chemical surprise in MSRE operation. This behavior does not, of itself, seem to pose serious problems for the MSRE or for subsequent large molten-tluoride re- actors, but it is clearly important to establish the mechanism by which these diverse species, notably 9%Mo, ?°Nb, '?%Ru, '°Ru, 132Te, and even !!'%Ag, get into the gas stream. It is possible that at least some of the gas-bome species listed above have volatilized from the melt We have, accordingly, initiated studies of the (less well known) lower fluorides of these materials both as pure compounds and as their solu- as fluorides. tions in molten LiF-BelF , preparations. The pre- liminary studies performed to date have been de- voted to the fluorides of molybdenum, but it is anticipated that, should the results warrant, the study will be broadened to include other elements on the list. 11.1 SYNTHESIS OF MOLYBDENUM FLUORIDES C. F. Weaver H. A. Friedman Molybdenum is a constituent of Hastelloy N, the MSRE structural metal, as well as an important high-yield fission product. Consequently, and especially since molybdenum is one of the elements appearing in the MSRE exit gas stream, the be- havior of molybdenum and its fluorides in molten fluoride fuels in contact with graphite and Hastel- loy N is of considerable interest. ™3 A search of the literature on the tluorides and oxyfluorides has been completed and reported.? The fluorides MoK _, M0F4, Mo, F,, MoF, and MoF _ have been shown to exist, but of these only MoF _ is commercially available. It has been neces- sary, therefore, to refine methods for preparation, 142 Behavior of Molybdenum Fluorides and to prepare gram quantities, of the lower- valence compounds for study. Preparation of MoF _ by the reaction SMOF6 + Mo~ EiMDFS has been described by Edwards, Peacock, and Small,® who have also established the crystal structure of MoF _. We have successfully employed this method in which Mol _ is refluxed over metal- lic molybdenum at temperatures in the ianterval 35 to 100°C in glass apparatus. Molybdenum pentafluoride is a yellow, hygro- scopic material which melts at about 65°C to a yellow liquid of high viscosity.® Its vapor pressure at the melting point is about 2 mm,® and it boils at about 212°C.%-° Disproportionation behavior of MoF appears to be very complex. We have observed that heating of the material to 100°C while maintaining a good vacuum leaves the MoF _ unchanged. However, we have shown that if Mok _ is heated at 200°C and if the volatile product is removed by pumping, the reaction is 3M0F5—> 2M0F6 + MOF3 i 'W. R. Grimes, Chemical Research and Development for Molien Salt Breeder Reactors, ORNIL.-TM-1853 pp. 61--81 (June 6, 1567). ’ 3. s. Kirslis, F. F. Blankenship, and C. F. Baes, Jr., Reactor Chem. Div. Ann. Progr. Rept. Jan. 31, 1967, ORNI.-4076, pp. 48—53. 3S. 5. Kirslis and F. F. Blankenship, MSR FProgram Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4119, pp. 124-43. 4C. F. Weaver and H. A. Friedman, A Literature Sur- vey of the Fluorides and Oxyfluorides of Molybdenuin, ORNL-TM-1976 (October 1967). SA. J. Edwards, R. D. Peacock, and R. W. H. Smail, J. Chem. Soc. 4486—91 (1962). 5D. E. LaValle et al., J. Am. Chem. Soc. 82, 243334 (1960). The MoF , so obtained has been identified by its x-ray diffraction pattem established by LaValle 143 ¢t al.® on the material synthesized by other methods. This convenient method of preparation, which seems not to have been used before, has served to pre- pare several pure batches of MoF . When MOF5 is pumped, in this same way at 150°C, the reaction has been reported® ” to be 2M0F5 —_— MOF4 + MOF6 . We have established that the solid product is not Mo¥ | (as produced in the 200°C reaction), but we have not vet completed our identification of the material. The compound MoF | has been shown to react with LiF to fom at least two binary compounds. These materials, whose stoichiometry has not yet been established, are both birefringent. The mean index of refraction for compound I is at 1.520, while its major x-ray diffraction peaks (copper x-tay target; 26 values in degrees) are at 21.4, 19.4, and 27.0; for compound II, the mean refrac- tive index is 1.480 and the major diffraction peaks ate at 22.3 and 20.3. The stoichiometry and the optical and x-ray data will be established as soon as well-crystallized samples are available. 11.2 REACTION OF MOLYBDENUM FLUORIDES WITH MOLTEN LiF.BeF, MIXTURES C. F. Weaver H. A. Friedman The appearance of ?Mo in the exit gas from the MSRE ? has suggested the possibility that a vola- tile fluoride of molybdenum exists in the MSRE. We have accordingly begun an examination of reactions of molybdenum {luorides with MSRE fuel and fuel solvent mixtures. Molybdenum hexafluoride is a very volatile material (boiling point, 34°C) and is a very strong oxidant. The nickel container was rapidly attacked when Mol diluted to ™ 50% with helium was passed through the MSRE fuel mixture (in which all of the uranium was as UF,) at 650°C; after 1 hr the fuel melt contained 9500 ppm of nickel. A small guantity of uranium was transported (presumably as UF ) by the gas, but, as expected, no appre- 7G. H. Cody and G. B. Hargreaves, J. Chem. Soc. 156874 (1961). ciable reaction of the MoF , with the melt was ob- served. When MoF was passed through a fuel mixture to which 0.08 mole % UF | had been added, the UF | was rapidly oxidized to UF . A small quantity of uranium was again observed in the exit gas lines, and violent corrosion of the nickel was observed. The valence state to which the MoF was reduced was not detemined in these experi- ments. It is probable, however, that the molybde- num was present as either molybdenum metal or MoF ,, since MoF and MoF , have not been reported to exist and MoF ,, Mo I, and MoF seem to be thermally unstable above 200°C. The compound MoF , dispropertionates readily® under a vacuum at temperatures greater than about 600°C to form molybdenum and MoF .. However, this material has been heated under its own pres- sure i closed nickel and copper capsules at 500 and 710°C, respectively, for perieds in excess of ten days without disproportionation. In addition, it has been shown that MoF , at 500 mm will react at 560°C with molybdenum to form MoF .. Conse- quently, an equilibrium 2MoF , T==Mo + MoF must exist in this temperature range with MoF pressures of a fraction of an atmosphere. When MoF . was heated in a nickel capsule to 500°C in the presence of 2LiF-BeF , (50-50 mole %), the reaction 2}\/[0173 + 3Nj —> EiNiF2 + Mo occurred. Apparently the 2LiF.BeF liquid dis- solved the protective coating on the nickel wall. Repeating the experiment with copper indicated that much less corrosion occurred. In this case the molybdenum was found both as MoF ; and as the 1.480 refractive index LiF-MoF , compound of 3 unknown stoichiometry. The reaction MoF , (1 mole %) + 3UF , (4 mole %)~—> Mo + 3UF, was demonstrated to proceed essentially to comple- tion at 500°C in a copper container. The reverse reaction was not observed, although it is possible that equilibiium was achieved with very small con- centrations of MoF , and Ul ,. Present information suggests that the following events (equations unbalanced) may be significant in the kinetics of the reduction of MoF | to molyb- denum metal by UF,: 144 UF source ——> MOF6 —_— M0F3 MoF , <= Mo + MoF \U F, T This scheme allows the molybdenum to be ‘‘trapped”’ in the trivalent state until the source of MoF _ is removed. Then the molybdenum is converted to the metal by the above reactions, which continue to produce the volatile MoF | at a decreasing rate until the process is complete. Attention is now being given to experimentally checking this hy- pothesis with molybdenum concentrations in the ppm fange. 11.3 MASS SPECTROMETRY OF MOLYBDENUM FLUORIDES R. A. Strehlow J. D. Redman The volatilization behavior of molybdenum and other fission product fluorides in the MSRE has led to a study of molybdenum fluorides. Mass spectro- metrically derived information is of particular value in studies involving volatilization, since, at least in principle, the vaporizing species ate analyzed with a minimum time lapse. This gives an oppor- tunity to observe some transient phenomena and to distinguish among various oxidation states and im- purities which may be present. The wotk so far has been concemed with the mass analysis of vapors from three molybdenum fluoride samples. The first objectives were to assess material purity and to establish the mass spectro- metric cracking pattems for these materials which have not previously been subjected to mass analy- sis. The three samples are designated and de- scribed in Table 11.1. Sample I, duting an increase of temperature from 400 to 725°C, yielded first M002F2 at the lowest temperature. As the temperature was increased, the peaks associated with this species decreased in magnitude and a family of peaks attributed to MoOF , appeared. Near the upper limit of the tem- perature excursion, a mass peak family was ob- served which is attributed to MoF _ and MoF ; vapor species. The large amount of volatile oxides indi- ] cated that an oxidation-hydrolysis had occurred - and that better, or at least fresher, material was needed. A somewhat increased amount of mass 96 was observed from this sample, which is attributed . to orthosilicic acid (H ;Si0O ) rather than to the molybdenum, since its peak height was not a con- stant multiple of the other Mo ' peak heights. Sample II, Mol ,, was prepared by C. F. Weaver and H. A. Friedman and was heated in the Knudsen cell inlet system of the Bendix time-of-flight mass spectrometer. The compound MoO,F was not ob- served, but some MoOF | was evident (along with the usual SiFS,Q,l ions) at temperatures as low as 350°C. Beginning at 275°C, M0F5+, M0F4+, M0F3+, MoF;, and MoF " were also observed. The M0F’5+/I\.roF4Jr peak height ratio was about unity, indicating some Mol as well as MoF ( (or MoF ). We have insufficient evidence to demonstrate that MOF4 has been part of our sampled vapor. At tem- peratures greater than 600°C, only fluoride species were observed. The spectra for sample I at tem- peratures of 250, 300, and 725°C are shown in Fig. 11.1. A photograph of an oscilloscope trace of . Table 11.1. Mass Analysis of Vapars from Three Molybdenum Fluoride Samples Nominal Sample Composition Source 1 MDF3 Exposed to air for D. EE. LaValle, Analytical Chemistry several years Division 11 MOF3 Recent synthesis C. F. Weaver and H. A. Friedman, Reactor Chemisiry Division 111 MOF5 Recent synthesis C. F. Weaver and H. A. Friedman, Reactor Chemistry Division 145 102 ORNL ~-DWG 87 - 103408 T i T T =725 MoF.* MoF,t MoFgt —4 MoF b : 10 MoF * ¢ J Mot 31077 : ‘ fi 34 :) ' ¢ g0 Lo L AL b d ! LU SR - & 1072 I T T T e, 4 7 = 300°C MoOF 3 4wt - E Ji l 5 402 ffi?‘[—_ T“_'1 """"" . 2 7= 250°C MoOF 5 % . MoF ! 3 MOFZ + *j 0 " MDF+ MOOF+ MOOFZ 0 + '{—:J) Mot MoO 1o L o L LAl b _ 4 oo 90 100 10 20 430 140 150 166 470 130 90 200 ’ amu Fig. 11.1. the MoOF ,*/MoF ¥, MoOF _ "/MoF ,*, and MoF _* peaks at 375°C is shown in Fig. 11.2. This shows, for illustrative purposes, the comhbined spectra for the various isotopes and compounds in the mass range from 150 to 194 amu. At 725°C the peak hmght for the Mol , * relative to the height of the MOF peaks as Lompared with the spectrum at 375D allowed calculation of a tentative cracking pattem assigned to MoF ,. This is seen only at elevated temperatures. An impurity at m/e = 167 was assigned the probable formula 5t ,0F . ' Mass analysis of these species is facilitated by the unique isotopic abundance ratios of molybdenum isotopes and the difference of 3 amu between oxy- gen and fluorine. This yields, if one selects ad- Jac,ent pairs of species from the MoO_F, */MoOF +/ MoF, " family of peaks, three unigue peaks for edch spec1e{~_., allowing a precise calculation of the peak height ratios for the ‘‘overlapping” peaks to be Mass Spectra of Sample 1l (MoF 4) During a Heating Cycle from 250°C 10 725°C, made. The dimer Mo ¥ has, of course, the 15 peaks which would be expected from the seven stable molybdenum isotopes. Figure 11.3 shows the good agreement between the calculated and the PHGTO 89172 155 193 Fig. 11.2. Oscilloscape Trace of Mass Spectrum of Somple H (M0F3). observed peak heights for Mo 2F9+. This, with the observed precision of the analyses of the various families of peaks, gives a satisfactory level of confidence in the applicability of mass spectral information to the molybdenum fluoride studies. Sample IHl, MoF, yielded the spectrum shown in Fig. 11.4 at 125°C. The presence of MOZFQ+ is attributed to Mo F , rather than to dimeric MoF _, since Mo, F is a reported compound which could be expected to yield a parent ionic species in the spectrometer. The Knudsen cell with the sample in it was removed from the vacuum sys- tem, exposed to air for ¥ hr, remounted, evacu- ated, and heated. No species was observed until a temperature greater than 450°C was reached. A scan at 525°C was made which yielded the spec- trum shown in Fig. 11.4. The sole parent species yielding this spectrum appeared to be MoO,F . Cracking patterns have been derived for the various species primarily based on scans in which the dominant or sole species was well character- ized. Since MoF _ has not yet been studied in our equipment, we have assumed the applicability of the reported data (see Table 11.2). Our derived cracking patterns are shown in this table. The compound designation for MoOF , is uncertain (MOO'F3 is possible as far as we know). The com- pound designated Mok _ may be a mixture of MoF and perhaps 5 to 10% MoF . The observation of Mo, 9+ at temperatures of 325°C and of MoF | at temperatures up to 725°C is noteworthy. The compound Mo k' is reported to decompose completely at 200°C, and MoF _ is re- portedly not stable at the temperatures at which we have observed it. One may, therefore, infer that the free energies of formation of these fluorides per fluorine atom are so nearly equal that kinetic factors are expected to dominate the descriptive chemistry of these substances. 146 ORNL~DWG 67-10342R FOOQ rormrmrmsmrssn oo s M02F9+ ISOTOPIC PATTERN AT 125°C 100 - 10 |— w },— T & L T 7 I { b d i JJ 1 ‘[ }‘ Ll a {100 ‘ I T T LéJ CALCULATED ISOTOFPIC PATTERN 2 FOR MopFy! w o {10 — . - Q.1 [ """" ! ! | 355 360 365 370 375 amy Fig. 11.3. Comparison of Calculated and Observed Isotopic Abundance for the Dimer M02F9. 147 -3 ORNL-DWG 67-103HR 10 Tt T T T i TR T fm \A,‘" BEYER Z0min EXPOSURE TO M002F-+ LLABORATORY AIR AT 25°C MoOF S Mo, FF 1g7* | T=525°C '\.400+ MoOF - MOF MOF? 5 L 1678 | e e S JOT2 e e — i T T : w’\/\“““ T 7 wy i 0 r=125° MoF:* . ; wrg - - oL g - MoF} MoOFy s o = i + + 02'9 - MoF MoOF o + 1075 |- MoG MoFe ] | fl Il r T 1OO HO 1¢O 130 440G 150 160 170 180 490 355 3ba amu Fig. 11.4. Mass Spectra of Sample 11l (MoF5) at 125°C (Note Dimer, M02F9+) and at an Elevated Temperature After Exposure to Air. Table 11.2. Mass Spectrometric Cracking Patterns for Yarious Molybdenum Compounds MoF MoF MoF 7 MoOF () MoO,F MoF,* <7 MoF," 100 MoF,' 100 MoOF," 100 MoO,F," 100 MoF,' 100 MoF " 15 MoF ,* 30.4 MoF 6 MoO,F 65 MoF, " 56 MoF 10 MoF ,* 17.5 MoOF " 10+ MoOF ,* 6 MoF," 18 Mo 10 MoF )" 13.1 MoF " 11 Mor 2 2 Mo 8 MoF 6.9 MoOF ¥ 10 MoOF * 23 Mo® 8 Mo” 4.5 MoF 9 MoF * 5 Mo 3.5 MoO 7 Mo 5 Mo 16 “From J. C. Horton, ORGDP. 12. Separation of Fission Products and of Protactinium from Molten Fluorides 12.1 EXTRACTION OF PROTACTINIUM FROM MOLTEN FLUORIDES INTO MOLTENMN METALS J. H. Shaffer D. M. Moulton F. F. Blankeaship W. R. Grimes The removal of protactinium from solution in IL.il- Bel’ -ThE, (73-2-25 mole %) by addition of thorium or beryllium metal has been demonstrated as a method for reprocessing the fertile blanket of a two- region molten-salt breeder reactor.! This investiga- tion has been directed toward the development of a liquid-liquid extraction process where reduced *33Pa would redissolve into a molten metal phase for sub- sequent back extraction by hydrofluorination into a second storage salt mixture. Upon #s radiolytic de- cay, fissionable 23U could be returned to the reactor fuel via fluoride volatility. Earlier experiments on static systems where the simulated blanket salt, in contact with bismuth, was contained in mild steel further substantiated the re- duction of protactinium from the salt mixture but failed to demonstrate its quantitative dissolution into the metal phase. Since the extraction method utilizes the molten metal only as a transport medium, the ap- parent very low solubility of 2*3Pa or that of its carrier in molten metal is not necessarily detrimental to the process. A dynamic system was constructed and was operated by recirculating bismuth containing dissolved thorium through the simulated blanket mix- ture.? Recovery of protactinium on a steel wool column in the bismuth circuit accounted for about 43% of the ?33Pa initially in the salt phase. How- 1MSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, p. 147. ‘MSR Program Semiann. Progr. Rept. Aug. 31, 1966, ORNL-4037, p. 148, 148 ever, the recovery of 2*3Pa was predominantly by filtration rather than by absorption. These filtered solids showed that ?33Pa was associated with both iron and thorium. More recent experiments have been designed to demonstrate the reductive extraction of 233Pa from the blanket salt mixture into molten metal and its back extraction by hydrofluorination into 2 second salt mixture. This apparatus, shown schematically in Fig. 12.1, contained all liquid phases in graphite to preclude the indicated absorption of 2?3Pa on solid metal surfaces. The initial experiment of this series demonstrated essentially complete removal of 233Pa from the blanket; only 65% of the ?33Pa was found in the recovery salt. The loss of ?3*Pa in this experiment and in a later experiment was attributed to its absorption on metal particles formed by re- duction of iron impurities in the blanket mixture. The solubility of iron in bismuth at 650%C is about 90 ppm.> Thus, the addition of a second component to the metal phase which would increase the solu- bility of iron might promote the ?33Pa extraction process. The solubility of iron in tin at temperatures proposed for the extraction process is in excess of 1000 ppm.* Since tin is completely miscible with bismuth below its melting point of 271°C, it was chosen as an additive for bismuth in a recent ex- periment. In this experiment approximately 1.78 kg of the simulated hlanket salt, I_,il?"-Ber-ThF4 (73- 2-25 mole %), with 1 mc of 2?3Pa was in contact with 2.98 kg of bismuth which contained 0.41 kg of tin. Approximately 0.533 kg of LiF-BelF, (60-40 mole %) was also in contact with the molten metal mixture as 3_]. R. Weeks et al., Proc. U.N. Intern. Conf. Peaceful Uses At. Energy, Geneva, 1955 9, 341 (1956). 1. A. Kakovskii and N. S, Smirnov, [zv. Akad. Nauk SSSR, Otd. Tekhn. Nauk 1957(11), pp. 44—51. 149 ORNL.-DWG &67-11821 ,_ £ o SAMPLE AND Th = ADDITION PORT e SAMPLE PORT . 0O o 1 He & TS oy oFFeas—9 HT L —— OFFGAS — o i T - BLANKET & ‘1. RECOVERY SALT - .l : SALT AL MATERIAL IN CONTACT WITH SALT AND METAL PHASES 1S GRAPHITE 233 Fig. 12.1. Extraction Yessel for Pa Removal from Molten Fluorides. a recovery salt. The results of 233Pa transfer during reduction by Th? additions to the blanket salt and hydrofluorination of the recovery salt are shown in Fig. 12.2. Approximately 90% of the 233Pa ac- tivity was found in the recovery salt mixture, 2% wass present in the molten metal phase, and 8% re- mained in the blanket salt. Thus, this experiment demonstrates quantitative accountability of ?33Pa by analyses of samples taken from the three liquid phases. Subsequent experiments will further eval- uate the effects of tin, as a metal phase additive, on the extraction process, 12.2 STABILITY OF PROTACTINIUM-BISMUTH SOLUTIONS CONTAINED IN GRAPHITE D. M. Moulton J. H. Shaffer Previous experiments on the extraction of #3%Pa from Li E‘—ThF4-BeF2 into Bi by Th or Li have re- ORNL--DWG 8714822 100 -~ 80 ] ] &~ 1 | o BLANKET SALT T \ & RECOVERY SALT o ® METAL PHASE o 60 |— L‘% H 5 | a = 3 BLANKET: 1.78 kg ~ 2, RECOVERY: 0.533 kg = o N METAL: 2.98 kg Bi = o A0 | L O AND 0.408 kg Sn — = T fij azz_ . = v z Pal~ {mc & T ooy g £ o oy - 200 N 10 15 25 THORIUM METAL ADDED (g) Fig. 12.2. Extraction of 23%Pa from LiF-BeF,-ThF, (73-2-25 mole %) by Reduction with Thorium Metal into LiF-BeF, (60-40 mele %) by Hydrofluorination via Molten Bismuth-Tin Mixture at $50°C, sulted in an apparent disappearance of Pa from the system. It seems as though protactinium is stable as an ion in the salt, and only begins to act peculiarly when it enters the liquid metal phase. We accordingly carried out an experiment in which thorium irradiated to produce 233Pa was dissolved in and kept in bis- muth in an all-graphite apparatus. Data obtained in this experiment are shown in Fig. 12.3, upon which the 233Pa decay curve has been superimposed. The solution of 1 me of protactinium with carrier thorium showed an initial ?**Pa count about one- third of that expected on the basis of 1-mc solutions made previously in salt. The protactinium stayed constant (though with wide scatter) for 16 days at 600 and 750°C with mote thorium addition. When the metal was cooled to 350°C so that ~98% of the tho- rium should precipitate, the protactinium fell (after an unexplained one-day lag) to ~12% of its former value. Reheating to 700°C and then to 725°C {where all of the thorium should redissolve) brought the prot- actinium back to only 30% of its initial value. When salt was added, protactinium did not appear until enough BiF, had been added to oxidize half the thorium added; then 82 to 86% of it appeared in the salt phase. In fact, protactinium should not have appeared until practically all the thorium was oxi- dized, so there must bave been about a 50% reductant loss. 150 5 ORNL-DWG 87-11823 ol 7o — S e ] s00C° \ 750° ‘ 350° ‘700" > counts min ! (g of B! L [ { - e — e e — L] STATIC J ® — 0 AGITATED A AVERAGE A SALT {(AS Bi EQUIVALENT} Q A0 20 3G 40 50 60 70 TIME (days) Fig. 12,3. Stability of Solutions of Protactinium in Bismuth, Behavior of this sort cannot be explained by a vigorously, suggestiog that the protactinium was on simple mechanism. However, slow reversible ad- a dispersed phase rather than on the vessel walls. sorption (negligible at high temperature) followed by Strips of various metals were immersed for 5 min irreversible diffusion into a solid (but oxidizable) and 2 hr in the bismuth to see how much protactinium phase is consistent with the data. This phase does was adsorbed; results are shown in Table 12.1. When not seem to be the thorium-bismuth intermetallic. the metals were dipped directly into bismuth, niobium After the cooling step, the metal samples, which were picked up the most protactinium, with beryllium, iron, taken with an open dipper, tended to show more prot- and carbon picking up less (carbon, much less). Ad- actinium when the liquid metal was being stirred sorption on iron reached about one-third of its 2-hr 151 Table 12.1. Adherence of Protactinium to Metal Samples at 725°C Pa on Metal Bi Equivalent to Metal E'Kplr)sure Are; (counts/min) Pz in ?11 » Counts of Pa (g) Time (cm®) T T T (counts min™ g~ ) T Total Per om Total Per om? x 104 x 10% x 103 Be-1 2 hr 5.52 6.27 1.14 5.9 11 1.9 Nb~1 2 hr 5.22 15.4 2.95 5.8 27 5.1 Fe-1 2 hr 5.52 2.84 0.514 5.2 5.5% 0.99 Nb-2 2 hr 4.34 11.2 2.58 5.1 22° 5.1 Fe-2 5 min 5.52 0.796 0.144 4.9 1.6 0.29 Fe-3 2 hr 5.52 2.59 0.469 4.9 5.37 0.96 C-1 2 hr 11.94 0.170 0.014 4,7 0.36 0.03 After C-1, a salt layer of 72.5 mole % LiF and 27.5 mole % BeF2 was added to the experiment, Metal samples were held in this salt for about 2 hr to remove surface oxides hefore ex- posure to the bismuth. Nb-3 2 hr 4.34 0.292 0.0673 3.4 0.86 Q.20 Nb-4 5 min 4.34 1.07 0.247 3.0 3.6 0.82 Nb-5 2 hr 4.34 0.617 0.142 3.0 2.1 0.47 Fe-4 2 hr 5.52 3.66 0.663 3.3 11 2.0 Weight gain (adhering bismuth) less than 1 g. Pa counted differentially at 305 to 315 kev. value in 5 min. When the metals had been exposed to salt to remove oxides, the amounts of adsorption were similar; but now, niobium picked up less protactinium than did iron, and the process was much faster, In all cases the protactinium adsorbed was not a large amount of the total. 12.3 ATTEMPTED ELECTROLYTIC DEPOSITION OF PROTACTINIUM D. G. Hill®> H. H. Stone A number of unsuccessful attempts to deposit protactinium electrolytically from molten fluoride mixes have been reported.®’” None of these experi- ments involved the use of a standard reference elec- trode such as the hydrogen—hydrogen fluoride couple.® Summarized here are an experiment to *Cons ultant, Duke University. 6C~ 1. Barton, MSR Program Semiann. Progr. Rept. Aug. 31, 1966, ORNL-4037, p. 156. c. J. Barton and H. H. Stone, MS5R Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4119, p. 153. IqGr, Dirian, K. A. Romberger, and C. F. Baes, Jr., Re- actor Chem. Div. Ann. Progr. Rept. Jan. 31, 1965, ORNL.-3789, p. 76. measure the deposition potential for thorium by using an Hz-HF anode and a nickel cathode, and an unsuc- cessful attempt to deposit protactinium with the same electrolytic cell. A nickel pot was fabricated for these experiments, The salt compartment had a diameter of 2.5 in, and was 8 in. high. It was fitted with two 0.5-in.-ID chimneys welded to the top of the pot and extending 9 in. above the pot, high enough above the furnace to permit the use of Teflon insulators for the electrodes or electrode supports. One of the chimneys also ex- tended to 1 cm above the bottom of the pot, and it was closged at the lower end by porous nickel filter material. Thus anode and cathode compartments wete isolated in such a manner that bulk diffusion would be very slight, but ions could migrate between elec- trodes. The pot was loaded with 1556 g of purified Lil- Bel" -ThF, (73-2-25 mole %). We calculated that the material would have a volume of 275 cm?® at 600°C and would fill the pot to a depth of 9 cm at this temperature. Initially, % -in. nickel tubing was inserted in both chimneys so that gas could be passed through each compartment. A cylinder of platinum gauze closed at the bottom and extending 1.5 cm below the end of the tube was tightly wired 152 to the end of the nickel anode. A mixture of H and HF that passed through this electrode estab- lished the reaction at the anode as 1/2H2+F_:HF+8—. The cathode reaction was the deposition of any of the cations present, particularly thorium at above its deposition potential, and hopefully protactinium after it was added to the melt. For the preliminary study of the decomposition po- tential of ThF4, the vessel was maintained at 650°C while the cathode compartment was swept for 3 hr with HF and then overnight with H,. The cathode coinpartment was capped both at the inlet and the outlet, and a stream of HF-H_ was introduced through the anode tube. The gas flow was adjusted to approximately 100 cm?®/min. The mixture was ' analyzed at the exit tube and was found to contain 3.8 mole % HF, Electrolysis was performed with dry cells as the source of power. The voltage was tapped off a slide-wire voltage divider, and simultanecus read- ings were taken of the voltage and the current. Four runs were made, starting each time at zero voltage and measuring the current at 0.2-v inter- vals. The plot of current vs voltage given in Fig,. 12.4 shows a decomposition potential in this cell of —0.86 v. The mole fractions of each of the reactants are re- lated by the equation RT [HF] E=FE - ln e 0 F [Hg]l/Z [T}lF4]1/4 4.59 {(923) ~-086=K -0 - o 2.3 «10* ORNL--DWG 67--118724 AOQ g -~ T . 75 . g /} é 50 /f— - 44 o5 | L _ ;/ T g A 0 e 8 e :‘,;__,_ 7“4%_J 0 05 1.0 L5 Ev) Fig. 12.4. Current-Voltage Curve. | (0.038) ) 0F e & (0.962)172 (0.25)t 74 from which E0 = —1.09 is calculated for the reaction 2H, 4 ThF, = 4HF ¢ Th. After completing these measurements the tlow of HF was cut off, while H, flow continued, After 1 hr, both electrodes were raised out of the melt, and the H, tlow was allowed to continue until the pot was cold. For the study of protactinium removal, the cathode compariment was opened with as much care as pos- sible to avoid the entrance of moisture, and to that compartment was added 5 g of LiF-ThF which had been irradiated in a reactor to give a 233Pa activity of about 1 mc. The salt also contained approximately 27 mg of ?31Pa. The vessel was closed and installed in the furnace in a glove box in the High-Alpha Molten- Salt Laboratory. After HF treatment for 4 hr, followed by H, overnight, a filtered salt sample was counted for 233pa_ The anode gas mixture was adjusted to be approxi- mately the same as that used in the previous run ex- cept that a slightly higher HF concentration, 7.5 mole %, was used. The decomposition potential for ThF, at this concentration of HI from Eq. (1) should be --0.92 v, so electrolysis was carried out at - 0.90 v. The current averaged about 10 ma, although it was less for the first 20 min, after which it rose rather rapidly to the 10-ma value expected from the earlier experi- ment. Assuming 100% Faraday efficiency, the depo- sition of 27 mg of protactinium would require 75 min at 10 ma. Electrolysis was carried out for 165 min, more than twice the minimum time for complete deposition. The gas at the end of the experiment contained only 4.1 mole % HF, although the H, rate was unchanged. There is no indication at what time during the elec- trolysis the decrease in HF occurred. At this lower concentration of HF the decomposition potential is calculated to be —0.87 v, so that if the decrease in rate took place early in the electrolysis period, the applied potential was too high. At the conclusion of the run the cathode was raised out of the melt, thus interrupting electrolysis. A sample of melt was then taken, and both cathode and sample were counted with a multichannel analyzer. The sample of melt gave essentially the same count as before electrolysis. A number of gamma peaks were observed in the cathode spectrum, but none corre- sponded to the ??3Pa peaks, which were strong in the melt sample. The location of the peaks and the limited data on their decay rates indicate that these activities can be attributed to various mewbers of the thorium decay series, such as 21%2Pb, ?12Rj, and 20871, although insufficient decay rate data were obtained for conclusive identification. The lack of 233Pa activity on the cathode, coupled with almost identical counts on samples of melt taken before and after electrolysis, indicated that deposition of a significant amount of protactinium on the cathode did not take place in this experiment. The failure to reduce protactinium in this experi- ment may be explained by its low concentration in the melt. The amount added, 27 mg in 1556 g of LiF-BeF -ThF,, is approximately equal to a mole fraction of 2 x 1073, If the protactinium is four- valent in the melt, and if it is assumed that its EG 15 close to that of thorium, that is, 1 v, the potential required to reduce it at that concentra- tion and at the HF-H_ ratio used in the electrol- ysis is 1.007 v. The very small denominator makes the logarithm term positive, so that the reduction potential is even more negative than Eo in spite of the low HIF pressure. The same argument would deny that protactinium can be reduced by thorium at this concentration, which is contrary to fact. The nicke! cathode did not provide a low-activity alloy in this attempt. This suggests the possibility of using a cathode that would form a thorium alloy in which protactinium dissolves at low enough activity to permit reaction. For further experiments of this type we recommend a potential of 1.1 v, at which thorium will be de- posited, hopefully, along with protactinium, and stopping the experiment when five equivalents of thorium have been deposited. This would give a concentration of about 0.2 for protactinium in the metal, and thus a low enocugh activity to permit reduction of essentially all the protactinium. 12.4 PROTACTINIUM STUDIES iN THE HIGH- ALPHA MOLTEN-SALT LABORATORY C. J. Barton H. H. Stone In the previous progress report’ we mentioned the possibility that iron coprecipitated with protactinium, when solid thorium was exposed to molten LiF-’_l‘hF’;, might be carrying the reduced protactinium to the steel wool surface in the Brillo process. It was also pointed out that variable iron analyses made it dif- ficult to determine the role of iron in protactinium re- duction experiments. We performed one reduction experiment with 5%Fe tracer dissolved in LiF-ThE, (73-27 mole %) in the absence of protactinium, and the results are summarized below. Most of the protactinium recovery experiments performed during the current report period were thorium reduction tests. An unsuccessful effort to deposit protactinium elec- trolytically is discussed in the preceding section of this report. Reduction of Iron Disselved in Molten LiF-ThF4 We studied the reduction of iron dissolved in molten LiF-ThF (73-27 mole %) by using hydrogen and metallic thorium as the teducing agents. The tracer iron tesults indicated that more than 40 hr was required to reduce the iron concentration from 330 to 2 ppm with hydrogen at a temperature of 600°C. During a 3-hr exposure to metallic thorium at the same temperature, the iron concentration in filtered samples, calculated from tracer counts, diminished from 550 to 13 ppm. Colorimetric iron determinations performed by two different laboratories agreed with the tracer iron data for about half the samples. In general, agreement was poorest for samples that gave °?Fe counts in- dicating an iron concentration less than 0.1 mg/g. It appears that the colorimetric iron method tends to give high results with samples having a low iron concentration. This finding is in agreement with results of earlier unpublished studies per- formed by Reactor Chemistry Division personnel. The results of this experiment will be discussed in more detail in another report. Thorium Reduction in the Presence of an lron Surface (Brille Process) One Brillo experiment of the type discussed in the previous report’ was performed during this report period. A salt solvent, LiF-ThF (73-27 mole %}, containing initially 17 mg of ?*'Pa and 57 mg of Fe was exposed to thorium rods for two 1-hr periods in the presence of 4 g of steel wool (0.068 m? per g of surface area). The tracers 233Pa and *°Fe were added to aid in following the behavior of these ele- ments in the experiment. The first thorium exposure removed 95% of the protactinium, as determined by analysis of a filtered sample of salt, and almost all the iron. The second thotium exposure resulted in only a slight further decrease in protactinium con- 154 centration (97% removed). The distribution of prot- 72.8% in the steel wool plus untransferred salt, 4.3% in the unfiltered salt transferred away from the steel wool, 8.7% in the steel liner and dip leg, and 12.0% in samples, for a total of 97 8% recovered. The data obtained in this experiment confirmed the resulis of previously reported” experiments that indicate the Brillo process may warrant further examination. actinium was as follows: Therium Reduction Followed by Fiitration We reported earlier® that a large fraction of the reduced protactinium that would not pass through a sintered copper filter was found in samples of un- filtered salt. This suggested the possibility of col- lecting the reduced protactinium on a metal filter from which it could presumably be removed by dis- solving it in a molten salt after passing HF through the filter. We performed several experiments to test the efficiency of protactinium recovery by filtration. The initial treatment of molten LiF—"l"I'lI*"4 (73-27 mole %) mixed with enough ?3'Pa to give a con- centration of 20 to 60 ppm was performed in unlined nickel pots. The molten salt was treated first with mixed hydrogen and HF, followed by a brief hydmgen treatment before it was transferred through a nickel filter into a graphite-lined pot equipped with a graphite dip leg. Here the reduction of protactinium was carried out in the usunal fashion, either by ex- posure to a solid thorium rod or to thorium turnings suspended in a nickel basket, taking samples of fil- tered and unfiltered salt after each thorium treatment of the melt. back into the nickel pot through the transfer filter. The reduced melt was then transferred In four experiments the amount of protactinium found in the transfer filter varied from 10 to 30% of the total amount present. This represented 40 to 95% of the amount of protactinium suspended in the reduced salt (average, 69%). The amount of prot- actinium in the graphite liner and dip leg varied from 20 to 57% of the total (average, 33%). The data show that a filtiation method will not catch protactinium on the filter, but nevertheless the re- moval of protactinium from a melt does appear fea- sible. One experiment was attempted with a niobium liner and dip leg. The reduction of protactinium with thorium proceeded normally (12% of the Pa re- mained in the filtered salt after 2 hr of thorium treatment), but transfer of the reduced salt through the filter could not be effected because of a clogged filter. A considerable amount of grayish material was found in the bottom of the pot after it cooled to room temperature. A sample of this material was reported to contain only 0.35 mg of Nb per g, but it is quite possible that this amount of impurity in the molten salt would have been sufficient to clog the filter. The effect of iron on the hehavior of protactinium in thorium reduction experiments has not been de- fined urambiguously as yet, but we continue to find reasonably good coirelation between the distribution of iron and protactinium in these experiments. Count- ing of 2?*Pa and *%Fe in both solid samples and solutions of samples provided a check on the ac- curacy of 2?!Pa alpha pulse-height analyses and colorimetric iron determinations. Conclusion Thorium metal is an effective agent for reducing protactinium in molten fluoride breeder blanket mixtures, but further study will be required to de- termine the best method of separating the reduced protactinium from the salt mix. 12.5 MSBR FUEL REPROCESSING BY REDUCTIVE EXTRACTION INTO MOLTEN BISMUTH W. R. Grimes J. H. Shaffer D. M. Moulton F. F. Blankenship An electromotive series for the extraction of fission products from 2LiF-BeF into liquid bis- muth has been constructed in the way described for the MSBR blanket materials in the preceding report in this series.® Briefly, standard half-cell reduc- tion potentials are calculated for each metal, using as the standard states the ideal solutions in salt and bismuth extrapolated from infinite dilution to unit mole fraction. The exception is lithium, for which the standard state in salt is 2LiF-BeF . (This standard state is also used for beryllium, but it is not assigned a standard state in bismuth be- cause of its very low solubility.) This choice of standard states is the normal practice when deal- ing with dilute solutions. These standard poten- tials are called 590' to distinguish them from 5’:‘0, IMSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNI.-4119, p, 150, the standard potential for reduction to pure metal. Thus, 86 is the potential corresponding to the ex- traction process and does not require the simulta- neous use of metal activity coefficients that are far from 1 at infinite dilution. Ii should perhaps be noted that the electromctive series is thermodynamically equivalent to other means of expressing the free energy of the extrac- tion process, such as equilibrium constants or de- contamination factors. It is used because it per- mits ranking the metals in an order which directly indicates their relative ease of extraction regard- less of their ionic valence and because it has proven useful in the experiments where a beryllinm reference electrode is used to monitor the reduc- tion process, From data previously reported!? the series has been constructed as shown in Table 12.2. The numbers in parentheses after the rare earths are the potentials calculated by using the experimental fractional valences. For beryllium, 80, the poten- tial for reduction to the pure metal, is shown. All of these except Be? * are calculated relative to the Li" value. Although better measurements may change these values somewhat, they indicate the relative ease of extraction. The numbers in parentheses are probably a better approximation to the true values despite the peculiar valences. All elements up through europium can be extracted completely be- fore metallic beryllium begins to form; this forma- tion of Be? represents the ultimate limit of the process. An order of magnitude change in the ratio [(mole fraction in metal)/(mole fraction in salt)] corresponds to 0.087 v for divalent and 0.058 v for trivalent species. Table 12.2. = at 600°C (volts vs Hy—HF = 0.00) Li 1.02 Bel® 1.81 &) Bal? 1.79 pu?t 1.62 (1.61) Na3t 1.58 (1.51) sm2t 1.58 (1.49) ce?? 1.87 (1.45) La3? 1.52 (1.47) 155 We have begun a reinvestigation of fission product extraction. In this study we will measure fission product distribution as a function both of [ithium concentration and of temperature with (hopefully) improved accuracy. Preliminary resulis of the first of this series, using cerium, are shown in Fig. 12.5. A failure ended the experiment hefore pood concentration data could be gotten, but those we did get indicate a valence close to 3, not 2.3 as before. The apparent minimum of E‘(; (Ce) at 700°C is not explained and is rather hard to believe. For lithium, ‘r.}f(; lithium concentration and the potential between the bismuth pool and beryllium metal electrode, for which the temperature coefficient iz known. "1 The was calculated from the measured PSR Program Semiann. Progr. Rept. Feb. 28, 1966, ORNI.-3936, p. 141. 11 .. .. ; C. ¥. Baes, Jr., Reactor Chem. Div. Ann. Progr. Rept. PDec. 31, 1965, ORNL-3913, p. 20. ORNL DWG €7~ 14825 2-1 } 20 |— L“W F S E U A Li FROM Be - EQUILIBRIUM \\. . CT >\k ® i Li HIS WORK - ARGONNE 18— —— D > AT —— ] e g 16 L e e S o -~ \ - P s -~ - e T - 400 500 00 T00 800 500 7 (°C) Fig. 12.5. &’ for Lithium and Cerium. Q ’ point marked with a cross is the value of 80(1,1) found earlier by equilibrating metallic beryllium with Li-Bi amalgams. Its distance from the line in this experiment may be the result of slightly in- creased lithium activity at higher concentration | X -0.16]. LB E‘,‘.(; for lithium calculated from the data of Foster The line marked ‘‘Argonne”’ is and Eppley'? on the excess chemical potential of Li in Bi. The rather small uncertainty here prob- ably does not matter much in the separation process, but it is important, in preparing the elec- tromotive series, to refer all potentials to the same The potentials at 600°C measured in this experiment are - 1.89, -1.81, and —1.46 for Li, Be, and Ce respectively. The extraction coetficients for the reaction are given in Table 12.3. This method of presentation shows more clearly than the potentials the sub- lithium or beryllium potential. stantial decrease in extraction with rising temper- ature. The difference between this value of Q at 600°C and that reported earlier by Shaffer!? is mainly due to his use of 2.3 rather than 3 as the exponent. The extraction equilibria look very favorable for using this process in MSBR fuel reprocessing. Re- covery of ‘Li from the bismuth will be helpful and should not be difficult. A perplexing problem which has occurred frequently is the apparent loss of total reductant species during the course of the extraction. The extent of this loss is not easy to measure because of uncertainties in the lithium analyses, but averages 25 to 50% of the lithium added. We think we have eliminated the possi- bility that the loss is due to exiraneous reactions such as lithium volatilization, dissolution of Li, Be, or Bi’~ in the salt, or formation of solid beryl- lium or beryllides. In the Zr-U separation, the usual experimental technique was changed to min- imize the possibility of air coming in the access ports (normally filled with hot helium) during addi- tions or sampling. The reductant loss was cut sharply to ~10%, which is within the accuracy of the measurements. In another experiment we ob- served a substantial increase in oxide content over the initial value; because of probable BeO precipi- tation, the amount of oxygen added could not be 12M. S. Foster and R. Eppley. Chem. Engr. Semiann. Progr. Rept. July—Dec. 1963, ANL-6800, p. 405, 13]. H. Shaffer, MSK Program Semiann. Progr. Rept. Feb. 28, 1966, ORNL-3936, p. 144, 156 Table 12. 3. Extraction Coefficient for Cerium Extraction intc Bismuth XCe(m) 4 X =Xy i Ce(s) Q T (C) 9.1 % 108 456 4.0 108 545 8.7 % 107 600 2.1 %10 697 2.1 x10° 817 determined. A few back oxidations with a meas- ured amount of BiFs have given more or less 100% efficiency, indicating that no unknown reduced species takes part in the reoxidation step. It seems, therefore, that at least a substantial part of the reductant loss is due to air entry and can be eliminated by improved experimental techniques. Since there is no evidence that any other reduction process is going on, we do not feel that this reduc- tant balance anomaly in the laboratory experiments will have any important bearing on the process when it is put on a practical scale. 12,6 REDUCTIVE EXTRACTION OF CERIUM FROM LiF-BeF, (66-24 MOLE %) INTO Pb-8i EUTECTIC MIXTURE W. J. Hunt'? W. . Bull'® J. H. Shaffer Current investigations on liquid-liquid extraction as a method for reprocessing the MSBR fuel are di- rected toward the use of bismuth-lithium mixtures for removing fission products from the reactor fuel solvent. A complementary program will survey other metal phase extractants for possible adapta- tion in the reprocessing method. This proposed experimental program will tieat the distribution of cerium, for reference purposes, between LiF-Bek, (66-34 mole %) and various molten metal mixtures 14 . . ORAU summer student from University of Missis- sippi. 15 . . Consultant, University of Tennessee. 2 as functions of temperature and metal-phase com- position. Experimental procedures will essentially duplicate those developed for the primary program. The results presented here on the reductive extrac- tion of cerium from LiF-BeF (66-34 mole %) into the Bi-Pb eutectic mixture (56.3 at. % Bi) are the first of this experimental program. As in previous related experiments, the distribu- tion of cerium between the two liquid phases was followed as a function of the lithium concentration in the metal phase. However, only small fractions of lithium added incrementally to the metal phase remained in solution under equilibrium conditions. Typical results obtained at 700°C (Fig. 12.6) show that only 32% of the cerium could be removed from the salt phase. In addition, analysis of the metal phase could account for no more than 1 to 2% of the cerium activity. Variations in temperature from 500 to 700°C had little or no effect on either the distribution of cerium between the two phases or on the lithium concentration of the metal phase. Attempts were made initially to measure the electrical potential between a beryllium metal electrode inserted in the salt phase and the molten metal pool. Despite short contact periods of the Be® rod with the salt mixture, substantial quanti- ties of cerium activity accumulated on the beryl- lium electrode. A large fraction of cerium activity missing from the two liquid phases collected on the beryllium electrode during its accidental over- night exposure to the salt mixture. 157 ORNL~-DWG 67 —-11826 12 oo e I [ | i [ i o ANALYSES OF SALT PHASE 2 o ANALYSES OF METAL PHASE - o MO e WEIGHT OF SALT PHASE: 2336 kg ] 2 WEIGHT OF METAL PHASE: {526 kg Bi 2 1474 kg Pb & ooeN —_— < £ 5 % B e \ _‘;;‘s_._“ ................ 4 8 e O, o o Z i_??- 4 — i ———— o3 £ = 2 E 21— . O O SR R (&) ® ¢ o """ * O { 2 3 4 5 6 LITHIUM ADDED TO METAL PHASE {(mole fraction xv102) Fig. 12.6. Reductive Extraction of Cerium from LiF- BeF2 {(66-34 mole %) into Molten Lead-Bismuth Eutectic Mixture at 700°C. Although the results of this experiment are some- what incoanclusive, they indicate that solid rare- earth beryllides might be stable in this extraction system. Because of this rather pronounced effect of lead, the system Pb-Bi can probably be excluded from use in the liquid-liquid extraction process for rare-carth removal from the MSBR fuel solvent. 13. Behavior of BF; and 13.1 PHASE RELATIONS IN FLUOROBORATE SYSTEMS C. J. Barton J. A. Bornmann' I.. 0. Gilpatrick H. Insley? T. N. McVay? H. H. Stone Interest in low-melting, low-cost molten salts for use as the coolant in molten-salt breeder reactors has prompted reexamination of phase rela- tions in fluoroborate systems. The system NaBF - KBP‘4 was studied earlier at this Laboratory, * while Selivanov and Stender* reported low-melting eutectics in the systems I\IaF‘—NaBF4 and KF- KBE . More recently, Pawlenko® investigated the ternary system KBE -KF-KBF ,OH and its as- sociated binary systems. Varying melting points reported in the literature for the compounds NaBF and KBF indicated the need for pure compounds and care in heating them to prevent hydrolysis or loss of BFs' We found that recrystallization of commercial preparations of NaBF , and KBF , from dilute hydrofluoric acid solutions, usually 0.5 M, gave compounds with higher melting points than any previously reported. We find 569°C for KBF (previous high, 552°)° and 406°C for NaBF (previous high, 368°).* The differential thermal analysis (DTA)® melting curve for K]BF4 was much 1 .. Research participant from Lindenwood College, St. Charles, Mo. 2Censultant. 3R. E. Moore, J. G. Surak, and W. R. Grimes, Phase Diagrams of Nuclear Reactor Materials, R. E. Thoma (ed.), ORNL-2548, p. 25 (Nov. 6, 1959). 4V. G. Selivanov and V. V. Stender, J. Inorg. Chem., USSR, IK2), 447—-49 (1958). ’S. Pawlenko, Z. Anorg. Aligenm. Chem. 336, 17278 (1965). ®L. 0. Gilpatrick, R. E. Thoma, and S. Cantor, Reactor Chem. Div. Ann, Progr. Rept. Dec, 31, 1966, ORNL- 4076, pp. 5—6. 158 Fluorcborate Mixtures sharper than that of NaBK ; this probably indicates higher purity of the potassium compound, but repeated crystallizations of NaBI*‘4 from dilute HF solutions failed to give a preparation having a ) melting point above 406°. We are considering . alternative methods of purifying this compound. The principal techniques used in the investiga- tion of phase relations in fluoroborate systems were DTA® and quenching.” By use of these techniques we have examined the binary systems NaF-NaBF“ and KF-KBF | as well as the compound joins NaBF -KBF and NaF-KBF in the ternary system NaF-KF-BF .. Diagrams for these systems are presented in Figs. 13.1 to 13.4. All except NaBF -KBF are simple eutectic systems, although a small thermal effect at 435°C in KBF -rich NaF- KBF, mixtures suggests that the compositions used were slightly off the compound join and that a small amount of NaF-KF-KBF , eutectic liquid was formed. Large thermal effects due to the KBF , phase transition (283 t 2°C) and the NaBF , transition (243 * 2°C) were noted in the DTA curves. Efforts to retain the high-temperature forms of the pure compounds and their mixtures with NaF or KF by rapid quenching were unsuccessful. The phase diagram for the NaF-NaBl system (Fig. 13.1) indicates that the previously reported data on this system®* showing a eutectic melting at 304°C and containing 37.5 wt % Nal3F is grossly in error. The most recently reported work® on the KF-KBT | system agrees with our data (Fig. 13.2) on the location of the eutectic composition, but we found that the eutectic temperature was 14° higher than reported. We found no evidence of the com- pounds KF - KBF, and KF - 2KBF, reported® in this system. The system NaF-KBF, (Fig. 13.3) has apparently not been studied by other investigators. The diagram for the NaBF -KBF, system (¥Fig. 13.4) . J. Barton et al., J. Am. Ceram. Soc. 41(2}, 63 (1958). ORNL-DWG 67 - 9425A e e e MPERATURE {°C) TE NaBF maie % NaBF4 Fig. 13.1, The System Naf-NaBF ,, ORNL- DWG 67- 3474 800 | T N } - — TEMPERATURE {°C) mote % KBF4 Fig, 13.2. The System KF-KBF4_ indicates that the high-temperature modifications of NaBF | and KBF ,, believed to be cubic, form a complete series of solid solutions exhibiting a shallow minimum (392°C) near 15 mole % KBF ,. Subsolidus relations are not clear at present. A thermal effect at about 190°C is evident in all mixtures in the system, while other subsolidus effects vary with composition. Because of the previously mentioned difficulty in guenching in the high-temperature forms of the compounds and their mixes, resolution of the subsolidus relations in this system may require detailed investigation with a high-temperature x-ray diffractometer. We have started to map the phase boundaries and isotherms in the region of the ternary system bounded by the compounds NaF-NaBF -KBF -KF. 159 600 | —— | — . 500 |- .. TEMPERATURE (°C) - ol 5 O j@] O O o : | | —————— & i | ] L 200 1 [ SR . Nof 20 40 &0 80 KBF, mole 7% KBF4 Fig. 13.3. The System NaF-KBF . ORNL-DWG 67-9426 500 T T T T T > DTA 1.1GUIDUS 1 550 « OTA SOLIDUS S & QUENCH LIQUIDUS \ T oo i & QUENCH 30LIDUS L) & 2 ’ { g 450 e o \ e : st - L : ~I = 400 | J : fi \ : F.. ! L 350 ] e Lo Lo -:!iOO el l.! N [ _J {J AL_L. KBk, 20 40 O 20 NoBiy £ maole % MNadh, Fig. 13.4. The System NaBF ,-K8F ,. Only a few compositions in addition to the two compound joins (Figs. 13.3 and 13.4) have been examined to date. Compositions in the ternary system containing more than 50 mole % BF probably exist in equilibrium with very high pres- sures of BF3- 13.2 DISSOCIATION YAPOR PRESSURES IN THE NaBF ,-NaF SYSTEM Stanley Cantor To determine BF dissociation pressures from salt mixtures composed of sodium fluoride and sodium fluoroborate, a static method is being used in which the melt and its vapor are essentially enclosed in a metal-glass system. The apparatus ORNL-DWG &7-11827 VACUUM =—x}= TO TAYLOR PRESSURE TRANSMITTER s =——x} GAS SAMPLING STATION THERMOCOUPLE WELL~ | Hg MANOMETER — . 7 L__I PRECISION ||~ PRESSURE ||~ GAUGE clebriele el bbb (it FURNACE Fig. 13.5. Apparatus for Measuring Decomposition Pressure of Fluoroborate Melts, is shown schematically in Fig. 13.5. The melt, contained in nickel, exerts its vapor pressure on a mercury manometer and on diaphragm gages. To remove adsorbed gases, the vessel, containing a weighed salt sample, is initially evacuated at 100 to 125°C for at least 15 hr, Then to release gases encapsulated within the salt crystals the sample is heated about 25°C above its liquidus temperature. The sample is subsequently cooled down to room temperature. If the apparent pressure is greater than 0.5 mm at room temperature, the system is reevacuated and then reheated and cooled until the vapor pressure at room temperature is essen- tially nil, After this desoibing procedure, the cbserved vapor pressures should be caused only by the dissociation NaBF (1) - BF ,(¢) + NaF(0) . Gas samples collected during the vapor pressure measurements and analyzed mass spectrometrically confirmed that the vapor was virtually pure BFs. Usually about 1% SiF4 is observed in the vapor, this probably originates from the reaction of desoibed HI" with the glass parts of the apparatus. After completing vapor pressure measurements, the containment vessel is cut open and the contents ORNL DWG 67-11828 10,000 - - P S 2 ~—e— MEASURED o // 1 T ——o— EXTRAPOLATED P 5000 - » | A gy 4 - T S L o — | 0 1'/ | ¢ /O/ //? e o e _ P 2000 399/-@ //,9 /o/o // / — ~ T ! 839?/ s %/ | f 1000 - T - ,;,/ ' Q * - ! ‘ . . = ‘/i 1 ¥ 500 !/ ?7/59./: e / 8 , ' £ | | . 7 s . i ‘U: / ,(OO/' ./. / . . x : ! 1 > 200" ‘ a/‘ - i ; / 660., / A . . [ ’ : ; , " oot | 9/ ot 10 : '/ O ' . , . . © ; , . . . . / ; /g o /j 50 . . - O . ! a ® . ; . , S ) ‘/l ‘/. | o4 20 2. For this reaction, AG?OOO, the standard free energy at 1000°K, is estimated to be ~14 1 20 kcal. The following table gives the free energies of formation: Reference or Method of Estimation From BFS dissociation equilibria By analogy with formation data of Na3A1F6 JANAYF Thermochemical Tables Based on estimate of CrzB by K. E. Spear, Metals and Ceramics Division 163 The two other experiments, one with Hastelloy N and the other with iron and molybdenum, were not monitored for vapor pressure. After desorbing the gaseous impurities from the sample charge (metal plus salt), the vessels were brought to the tempera- tures shown in Table 13.1 and were kept there for the indicated time. After completing these two experiments the samples were examined; very little or no visible attack was apparent on the metal specimens. Also, there were no visible color changes in the salt mixtures. However, the weight losses in the Hastelloy and in the iron specimens (sce Table 13.1) suggest the need for more thorough investigation of the corrosion reactions of chromium and iron with fluoroborate salt melts. Apparent Mass Tronsfer of Nickel After completing vapor pressure measurements on each fluoroborate mixture, the nickel vessels are always cut open for examination of the con- tents. In almost every case, the top surface of the solidified salt contains a small amount of black material. In all cases the inner metal surfaces are shiny. The top portions of two salt cakes (one 97.5-2.5 mole %, the other 65-35 mole % NaBF - Nal') when dissolved in water yielded silvery residues. These residues were ferromagnetic, and, by z-ray diffraction, were identified as nickel metal. The reason that nickel metal particles appeared on the melt is not readily evident. It is unlikely that metal particles were present in the vessel prior to loading with salt, nor is it likely that the stock salts contained nickel metal. The luster of the walls in contact with salt suggests that nickel may have been mass transferred. Further investiga- tion should provide additional information on this unexpected deposition of nickel metal on the salt. 13.4 REACTION OF BF , WITH CHROMIUM METAL AT 650°C J. H. Shatfer H. ¥. McDuffie Current plans to replace the secondary coolant of the MSRE with a fluoroborate mixture have prompted studies of the compatibility of these materials with structural metals of the reactor system. Since fluoroborates of interest to this program exert measurable vapor pressure of 1BF3 at operating temperatures, covering atmospheres containing equivalent concentrations of BE, must be maintained in the free volume of the pump bow! of dynamic systems or used for gas sparge opera- tions. This experimental program will examine the reactions of BF | with various structural metals and alloys that might be applicable to the program. Preliminary results obtained by contacting BF with chromium metal at 650°C ate presented here. The experimental method utilizes a 30-in. length of 2-in. IPS nickel pipe, mounted horizontally in a 3-in. tube furnace, as the reaction chamber. The reacting gases are admitted through a penetration in the end plate that is welded to one end of the nickel pipe. A sheathed thermocouple also pene- trated the end plate and extended into the central region of the heat zone. The other end of the reac- tion chamber, which extends some 10 in. out of the tube furnace, is closed by Teflon in a threaded pipe cap. The gas manifold system provides for the introduction of helium, BFs, or mixtures of these gases at known flow rates into the reaction chamber, The system is sealed from the atmosphere by bubbling the gas effluent through a fluorocarbon oil. Concentrations of BF _ in helium can be determined continuously from the recorded signal of a calibrated thermal conductivity cell. Metal samples or speci- mens are carried in nicke! boats inserted through the threaded access port. The reaction of BF, with chromium metal was followed by periodic determination of weight gain of about 10.88 g of prepared chromium flakes during a 60-hr reaction period at 650°C. The chromium metal was prepared by electrolytic deposi- tion on a copper sheet which was subsequently dis- solved by an acid leach. The thia film of chromium metal which remained was broken into smal] flakes for this investigation. Reaction periods commenced by introducing BF | at 1 to 2 ce/min after heating the sample to 650°C in flowing helium. The sample was also cooled to room temperature under flowing helium to permit inspection and weight gain deter- minations, As shown by Fig. 13.8, the weight gain of the chrominm sample showed a linear dependence on the square root of reaction time., The overall reac- tion period showed that the chromium sample in- creased its weight by about 4%; there was no significant difference in the weight or appearance of the nickel boat during this experiment. Examina- tion by x-ray diffraction techniques showed that the chromium sample contained substantial quantities of Crz()s; minor fractions of the mixed fluoride, ORNL-DWG &7—14830 044 I | T INITIAL 0.40 ) WEIGHT GAIN OF REACTANT { \ 8 | 4 1 SQUARE ROOT OF REACTION TIME (hr}/2 O 1 2 3 5 Fig. 13.8. Reaction of BF 3 with Chromium Metal at 650°C. CrF, « CrF , were identified by petrographic tech- niques, The material is currently being chemically analyzed for boron. Although these results are inconclusive with respect to identification of the oxidation species in the gas phase, they do illustrate that the direct use of chemically pure, commercially available BF3 in high-temperature systems will promote the oxidation of nearly pure chromium. Further studies will evaluate the effects of BF _ concentrations on oxidation rates and will investigate methods for improving the purity of BF . 13.5 COMPATIBILITY OF BF, WITH GULFSPIN-35 PUMP OIL AT 150°F F. A. Doss PP. G. Smith J. H. Shaffer The proposed use of a fluoroborate mixture as the secondary coolant in the MSRE will require that a covering atmosphere containing BF3 be maintained above the salt in the pump bowl. Since BF is 164 known to catalyze the polymerization of certain organic materials, its effect on the lubricating properties of the pump oil needs evaluation. Although these effects will be observed directly during the planned operation of the PKP-1 loop with a fluoroborate salt mixture, a preliminary ex- periment is in progress to determine relative polymerization rates of Gulfspin-35 oil under con- ditions which can be related to actual pump opera- tions. The degree of polymerization should be indicated by measured changes in oil viscosity during the experiment. MSRE-type pumps use a helium purge down the pump shaft to isolate the lubricated parts of the pump assembly from the gas environment of the pump bowl. Previous tests on the prototype pump loop, using 3 Kr as an indicator, showed that this isolation technique reduced the concentration of pump bowl gases at the cil-gas interface to about 1 part in 20,000.% Accordingly, this current study provides accelerated test conditions by contacting the pump oil with helium containing about 1000 ppm of BF3 at a maximum operating temperature of 150°F. Two experimental assemblies have been operated concurrently to provide comparative data. In each assembly, helium was bubbled at a rate of about 1 liter/min through 1.5 liters of pump oil. The BF3 was introduced into the helium influent stream to one experiment at a rate of about 1 cc/min. Sam- ples of the oil were drained from each experiment periodically and submitted to the Analytical Chem- istry Division for viscosity measurements. As described in a later section, a continuous gas analysis system was installed and calibrated by A. S. Meyer, Jr., and C. M. Boyd of the Analytical Chemistry Division. Concentrations of BF , in the gas influent and effluent of the experiment are recorded from the output signal of a thermal con- ductivity cell. The light hydrocarbon content (products of o0il polymerization) of the gas effluent can also be monitored on a semicontinuous basis. The results obtained through approximately 600 hr of continuous operation of the experiments are illustrated in Fig. 13.9. Although some discolora- tion of the oil exposed to BF3 was noted, there is no distinguishable difference in oil viscosity between comparative samples. The increase in oil viscosity (original value, 15.7 centistokes) during 8 A. G. Grindell and P. G. Smith, MSR Program Semi- ann, Progr. Rept, July 31, 1964, ORNL-3708, p. 155. 25 % { (centistoxes ) VISCOSITY GF OIL AT W ORNL-DWG &7-11831 [ ! 1 l l 1 o-QIL CONTACTED WITH 8F 3 (~1000 ppm) a0 | IN HELIUM. *-0OIL CONTACTED WITH HELIUM ALONE. | VOLUME OF OiL IN EACH EXPERIMENT: 1.5 liters. GAS FLOW RATE IN EACH EXPERIMENT:1 liter/min. 0 100 200 300 400 500 600 TOOQ CONTACT TIME (hr) Fig. 13.9. EHect of BF 3 on Viscosity of MSRE Pump Oil (Gulfspin-35) at 150°F. 165 the experiment is no more than 0.3 centistoke at 25°C. Concentrations of BF, in the gas effluent have remained essentially the same as influent concentrations throughout the experiment. Hydro- carbon concentrations in the gas effluent were insignificantly low. On the basis of these results, the introduction of BF3 as a covering atmosphere in MSRE-type pump bowls should have negligible effects on this lubricating propetty of the pump oil. The experiment will be continued to examine long-term effects of BF, on the pump oil. 14. Development and Evaluation of Analytical Methods for Molten-Salt Reactors J. C. White The determination of oxide in highly radicactive MSRE fuel samples was continued. The replacement of the moisture-monitor cell was the first major maintenance performed since the oxide equipment was installed in the hot cell. The U*" concentrations in the fuel samples run to date by the transpiration technique do not reflect the beryllium additions which have been made to reduce the reactor fuel. This may be accounted for by an interference stemming from the radiolytic generation of fluorine in the fuel samples. This problem will receive further investigation. Ex- perimental work is also being carried out to develop a method for the remote measurement of ppm con- centrations of HF in helium or hydrogen gas streams. Design work was continued on the experimental molten-salt test loop which will be used to evaluate electrometric, spectrophotometric, and transpiration methods for the analysis of flowing molten-salt streams. Controliled-potential voltammetric and chrono- potentiometric studies were carried out on the reduction of U(IV) in molten fluoride salts using a new cyclic voltammeter. It was concluded that the U(IV) - U(III) reduction in molten LiF-BeF - ZrF, is a reversible one-electron process but that adsorption phenomena must be taken into account for voltammetric measurements at fast scan rates or for chronopotentiometric measurements at short transition times. An investigation of the spectrum of U(VI) in molten fluoride salts has been initiated. It was found that the spectrum of Na UF_ dissolved in LiF‘-BeF2 in an 3i0, cell with SiF, overpressure was identical to the spectrum of UO F , dissolved under identical conditions. It appears that the 166 equilibrium concentration of 02~ may be sufficient to react with the components of the melt, An attempt to use the SiOQ—SiF4 system in the spec- trophotometric investigation of electrochemically generated species in molten fluorides also met with difficulties. The Sil", overpressure interferes with cathodic voltammetric studies by causing very high cathodic currents. It is planned to install a spectrophotometric facility with an extended optical path adjacent to a high-radiation-level hot cell to permit the observa- tion of absorption spectra of highly radioactive materials. The basic spectrophotometer and as- sociated equipment have been ordered. The effects of BF | on MSRE pump oil have been investigated. Measurements were made of increases in hydrocarbon concentrations of an He-BF | gas stream after contact with the oil. A thermal con- ductivity detector was used to monitor the BF 3 concentration in the test gas stream. - Development studies are being made on the design of a gas chromatograph to be used for the continuous determination of sub-, low-, and high- . ppm concentrations of permanent gas impurities and water in the helium blanket gas of the MSRE. This problem of analyzing radioactive gas samples prompted the design and construction of an all- metal six-way pneumatically actuated diaphragm valve. A helium breakdown voltage detector with a glass body was designed and constructed to pernnit the observation of the helium discharge. Under optimum conditions this detector has ex- hibited a minimuin detectahle limit below 1 ppb of impurity. It appears to be possible that the detector will also operate in the less-sensitive mode nec- esgary for the determination of high-ievel con- centrations of impurities in the blanket gas. 14.1 DETERMINATION OF OXIDE IN MSRE SALTS R. F. Apple J. M. Dale A. 8. Meyer During the last week of December the moisture- monitor cell in the oxide apparaius became inop- erative. Because of other experiments being petformed in the same hot cell, the moisture- monitor cell was not replaced until March. The insensitive cell showed some supetficial evidence of radiation damage in that the potting compound (an RTV preparation which is used to seal the tube containing the spiral electrodes in a stainless steel housing) had shrunk and cracked. Flow checks revealed that substantially all the flow was still passing through the electrolysis tube, so that the damage to the potting compound could not have been responsible for the cell failure, Resistance measurements indicated that the failure was caused by either removal of or some alteration to the PO electrolyte film. The analyses of eoxide in radioactive salt samples from the MSRE for this period are summarized in Table 14.1. Two samples of radioactive fuel (IPSL-19 and IPSL-24), submitted from the In-Pile Salt Loop 2, were found to contain 265 and 240 ppm of oxide respectively. Sample IPSL-24 was stored under helium at 200°C for a period of about six months from the time of sampling uatil the analysis was made, 14.2 DETERMINATION OF U3* IN RADIOACTIVE FUEL BY A HYDROGEN REDUCTION METHOD J. M. Dale R. F. Apple A. S. Meyer A transpiration method is currently being used to determine the U3" concentration in molten radicactive MSRE fuel. The molten fuel is sparged with hydrogen to reduce oxidized species according to the reaction n - m MF 4 —5 H, —> MF_ + (0 — ;) HF , in which MF, may be UF _, NiF,, FeF,, CiFF,, or UF4 in order of their observed reduction potentials. 167 Table 14.1. Oxide Concentrations of MSRE Salt Somples g Oxide Concentration Sample Date Receive (ppm) FP-11.28 (fuel) 3-21-67 58 FP-12~4 (flush) 6-17-67 41 FP=12.18 (fuel) 7=11.67 57 The rate of production of HF is a function of the ratio of oxidized to reduced species in the melt. The theory of the method has been described pre- viously.’ The computer program which was under develop- ment has been completed and permits the calcula- tion of expected HF yields for any preselected reduciion steps on any melt composition. Using the present fuel composition and the experimental conditions of the transpiration experiment as input data to the program, HF yields were calculated for varying initial concentrations of U%*, Sample concentrations of U3 " were determined from the comparison of the experimental and calculated HF yields. Table 14.2 shows the U® " results obtained from the HF yields of the third and fourth reduction steps of the analyses and compares them with ex- pected values calculated by W. R. Grimes. The calculated results assume that 0.16% of the uranium in the fuel was originally present as U3, that the chromium concentration increase from 38 to 65 ppm which occurred before the first sample was taken resulted in the reduction of U*™ to U3 +, that each fission event results in the oxida- tion of 0.8 atom of U3+, and that there have been no other losses of U3¥, It will be noted that no analysis results are listed for samples FP-11-38 and FP-11-49. Although these samples were run in the normal manner, a total of over 2000 micromoles of HF was evolved for the four hydrogen reduction steps for each sample as compared with about 55 micromoles for the previous runs. Since this increased HF yield coincided with an increase in activity in the traps used to collect the HF, it appeared likely that the buildup in sample activity during the extended period of reactor operation might be responsible. 11, M. Dale, R. F. Apple, and A. S. Meyer, MSR Program Semiann. Progr. Repf., Feb. 28, 1967, ORNL- 4119, p. 158, 168 Table 14.2. Concentration of U3+ in MSRE Fuel Salt Ay Burnup U3+/Utota1’ U3+/U , Analysis (%) Sample No. Mwhr Oxidation Added Calculated total - {(equivalents) (equivalents) (%) Step III Step 1V FI?.9-4 10,978 1.87 0.31 0 0.1 FP-10-25 16,450 0.93 3.61 0.58 0.35 0.45 FP-11-5 17,743 £2.40 .54 0.37 0.37 FP-11-13 20,386 0.26 2.59 0.77 0.37 0.42 FP-11-32 25,510 0.88 0.69 0.33 0.34 FP-11-38 27,065 0.26 0.66 FP-11-49 30,000 0.50 1.86 0.80 FP-12-6 32,450 0.40 0.76 0.42 0.37 FP-12.11 33,095 0.10 3.95 1.14 0,38 1.2 FP«12-21 35,649 0.50 4.44 1.54 0.39 .50 If the induction period for the radiolytic generation of fluorine were shortened due to the increased activity level of the sample, the fluorine evolved during the loading of the sample could react with the inner walls of the Monel hydrogenation vessel. The copper and nickel fluorides formed would be subsequently reduced during the hydrogenation steps to produce HF. The above hypothesis appears to be supported by the results of the following experiment. One of the samples which produced the high HF yields was allowed to stand in the hydrogenator at room temperature for about a week. The sample was then subjected to additional hydrogenation steps, and HF was produced in quantities comparable with that obtained in the original runs. After standing several more days at room temperature, the sample was removed from the hydrogenator. Smaller but significant quantities of HF were obtained when the empty hydrogenator was subjected to the high-temperature hydrogenation procedure, The last three samples, FP-12-6, FP-12-11, and FPP-12-21, were all taken after a relatively brief period of reactor operation following a lengthy reactor shutdown period. None of these analyses produced the excessively high HF yields which were observed for the previous two samples. This appears to be further confirmation that the exces- sive HF yields resulted from a buildup in sample activity with extended reactor operation. Since the first addition of beryllium to the fuel, a1l the determinations of U®" not obviously affected by radioactivity have fallen in the 0.33 to 0.50% range (the one result of sample FP-12-11 could be explained by a leaky valve) and do not reflect the beryllium additions in the periods between the samplings. This could be accounted for by the evolution of fluorine in much smaller quantities than appeared to be the case in samples FP-11-38 and FP-11-39. If this is the case, the only per- manent solution would be to maintain the samples at 200°C during the time of transfer to the hot cell for analysis. However, an apparatus is now being designed which will permit the hydrogenation of synthetic fuel samples under carefully controlled conditions. It is felt that this experiment will provide a check of the validity of the transpiration method and will give further evidence as to whether or not the fluorine evolution is a real problem. Experimental work is also being carried out to develop a method for the remote measurement of ppm concentrations of HF in helium or hydrogen gas streams. The technique is primarily for appli- cation to the U?”™ transpiration experiment but, if successful, should also be applicable to the deter- mination of HF in the MSRE off-gas. The method is based on the collection of HF on a small NaF trap which is held at 70°C to prevent the adsorp- tion of water. This is followed by a desorption at a higher temperature to give a concentrated pulse of HF that can be measured by thermal conduc- tivity techniques. A components testing facility has been set up which includes a dilution system to produce HF ‘‘standards’’ as low as 20 ppm, a thermostatted trap with self-resistance heating for fast heatup, and a themmal conductivity cell with nickel fila- ments. Initial tests have revealed that the thermal conductivity cell presently being used is too sensitive to perturbations in the carrier flow and that the last traces of HF are desorbed too slowly from commercial pelletized Nal". The use of thermal conductivity cells of different geometry and better flow-control valves is expected to eliminate the {irst of these problems. The second problem will require further development studies to delermine whether the slow desorption rate repre- sents an inherent property of the NaF or is a result of impurities in the NaF. Other trapping waterials will also be investigated. A special Monel valve has been made and will be used in the remote HF measuring system. The valve incorporates two valving systems in the same metal body and will provide for the simultaneous adsorption and desorption of HF from alternate traps. Figure 14.1 shows a schematic of the flow pattern. This arrangement will eliminate any dead legs in the HF trapping system and will permit the measurement of incremental quantities of HF evolved from one hydrogenation step in the U3" transpiration experiment. One additional step is being taken with regard to the computer program which is being used for data analysis. Due to equilibrium shifts of oxidized and reduced species in the molten fuel salt with temperature changes, the starting U3 * concentration in the analysis sample at the tem- perature of the initial hydrogenation steps will necessarily be different from the U™ concentra- ORNL-OWG 67-39009 LVENT HF TRAP 1 THERMAL CONDUCTIVITY CELL TRAP 2 Fig. 14.1, System. |/ Hy PURGE Schematic Flow Diagram of HF Trapping tion in the fuel in the reactor. The computer program is presently being modified to take this into account, 14.3 IN-LINE TEST FACILITY J. M. Dale R. F. Apple A. S. Meyer Design work has been continuing with the as- sistance of J. H. Evans on the experimental molten-salt test loop which will be used to eval- uate electrometric, spectrophotometric, and transpiration methods for the analysis of flowing molten-salt streams.? The operation of flow equipment such as capillaries, corifices, and freeze valves will also be tested. A schematic tlow diagram of the proposed test loop is shown in Fig. 14.2. The first draft engineering drawings have been completed, and it is planned to start construction of the various components of the system. 14.4 ELECTROREDUCTION OF URANIUM(1Y) IN MOLTEN LiF-BeF -ZrF , AT FAST SCAN RATES AND SHORT TRANSITION TIMES D. IL.. Manning Gieb Mamantov? Controlled-potential voltammetric and chrono- potentiometric studies were carried out on the reduction of U(IV) in molten LiF-BeF ,-Zik (65.5-29.4-5.0 mole %). The controlled-potential, controlled-current cyclic voltammeter was con- structed in the Instrumentation Group of the Analytical Chemistry Division at ORNL. In the controlled-potential mode, scan rates from 0.005 to 500 v/sec are available and cell currents to 100 ma can be measured. In che controlled-current mode, currents ranging from a few microamperes to 100 ma can be passed through the cell. The built- in time base allows transition times from 400 sec to 4 msec to be measured. The instrument can also be operated in a potential-step mode for chronoamperometric experiments. Readout of the curves is accomplished with a Tektronix type 549 storage oscilloscope. 2“Ana1ytical Methods for the In-Line Analysis of Molten Fluoride Salts,” Ansal. Chem. Div, Ann. Progr, Rept. Oct. 31, 1966, ORNL-4039, p. 18. 3Consu1tant, Department of Chemistry, University of Tennessee, Knoxville, ORNL-DWG. G7-3792R ...... - i B e Q; e CONSTANT ! ' Ui =l HEAD POT (}g $—fi q,;* Jélfi : | I | | | TS| - 5 ! o @ L0 it g ! 5 i Y, | Z W l’h ! :u Z T e ! 5 2 :I | ll>:-| * -~} o I ] 5 Eu, i 131X ! T OUI i i [ @ ! o = Ll | | e [ I % a || ke | | .- : L&__.A\_ ______ J’\.-.L...,quFL..) | _;l_,r,f\, _______ MM e e — - “- z ) H,~ HF 8 r““‘T‘Al ,,,,,,,, Ao 5 a8 I HT"\----—;\G---;‘”‘-” . _fl__g;’fx wwwwww T e ANOF > O I et | < 1] 'l o Mo T b N - | r; [‘_ - He a E’ :‘g‘ || = w el I HF REMOVAL 4 z mlht O E - | || - s 5 5 &’ = Qo >| :| wn o EII II.J'\ ______ E ml Il HOLD POT g alyl - = I . o ; j Lo ~~~~T—-——Lfi1”1mmw—<- |__ fi E | E : He 1 X - t | N ' r [ — -7 _Lt"# —— I T - 7T _| 20 kg SALT i — = RESERVOIR | L - : — | ——a | | - — | L. GL . r——O_I CIRCULATING 1 PUMP ] riq | | 1 |DRIER r | L ) b e e e e J Fig. 14.2. Schematic Flow Diagram of an Experimental Moltzn-Salt Test Loop. In LiF-BeF ,ZtF , at 500°C, U(IV) is reduced to U(III) at approximately — 1.2 v vs a platinum quasi- reference electrode. For voltammetry with linearly varying potential, the peak current (ip) at 500°C is given by the Randles-Sevcik equation as i,=175x 10°a37%2AD Y 2Cvt/? where v is the rate of voltage scan (v/sec) and the symbols n, A, D, and C represent electron change, electrode area (cm?), diffusion coefficient (cm?2/ sec), and concentration of electroactive species (moles/cm?) respectively. The peak current is proportional to the concentration of uranium and also to the square root of the rate of voltage scan (v'”?) from approximately 0.02 to 1 v/sec. The diffusion coefficient at 500°C is approximately 2% 1078 cm?/sec. At faster scan rates, however, the i vs v!/? plots frequently exhibited upward curva{t)’rure, particularly at platinum and platinum- rhodium indicator electrodes. It is believed that this deviation from the Randles-Sevcik equation is caused in part by adsorption of uranium on the surface of the electrode. Conway* postulated that for a charge transfer examined by the voltage sweep method there are three contributions to the time- dependent current: (1) a non-Faradaic current C 4 @v/dt associated with charging or discharging of the ionic double layer, (2) a pseudo-Faradaic current C dv/dt associated with the change of extent of coverage by adsorbed species formed or removed in the electrochemical step, and (3) Faradaic current iF associated with any net reactions which can occur within the range of potential scans employed. Therefore the net current 1, that is recorded as a function of time during a potential sweep is itsz[dV/dt+CdV/dt-+iF ] Since dv/dt is the scan rate (v) and i , kvl/2, it can be written as 1 V1/2 where k = 1.75 x 10%23/24AD'/2C. Corrections for adsorption plus charging effects, when encountered at fast scan rates, were made by plotting ip/vl/2 vs v1/2, where the slope reflects adsorption phenomena and the intercept contains the Faradaic term. The diffusion coefficient for U(IV) calculated from the intercept was in good agreement with the value from the linear i vs v'’/ 2 plots. In chronopotentiometry, adsorption of reactant causes the io’fl/2 product, where 1 = current density (amp/cmz), to increase as 7 (transition time) decreases; this effect was observed for uranium at short transition times (<100 msec). The potential-time traces were not as well defined as the voltammograms; howevei, reasonably precise transition times could be established. Approximate corrections for adsorption effects were made utilizing a method set forth by Tatwawadi and *B. E. Conway, J. Electroanal. Chem. 8, 486 (1064), 171 Bard.® The diffusion coefficient calculated from chronopotentiometry was in agreement with the voltammeiric value. It is concluded that the U(IV) -s U(II) reduction in molten LiF-Bel ,~ZyF , is a reversible one- electron process. However, for the analyses of current-potential curves at fast rates of voltage scan or potential-time traces at short transition times, adsorption phenomena could be pronounced and must be taken into account. 14.5 SPECTROPHOTOMETRIC STUDIES OF MOLTEN FLUORIDE SALTS J. P. Young Spectrophotometric studies of interest to molten- salt reactor problems have continued. A major portion of the work for this period has been con- cerned with the use of S5i0, as a container for LiF-BeF , melts. Spectrophotometric techniques were applied to the evaluation of the compatibility problems invelved. This work is reported in a preceding section ot this report. An investigation of the spectrum of U(VL) in molten fluoride salts has been initiated. By means of spectral meas- arements it should be possible to measure the solubility of U(VI) in these solutions. The spec- tral measurements can also be nsed analytically to measure concentrations of U(VI) and to follow the change in concentration of this species during various reactions. This work is being done in cooperation with G. I. Cathers, who has furnished the solute salt, Na ,UF . The container material for molten-salt solutions of U(VI) must be inert to oxidation. It was hoped that Si0 , would satisfy these requirements. It was found, however, that the spectrum of Na UF dissolved in LiF-BeF in an Si0, cell with SiF, overpressure was identical to the spectrum of UQ ¥ dissolved under identical conditions. This would suggest that the oxide concentration in the melt was suf- ficient to cause the formation of uranyl ion. The dissolved species at 550°C exhibited an absorption peak at 419 nm and the foot of a strong absorption peak in the ultraviolet with a shoulder at 310 nm. The absorption peaks generally correspond quite well to that reported® for UQ ? * in aqueous HCIO o %S, V. Tatwawadi and A. J. Bard, Anal. Chem, 36, 2 (1964). 1. T. Bell and R. E. Biggers, J. Mol. Spectry. 22, 262 (1967); ibid., 22, in press. This uranium species disappeared slowly when iron wire was introduced into the melt, The resultant spectrum was that of tetravalent uranium; thus, it is demonstrated that an oxidized species of uranium was originally present. It would appear, then, that although SiO, containers are stable to oxidation, the equilibrium 0%~ concentration may react with components of the melt. The work on electrochemical generation of solute species and their spectrophotometric characteriza- tion is continuing.” This work is carried out in cooperation with F, L. Whiting® and Gleb Mamantov.?® Again 510, would offer several advantages in this study compared with window- less cell techniques which had been used. Under the experimental conditions presently required, however, it was found that the presence of 1 atm of SiF , gas over the melt under study interferes with cathodic voltammetric studies by causing very high cathodic currents. The reasons for this are not yet understood, but the U(IV)/U(III) reduc- tion wave is completely lost in the presence of SiF,. Conversely, SiF , does not affect anodic voltammograms. Since SiF is a product of fluoride salt reaction with SiOz, these results suggest that Si0, containers may have limited use in voltam- metric studies. The development of a recording spectrophoto- metric system is continuing that will permit spec- tral measurements of highly radioactive solutions to be obtained on samples located in a high- radiation-level cell. The system will make use of the extended optical path length similar to that being considered for use in the in-line spectral measurements of molten-salt reactors. The basic spectrophotometer and associated eguipment have been ordered, and the design of the physical and optical arrangements of the compouents for the extended path is being considered. The optical arrangement will be such that both windowed and windowless cells can be used. 14.6 ANALYSIS OF OFF-GAS FROM COMPATIBILITY TESTS OF MSRE PUMP OIL WITH BF, C. M. Boyd A total hydrocarbon detector and a thermal conductivity detector have been used to determine the hydrocarbons and BF | in the gas from tests on the effects of BF , on MSRE pump oil (Gulf- spin 35). These tests are described in an earlier section. A gas sampling manifold permits the measurement of increases in hydrocarbon concen- tration of an He-BF ; stream on contact with the A parallel stream of helium through a separate oil reservoir serves as a reference. The hydro- carbon analyzer was modified by the addition of a sampling pump which draws in the gas at near atmospheric pressure, compresses it, and then passes it through the analyzer. The portions of the test gas to be analyzed were {irst passed through a saturated solution of KF to remove the BY and thereby protect the pump and other components of the analyzer. This solution has a oil, low water vapor pressure, which prevents water condensation in the analyzer. In the first tests, with the BF , at the 2000-ppm level and the oil at 150°F, the hydrocarbon level in the off-gas was less than 50 ppm. The thermal conductivity detector was used to monitor the BF | concentration in the test gas, which was produced by mixing flows of pure BF | and He. The detector and flow capillary were calibrated by passing a known volume of the gas mixture through the detector and then through a trap of dilute NaOH. The solution was analyzed for boron and fluoride, and the BF _concentration of the gas was then calculated. The sensitivity of this detector under the conditions nsed was 10 ppm of BF .. 14.7 DEVELOPMENT OF A GAS CHROMATOGRAPH FOR THE MSRE BLANKET GAS C. M. Boyd A. S. Meyer Development studies are being made on the design of a gas chromatograph to be used for the continuous determination of permanent gas im- purities and water in the helium blanket gas of the MSRE. The chromatograph will require two columns to separate the components desired. Tests have shown that a paralle! column system composed of a 5A molecular sieves column and a Porapak S column should be practical for analyzing gases which are not highly radioactive. The H , O, 7]. P. Young, MSR Program Semiann. Progr. Rept. Feb, 28, 1967, ORNL-4119, p. 163. 8University of Tennessee, Knoxville. Nz, CH,, and CO are separated on the molecular sieves column, and the H,O and CO, are separated on the Porapak column. The chromatograph will tequire separate compartments controlled at dif- ferent constant temperatures. The sample valve, columns, and detectors may require different tem- peratures for optimum operation. The requirement that determination of impurities be made at sub-, low-, and also at higher-ppm levels necessitates the use of two different types of systems or detectors. A helium breakdown detector can be operated in either the high- and low-sensitivity modes or in the high-sensitivity mode in conjunction with a thermal conductivity detector. The chromatograph which is used for analyzing the reactor off-gas will be subjected to radiation from samples at the l-curie/cc level. This re- quires a system free of organic construction materials. An all-metal sampling valve and a detector unaffected by this radiation are necessary. The organic Porapak column cannot be used at this level of radiation. This eliminated the pos« sibility of H,0 analysis on these samples, A silica gel column will be used to resolve CO.. The problem of radioactive samples and the transmittal of gas samples containing ppm levels of H,0 from the source to the detector require the use of a heated all-metal sampling valve. Such a valve has been designed and constructed. The valve is similar to the conventional Phillips six- way pneumatically actuated diaphragm valve (Fig. 14.3). The Teflon diaphragm has been replaced by a 1-mil-thick Inconel diaphragm which contacts and seals the redesigned entry orifices (Fig. 14.4). With this metal diaphragm a spacer is required to allow gas flow without excessive backpressure. Spacers made from 2-mil gold were necessary to give a leak-tight seal between the diaphragm and the valve faces. The gold was annealed at 1500°F and, after installation in the valve, subjected to 32,000 psi pressure. Pressure maintained by the assembly screws was sufficient for retaining this seal. The choice of detectors which are sensitive to sub-ppm levels of permanent gases is limited to the helium ionization types. A helium breakdown 173 voltage detector was used on the MTR Capsule Test Facility (test 47-6)° and was not affected by the radiation present in the gas samples. This may be explained by the relatively high current levels (microamperes) used with this type of detector. A constant-curient power supply is used with this detector, and a decrease in breakdown voltage, rather than an increase in ionization cur- rent, indicates an increase in the electrical con-~ ductivity of the gas. The breakdown voltage of pure helium is about 500 v and is lowered by about 50 v by 1 ppm of impurity. The minimum detectable limit is controlled primarily by the noise level present, but under optimum conditions is below 1 ppb. In previous attempts to improve the stability of the detector discharge, various electrode radii were used with the anode and cathode in a con- centric arrangement. In this early model of the detector, which was contained inside a Swagelok tube fitting, it was impossible to observe the helium discharge. A test detector was therefore constructed which has a glass body with Kovar seal tube connections through which the electrodes are mounted (Fig. 14.5). These electrodes have removable tips which allow the testing of various electrode shapes and spacings. The effects on the helium discharge can be observed through the glass. Tests with this detector indicate that a minimum noise level is obtained with a smooth flow discharge on the anode probe. Maximum sensitivity dictates the use of a very pure helium carrier gas, but this purity level also causes a sparking or arcing in the helium discharge. The addition of mercury vapor by the presence of a small source of the metal in the tip of the anode stabilized the discharge. The more practical solution of adding a contaminant by a controllable gas flow is being tested. This approach may also permit addition of larger amounts of contaminant to allow the detector to operate in the less-sensitive mode necessary for the determination of high levels of impurities in the blanket gas samples. IMSR Program Semiann. Progr. Rept. July 31, 1964, ORNL~3708, p. 328. 174 ORNL-DWG. 6577 —————> CARRIER — > SAMPLE — - =»A TO SAMPLING LOOP B——-——> FROM SAMPLING LOOP > ACTUATING GAS Fig. 14.3. Conventional **Phillips’* Valve. . ACTUATING VENT PRESSURE CONVENTIONAL "PHILLIPS" VALVE D o 2mit GOl pC BN 22 R ! 30 ')’FLJ mil INCONEL T ACTUATING VENT FRESSURE ALL METAL "PHILLIPS' VALVE Fig. 14.4. Modifisd All-Metal Sample Valve. CRNL-LWG. 67-3043 OUTLET 3C4 55 TIPS Fig. 14.5. Diagram of He!ium Breakdown Detector. Part 4. Molten-Salt Irradiation Experiments E. G. Molten-salt breeder reactors are expected to operate with a high rate of production of fission products as a result of fuel salt power densities in excess of 200 w/cc. The effects of long-term exposure under such conditions on the stability of fuel salt, the compatibility of salt with graphite and metal (Hastelloy N), and the fate of the fis- sion products are of interest. Undue buildup of fission product poison on core graphite, for ex- ample, could reduce the breeding efficiency of the reactor if not mitigated. Initial loop experiments are being directed largely at understanding the fate of important fission products. The second thermal convection in-pile loop ex- periment was teminated by the appearance of a Bohlmann crack in the core outlet pipe, probably caused by radiation embrittlement of the alloy and stresses encountered during a reactor setback. Sufficient operating time had, however, been achieved to produce fission product concentration levels equivalent to those estimated to be present at processing equilibrium in a breeder; therefore, an exhaustive evaluation of the experiment is in progress. Future loops will be fabricated of Hastelloy N modified by titanium additions shown to inhibit the radiation effects. Experiment objectives irni- clude study of materals compatibility, fission product behavior, and effects of operation at off- design conditions. 15. Molten-Salt Convection Loop in the ORR E. L. Comperse Resulis of the first in-pile molten-salt convec- tion loop experiment in this program have been reported. Irradiation of the second molten-salt convection loop in beam hole HN-1 of the Oak Ridge Research Reactor began' January 12, 1967, and was temminated April 4, 1967, after develop- ment of 8.2 x 10!® fissions/cc (1.2% 233U burnup) in the "LiF-Be¥ -ZtF -UF, (65.3-28.2-4.8-1.7 mole %) fuel. Average fuel power densities up to 150 w per cc of salt were attained in the fuel channels of the core of MSRE-grade graphite. . c. Savage, E. L. Compere et al., MSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNI.-4110, pp. 167--73. H. C. Savage J. M. Baker The experiment was temminated alter radioactivi- ty was detected in the secondary containment sys- tems as a result of gaseous iission product leak- age from a crack in the core outlet tube. Operation and postirradiation examination of the second loop are described below. 15.1 LOOP DESCRIPTION A diagram of the second in-pile molten-salt coi- vection loop is shown in Fig. 15.1. The core section of the loop consists of a 2-in.-diam by 6-in.-long cylinder of MSRE graphite. Eight vertical 1,/4—in~—diam holes for salt flow are bored through the core in an octagonal pattemn with 177 ORNL -DWG 67-11332 Y LTYPICAL SALT LEVEL ~-CORE OUTLET PIPE / I INCONEL COOLING COILS ™ TR RETURN LINE ") (COLD LEG) }:\ SALT SAMPLE LINE Fig. 15.1. Diagram of M .'3, centers 7 in. from the graphite center line. A horizontal gas separation tank connects the top of the core through a retum line to the bottom of the core, completing the loop circuit. These and the core shell were fabricated of Hastelloy N, as was the 12-ft-long sample tube which connects the loop to the sample station in the equipment cham- ber at the ORR shield face. The Hastelloy N was from material in fabrication of the MSRE and which had not been modified to improve its resistance to irradiation embrittlement. The electric heating elements and cooling tubes surrounding the com- ponent parts of the loop were embedded in sprayed nickel. Figure 15.2 is a photograph of the partially assembled loop showing the configuration of the salt flow channels in the top of the graphite core. An identical configuration was used for the bottom of the core section. Total volume of the loop was 110 cc, with a 43-cc salt volume in the graphite. ~NICKEL SPRAY‘.._\__?::“. a4 Uy DL | - }. it ::.O:::. 0’".‘ - 4~ THERMOCOUPLE WELL 9'0‘:’0: 35‘35& ~HASTELLOY N CORE BODY (RN i;fvm'/q -in.~diom FUEL RO CHANNEL (8) 54 in. clten-5alt In-Pile Loop 2. 15.2 OPERATION After assembly, the loop was tested for leak- tightness, flushed with argon gas, and vacuum pumped at 600°C for 20 hr to remove air and moisture. The loop was flushed with solvent salt, then recharged with fresh solvent salt and operated at temperature for 248 hr in the mockup facilities in Building 9204-1, Y-12. During the out-of-pile test period, 15 salt sam- ples were removed from the loop and 12 salt addi- tions were made, providing a good demonstration of the salt sample and addition system. The loop package containing the solvent salt used in the preirradiation test period was in- stalled in beam hole HN-1 of the ORR, and in- pile operation was started on January 12, 1967. Samples of the solvent salt ("LiF-BeF ,~ZIF 4, 64.8-30.1-5.1 mole %) were taken on January 13 before the start of irradiation and again on PHOTQ 74244 # COLD LEG RETURN LINE” % HASTELLOY N BODY “ SALT CHANNELS Fig. 15.2. Photograph of Partially Assembled In-Pile Convection Loop 2 Showing Design of Salt Flow Channels in Graphite Core. January 16, 1967, after the reactor was brought to its full power of 30 Mw. Loop operation was continued with solvent salt under irradiation at temperatures ranging between 550 and 650°C. During this period the equipment and instrumenta- tion were calibrated and tested, loop performance was evaluated, and reactor gamma heat was de- termined as a function of the depth of insertion of the loop into the beam hole. lLoop operation was entirely satisfactory during this period, and on January 30, 1967, 7LiF—UF4 (63-27 mole %) eutec- tic fuel (93% 23°U) was added, along with addi- tional solvent salt, resulting in a fuel composi- tion of "LiF-BeF ,-ZrF -UF, of 65.26-28.7-4.84- 1.73 mole %. The nuclear heat generated in the loop (fission plus gamma) was again determined as a function of the depth of insertion, two fuel salt samples were removed from the loop, and on February 21, 1967, operation at the fully inserted, highest flux position was achieved. Operation in the highest flux position was continued to the end of ORR cycle No. 71 (March 5, 1967). The ORR was down from March 5 until March 11, 1967, for the regular between-cycle maintenance and refueling operations. Loop operation con- tinued throughout this period. A fuel salt sample was removed, and, by the addition of eutectic fuel and solvent salt, the uranium concentration of the fuel salt in the loop was increased to the composition of 'LiF-BeF g tF AIF , of 65.4- 27.8-4.8-2.0 mole % in order to attain the experi- mental objective of 200 w/cc fission-power densi- ties in the fuel salt. Also a sample of the cover gas in the loop was taken for analysis. The ORR was brought to full power of 30 Mw on March 11, 1967. When the molten-salt loop was placed in the fully inserted, highest flux position on March 14, 1967, it was found that the fuel fission-power density in the graphite core was 150 w/cc instead of the expected 200 w/cc. This resulted from a rearrangement of the ORR fuel between cycles 71 and 72 which caused a reduc- tion in thermal flux in beam hole HN-1 in an amount sufficient to compensate for the increased uranium in the loop fuel salt. This reduction in flux was qualitatively confirmed by other experi- menters in an adjacent beam hole facility. When the ORR was started up on March 11, 1967, it was observed that the radiation monitor on a charcoal trap in the loop container sweep-gas line read 5 mr/hr, whereas normally this monitor read zero. We concluded that this increased activity was caused by radiation from a nearby ORR primary coolant water line instead of fission product leakage from the loop into the secondary containment, and loop operation was continued. Shortly after full power operation was reached on March 14, the radiation monitor on the charcoal trap in the container sweep-gas line increased to 18 mr/hr. Some 8 hr later a further increase to ~3.4 r/hr was noted. No further increase oc- curred until March 17, when the radiation from the charcoal trap increased rapidly (over a period of ~3 hr) to ~100 r/hr, indicating a significant leakage of fission products from the loop. At this point the loop was retracted out of the high-flux region to a position at 1 to 2% of the highest flux, and the fuel salt in the loop was frozen by re- ducing the loop temperatures to ~400°C in order to prevent possible salt leakage from the loop. As a result of these actions, the charcoal trap activity decreased to ™~ 1 r/hr over a 15-hr period. From March 17 to March 23, 1967, the loop was operated in a retracted position at 1 to 2% of full power, and the fuel salt was kept frozen at a temperature of 350 to 400°C, except for brief periods of melting to determine the location of the leak. It was concluded that the fission prod- uct leak was in the vicinity of the gas separation tank and that loop operation could not be con- tinued. Three fuel salt samples were removed from the loop during this period. Beginning on March 27, 1967, the fuel salt was drained from the loop by sampling to facilitate the removal of the loop from the reactor and subse- guent examination in hot-cell facilities. By this procedure the fuel salt inventory in the loop was 179 reduced from 151.6 g to 2.1 g, requiring ten sam- ples (12 to 25 g per sample). On April 4, 1967, the ORR was shut down, and on April 5, 1967, the loop package was removed from the reactor into a shielded carrier and transferred into a hot cell without difficulty. During the in-pile operating period, fuel salt containing enriched uranium was exposed to re- actor irradiation for 1366 hr at average fission- power densities up to 150 w/cc in the graphite core fuel channels. For both out-of-pile and in- pile operating periods, various salt additions and removals were made, including samples for analy- sis. During the in-pile period, the reactor power was altered appreciably 38 times, and the distance of the loop from the reactor lattice, that is, the loop position, was changed 122 times. Thus, the equivalent time at full loop power during fueled operation was 547 hr, while the ORR was operated for 937 hr. Table 15.1 summarizes the operating periods under the various conditions for molten- salt loop No. 2. Details of some of the more significant observa- tions made during operation and results of hot- cell examinations and analyses are given in sections which follow. 15.3 OPERATING TEMPERATURES The salt in the loop was kept molten (> 490°C) during all in-pile operations until it was frozen on Table 15.1. Summary of Operating Periods for In-Pile Molten-Salt Loop 2 Operating Feriod (hr) Salt Additions and Withdrawals Full Power Total Irradiation Dose Equivalent Additions Samples Salt Removal Out-of-Pile Flush 77.8 7 1 13 Solvent salt 171.9 5 1 0 In-Pile Preirradiation 73.7 1 2 0 Solvent salt 343.8 336.5 136.0 1 2 0 Fueled salt 1101.9 G37.4 547.0 2 3 0 Retracted-fuel removal® 435.0 428.3 11.2 0 4 9 Total 2204.1 1705.2 694.2 16 13 22 “Maintained at 350 to 400°C (frozen) except during salt-removal operations and fission product leak investiga- tions. March 17 following the fission product leak. At full power, temperatures in the fuel salt ranged from ~545°C in the cold leg return line up to ~720°C in the core outlet pipe. Since nuclear heat was removed through the core graphite to the cooling coils around the outside of the Hastelloy N core body, the graphite temperature was ™~ 550°C, or some 70°C below the average fuel salt tempera- ture in the core. Temperatures of the metal walls of the component parts of the loop (Hastelloy N) ranged from a low of 510°C in the core body (where maximum cooling was used) to ~730°C in the core outlet pipe, where there were no provisions for cooling. The temperature distribution of the salt and loop components was significantly altered when the reactor was down. Under this condition, salt temperatures ranged between ~535°C in the cold- leg return line to ~670°C in the top of the core graphite fuel passages. The core outlet pipe was 670°C with no nuclear heat, as compared with ~730°C for full power operation. The core outlet pipe temperature of 730°C is believed to have contributed to the outlet pipe failure as discussed in a following section. 15.4 SALT CIRCULATION BY CONVECTION In the first in-pile molten-salt convection loop, the salt circulation rate of 5 to 10 cm?/min was substantially below the calculated rate of ~45 cm®/min at operating temperature, and frequent loss of flow occurred. In order to improve salt circulation rate and reliability in the second loop, the salt flow channels at the top and bottom of the graphite core were redesigned to increase the flow area and improve the flow path (refer to Fig. 15.2). Further, the top and bottom of the core section that were horizontally oriented in the first loop were inclined 5° to minimize trap- ping of any gas released from the salt, since it is believed that formation of gas pockets often re- sulted in flow stoppage. Salt circulation in the second loop was estimated to be 30 to 40 cm3/min; this was determined by making heat-balance measurements arcund the cold-leg return line and by adding an increment of heat in a stepwise fashion to one point in the loop and recording the time required for the heated salt to traverse a known distance as monitored by thermocouples around the loop circuit. This flow 180 rate is a fivefold increase over that observed in the first loop and is attributed to the modification described. However, occasional loss of flow still occurred. One possible explanation for this is that a sufficient temperature difference was not maintained between the salt in the hot and cold legs. This is supported by the fact that flow, when lost, could be restored by adjusting the temperatures around the loop circuit. Since oc- casional flow loss did not adversely affect in-pile operation, this was not considered to be a problem of any serious consequence. 15.5 NUCLEAR HEAT, NEUTRON FLUX, . AND SALT POWER DENSITY - Nuclear heat was detemined at various loop positions by comparing electric heat input and cooling rates with the reactor down and at full power (30 Mw). Figure 15.3 shows the results of these measurements. Although nuclear heat meas- urements based on such heat balances are not precise because of variations in the temperature distribution around the loop and variations in heat loss at different loop positions (different power levels) as well as necessary estimates of the exact inlet and outlet temperatures of the several air-water coolant mixtures, they provided a good basis for determining nuclear heat and resultant fission-power density during operation. Based on these determinations, reactor gamma heat with the loop fully inserted and filled with unfueled salt was 4200 w. Fission heat was computed by sub- tracting this value from the total heat generation - obtained when fuel salt was in the loop. In the earlier part of the operation with fueled ] salt (1.73 mole % U), the loop contained 17.84 ¢ - of uranium, 93% enriched, in a salt volume of 76.2 cc. A fission heat of 8600 w in the fully inserted position was determined, leading to an estimate of the average thermal neutron flux of 1.18 x 1013 neutrons cm™ % sec™?, or an average fuel-salt power density of 113 w/cc. Assuming from a neutron transport calculation by H. F. Bauman that the core/average flux ratio was 1.33, the average power density in the core salt was 150 w/cc at full power. The power density in the forward core tubes is estimated by using Bauman’s results to be 180 w/cc. The average thermal neutron flux in the salt was also independently determined from flux monitors CRNL-DWG 67-11833 » 104 L N e & A{SSION +GAMMA HEAT— ... I,\ R L o——— AR HEAT {w) ~ NUCL & o we . 23 43 6.3 8.3 103 12.3 LOOP POSITION-DISTANCE FROM REACTOR TANK TO LOOP CORE CENTER (in.) Fig. 15.3. Mucliear Heat Generation in Molten-Salt Loop 2. recovered during postirradiation disassembly of the loop and from the fission product activity in the final salt sainples, as discussed below. Calculations involving activity of type 304 stainless steel monitor wires or of fission products in salt samples required taking into account the relative flux history for the particular isotope as described in the section on activity calculation. Thermal neutron flux levels calculated from type 304 stainless steel monitor wires attached to various regions on the outside of the loop were (in units of neutrons ecm~ % sec™ 1): core shell, front, 4.4 to 5.5 x 10!3; coie shell, bottom (rear to front), 1.7 to 3.8 x 10'3; central well in core graphite, 2.7 to 3.4 x 10'3; core shell, rear, 1.4 to 1.9 x 10'3; gas separation tank, 1.4 to 2.6 x 103, Applying a calculated attenuation factor of 0.6 and a fuel blackness factor of 0.8 to a 181 Table 15.2. Mean Nevutron Flux in Salt as Calculated from Fission Product Activity Flux Estimate Isotope (neutrons cm ™2 sec_l) Sample 23 Sample 26 x 1013 x 1013 30-y 137¢s (6.0%) 0.82 0.90 284-d 1**Ce (5.6%) 0.93 1.05 65-d 2°Zr (6.3%) 0.65 0.96 58.3-d °lv (5.8%) 1.14 0.58 50.4-d %%sr (4.79%) 0.61 0.76 32.8-d 1*!Ce (6.0%) 0.64 0.70 12.8-d 14%Ra (6.4%) 0.27 0.28 11.1-d 1*'Nd (2.6%) 0.59 0.66 weighted mean monitor flux of 2.4 x 10!3 results in an estimated mean flux available to the fuel of 1.15 x 1013, The flux was estimated on the basis of the activity of varipus fission products in final salt samples after accounting for the relative flux history of the salt. Table 15.2 shows flux esti- mates hased on the activity of a number of iso- topes that might be expected to remain in the salt. As is frequently the case, the flux estimates based on fission product activity in salt samples are somewhat lower than the values obtained in other ways. Generally, chemical separations are used prior to counting. Because of favorable half- lives, well-established constants, low neutron cross sections, good separations, and less inter- ference from other isotopes in counting, 137Cs, 144Ce, and ?°Zr are regarded as the more reliable measures of fissions (or flux). The average of the flux estimates for these three isotopes is 0.88 x 1013, The results from the different methods of esti- mating flux are compared in Table 15.3. The value of 0.88 x 1013 obtained from the fission product activity is the most direct measure of fissions in the fuel, and, therefore, it will be used in estimates of the expected activity of other fission products produced in the loop. 15.6 CORROSION Evidence of corrosion was obtained from chemi- cal analysis of salt samples withdrawn from the Table 15.3. Comparison of Yalues for Mean Neutron Flux in Salt Obtained by Various Methods Mean Neutron Flux Method (neutrons cm 2 sec"l} Power generation rate 1.18 x 10'3 Type 304 stainless steel 1.15% 1013 monitor wires Activities of 137Cs, '*4Ce, 0.88 x 103 and J S'Zr loop and from metallographic examination of samples cut from various regions of the loop. A dissolved-chromium inventory based on analy- sis of samples indicated that 13 mg of chromium was dissolved by the flush salt, an additional 35 mg was dissolved during preirradiation opera- tion, 20 mg more during solvent salt in-pile opera- tion, and 19 mg more during the ensuing fueled operation with fissioning, for a total of 87 mg overall. The flowing salt contacted about 110 cm 2 of loop surface. A uniform 0.5-mil thickness of metal (7% Cr) from over this area would contain about 87 mg of chromium. The largest increases in dissolved chromium came during the first part of the run; this is con- sistent with out-of-pile behavior reported by DeVan and Evans.? There was no indication of any ag- gravation of corrosion by irradiation. The regions of the loop contacted by flowing salt, particularly the core outlet and cold leg, were seen on metallographic photographs to be attacked to depths of 1/10 to 1 mil. This agrees reasonably with the chemical value, which of necessity was calculated on an overall basis. The gas separation tank between the core out- let line and the cold leg showed less attack than the tubing sections, thereby indicating varying susceptibility of different items of metal. Examination of metallographic photographs of Hastelloy N from the core shell and end pieces showed darkened areas l/2 to 1 mil deep in re- gions where the metal was in contact with the core graphite, indicative of carburization there. 2}, H. DeVan and R. B. Evans ITI, ‘‘Corrosion Be- havior of Reactor Materials in Fluoride Salt Mixtures,'’ pp. 557-79 in Conference on Corrosion of Reacfor Materials, June 4—8, 1962, vol. 11, International Atomic Energy Agency, Vienna, 1962. 182 Similar darkening along the core outlet tube bottom could be either carburization or corrosion. 15.7 OXYGEN ANALYSIS Oxygen in the salt may come from moisture (or other oxygen compounds) absorbed by the salt, either from the atmosphere, the graphite, or else- where, or from the dissolution of metal oxides previously formed. Three oxygen deteminations on salt samples were made. The original solvent salt contained 115 ppm. Solvent salt withdrawn from the loop after 353 hr of in-pile circulation contained 260 ppm. The increase is equivalent to about 25 mg of oxygen (or 80 mg of chromium oxi- dized to Cr2"). Fueled salt withdrawn from the loop after retraction and freezing showed 241 ppm of oxygen. These values are well below levels expected to cause precipitation of zirconium or uranium oxides. However, some or all of the 69-mg increase in chromium content of the salt noted during the solvent-salt operation could have been due to corrosion if the oxygen increase in this period is attributed to moisture, all of which reacted to dissolve chromium from the metal. An analysis for the U3*/U*" ratio in a salt sample was attempted, but a valid determmination has not been reported. 15.8 CRACK IN THE CORE OUTLET PIPE Following its removal from beam hole HN-1, the loop was transferred to hot-cell facilities for examination. Cutup of the loop is described in a later section. After the containment vessel was opened, no evidence of salt leakage from the loop was seen by visual examination. The loop was then pressurized to ~ 100 psig with helium, and L.eak-Tec solution was applied to the extemal loop surfaces. By this technique a gas leak was observed in the core outlet pipe adjacent to its point of attachment to the core body. Subse- quently, the loop was sectioned for metallo- graphic examination, and a crack through the wall of the Hastelloy N outlet pipe (0.406 in. OD x 0.300 in. ID) was found. Figures 15.4 and 15.5 are photomicrographs of the crack, which ex- tended almost completely around the circumference of the pipe. Analysis of the cause of failure in the core out- let pipe indicates that this failure was probably e 183 i R36583 Fig. 15.4. Inner Surface and Crack in Core Outlet Pipe of In-Pile Loop 2. Top side, near core. 250x. caused by stresses resulting from differential thermal expansion of the loop components. Calculation of the piping stresses in the loop has been made for two conditions: (1) for the temperatuce profile around the loop at normal, full-power in-pile operation, and (2) for the tem- perature piofile observed during a reactor setback (change from full power to zero in ™ 11/2 min). For both conditions (1) and (2) the piping stress analysis indicates that the maximum stress from thermal expansion occurs in the core outlet pipe where the failure occurred. For the nomal operat- ing condition the bending movement produces a stress of ~ 10,000 psi in the pipe wall (tension on the top and compression on the hottom). For the temperature distribution encountered during a reactor setback, the direction of the bending move- ment is reversed, causing a stress of ~ 17,000 psi in the pipe wall (compression on top and tension on the bottom). It appears that two factors could have caused the failure in the core outlet pipe. First, the section of pipe where failure occurred was at a temperature of ~1350°F. Stress-rupture properties of Hastelloy N at 1350°F (732°C) are below those at 1200°F (650°C) used for design purposes, and these properties are further reduced®'* by the accumulated irradiation dose of ~5 x 10'? nvt. Under these conditions (1350°F and 5 x 1019 nvt), it is estimated that stresses of about 10,000 psi could produce rupture within moderate times, possibly of the order of days. Fuither, the duc- tility of Hastelloy N is reduced such that steains of 1 to 3% can result in fracture. Thus the thermal stress of ~ 10,000 psi calculated to exist in the outlet pipe at full power operation may have been sufficient to cause failure. A second and more 3H. E. McCoy, Jr., and J. R. Weir, Jr., Materials de- velopment for Molten-Salt Breeder Reactors, ORNL- TM-1854 (June 16, 1967). ‘n. E. McCoy, Jr., and J. R. Weir, Jr., In- and K x- Reactor Stress-Rupture Properties of Hastelloy N Tubing, ORNL-TM-1906 (Septembar 1967). 184 | R36587 W e = & 5 I 250x —— 3044 NCRES N ] = S o Fig. 15.5. Inner Surface and Crack in Core Qutlet Pipe of In-Pile Loop 2. Bottom side, near core. 250x, likely cause of failure is the rapid stress reversal (+10,000 to —17,500 psi) calculated for the thermal shock caused by the rapid loss of fission heat during a reactor setback. Approximately half a dozen such cycles were encountered during in-pile operation. In particular, one such cycle occurred on March 3 after a dose accumulation of ~4 x 10!? nvt, and it was on March 11 that evidence of fis- sion product leakage from the loop was first ob- served. Thus the stress reversals resulting from such cycles are very likely to have contributed to failure of the loop, since the rupture occurred at the point where the stress was a maximum, the temperature was 1350°F, and a radiation dose suf- ficient to affect the strength and ductility of Hastelloy N had been accumulated. It is evident that for future in-pile loops with similar configuration and temperatures, a material superior to the present Hastelloy N in high- temperature strength under irradiation is required. 15.9 CUTUP OF LOOP AND PREPARATION OF SAMPLES Promptness in the examination of the loop was essential. Major irradiation of the loop had ceased on March 17, 1967, when the loop was retracted following the identification of the fission gas leak. Since such short-lived isotopes as 66-hr *°Mo and 78-hr 13 2Te were of interest, it was necessary that all such samples should be counted within less than about eight weeks after this time. On April 5, 1967, following the ORR shutdown on April 4, the loop package was removed from the beam hole and transferred to the segmenting facility. The sample and addition system was stored, and inlet (‘‘cold”’) and outlet {*‘hot’’) gas tubing sections were obtained from the external equipment chamber region and submitted for radio- chemical analysis. ¥ The assembly was segmented into shield plug, connector, and loop container regions. Sections of the gas addition and gas sample tubing and sections of the salt sample line from these re- gions were obtained for radiochemical analysis. The loop container region was opened; no evi- dence of salt leakage from the loop was seen. Several sections of the gas addition and sample lines and of the salt sample line were taken. The loop was then pressurized with dry argon, and bubbles from a leak-detecting fluid indicated the crack on the top of the core outlet tubing near the core. During these operations, the core region was kept at 300°C in a fumace when not being worked on to keep radiolytic fluorine from being generated in residual salt, although the salt inventory was only about 2 g. The loop was cut into three segments — gas separation tank, cold-leg retuin line, and core — and was transferred to the High- Radiation-Level Examination L.aboratory for further cutup and examination. There the Hastel- loy N core body was removed, and the graphite was cut into upper and lower sections with thin sections removed at top, middle, and bottom for metallographic examination. Samples of loop metal were taken such that surfaces representing all regions of the loop were submitted for both radiochemical and metallo- graphic examination. In total, some 16 samples of loop metal, & of the salt sample line, 11 of the “‘hot’’ gas sample tubes, and 9 of the ‘“‘cold’’ gas addition tube were submitted for radiochemical analysis. A dozen specimens of metal from the loop, some of which contained parts of several regions of interest, have been subjected to metal- lographic examination. The results of these analyses and examinations are described in sec- tions which follow. Small amounts of blackened salt were found in the gas separation tank near the outlet, in the core bottom flow channels, and in the first few inches of salt sample line near the core. A drop- let also clung to the upper thermocouple well in the fuel channel. The total residual salt in the loop did not appear to exceed the inventory value of 2 ¢g. The penetration profiles of the various fission products in graphite were determined by collecting concentric thin shavings of core graphite from representative fuel tubes for radiochemical analy- sis. For this purpose a graduated series of 185 broaches or cylindrical shaving tools were de- signed by S. E. Dismuke. Fourteen broaches per- mitted sampling of the core graphite fuel channels (‘/4 in. ID) to a depth of 45 mils, ia steps nomi- nally ranging from 0.5 mil for the first few mils in depth up to 10 mils each for the final two cuts. In some cases, two or three cuts were collected in the same bottle in order to reduce the number of samples to be analyzed. A sample bottle was attached directly below the hole being sampled, and the broach with shaved graphite was pushed through the hole into the bottle from which it was subsequently retrieved after brushing into the bottle any adhering graphite particles. The bottle was closed, a new bottle clipped into place, and the next larger broach used. For each bottle, all the graphite sample was weighed and dissolved for radiochemical analysis. Total recovery from given holes ranged from 94 to 111% of values calculated from the broach di- ameter and graphite density. The higher values were almost entirely due to high initial cuts, indi- cating fuel channels narrower than the nominal 0.250-in.-diam, irregular original holes, or in some cases, some salt adhering to the surfaces. Since penetration depth should be measured from the original surface, actual depths were calculated from the cumulated weight of material actually removed. Forward, next-to-forward, next-to-rear, and rear fuel tubes, in top and bottom sections, were sampled in this way; a total of 76 such samples were submitted for analysis. Graphite was also shaved from the outer surface of the core cylinder in four samples to a depth of 19 mils. In addition, eight 1/8— by 3/4-in. core drillings were taken of the interior central part of the graphite. 1510 METALLGGRAPHIC EXAMINATION Samples of loop metal taken from the core out- let, gas separation tank inlet and onutlet ends, cold leg, and other regions were subjected to metallographic examination. This examination defined the nature of the break in the core outlet line and gave evidence of some carburization and corrosion of loop metal surfaces. Bottom and top inner surfaces of the core outlet tube near the core are shown in Figs. 15.4 and 15.5. These photographs give a metallographic view of the break in this tube. The break appears 186 [ T& o 0.014 INCHES lo 250x 1o Fig. 15.6. Inner Surfuce of Cold-Leg Tubing from In-Pile Loop 2 (Hastelloy N). 250x. to have been intergranular and without any indi- cation of ductility. Such behavior at temperatures in excess of 650°C and at a thermal neutron dose of 5 x 10'? neutrons/cm ? is consistent with recent ORNL studies?® of the effect of irradiation on elevated-temperature properties of Hastelloy N, as discussed in Sect. 15.8 in connection with the break in the core outlet pipe. Some evidence of attack on the inner surface of the core outlet may be noted in Figs. 15.4 and 15.5. The upper inner surface shows evidence of corrosion, even though part of the 1-mil, more finely grained layer had been removed by reaming prior to the assembly of the loop. The lower surface does not show corrosive pitting but does have a datkened bank almost 1 mil deep that could be carburization. The bottom surface (not shown) at the core end of the gas separation tank showed no evidence of either corrosion or carburzation. The cold leg of the loop showed substantial corrosive attack — probably largely intergranular in the 1-mil, more finely grained inner layer as shown in Fig. 15.6. An unexposed piece of the same tubing is shown for comparison in Fig. 15.7. This tubing was also used for fabrication of the core outlet pipe. The depth of the attack on the cold-leg pipe is of the same magnitude as would be anticipated from chtomium and oxygen analyses of the salt reported above. The tubing used to fabricate the cold leg and core outlet appears to be more sensi- tive to corrosion than materials used in other parts of the system. Carburization to a depth of about 1 mil appears to have occurred in the inner surface of the core shell wall in contact with the graphite core, as shown in Fig. 15.8. Similar carburization was also noted on core top and bottom pieces (not shown). Hardness tests were taken at various depths below the surface of the metal, the nearest about 1 mil. The test nearest the surface showed definitely greater hardness, as would be expected from carburization. - 187 i . L. . g . i o o LN A A Ty o ’@—*’fi : e - SRl % « £5F " j R—39708 o i e 260x [WO Fig. 15.7. Inner Surface of Unexposed Hastelloy N Tubing Used in Cold Leg and, After Slight Reaming, in Core Outlet of In-Pile Loop 2. 250x. 15.11 ISOTOPE ACTIVITY CALCULATION FROM FLUX AND INVENTORY HISTORY It was useful to estimate how much of a given isotope (either fission product or activation prod- uct in flux monitors, etc.) was to be expected in the system at a patticular time. This may be done by detailed application of standard equations® to the individual irradiation and inventory periods with appropriate adjustment for decay tc a standard reference time (reactor shutdown, 4-4-67, 0800). Counting data on the various samples were also referred to this time. In the in-pile period of the experiment, 38 changes in ORR power and 122 changes in experiment position altered the neutron flux, 3 salt additions altered the fuel composition, and 18 withdrawals, including 9 samples, were made. 5_]. M. West, pp. 7—-14, 15 in Nuclear FEngineering Handbook, ed. by H. Etherington, McGraw-Hill, New York, 1958. The equations described below, along with the detailed flux and inventory history, were pro- grammed for computer calculation and estimates of activity of the respective isotopes made as described below. The equations will be discussed in terms of fission products but are readily adapted to activation piroducts. The activities (A-)\] and B.)\ ) of isotopes that are the first and second significant members of a decay chain produced by a given period, ¢, of steady irradiation fol- lowed by a decay period, t, are givens by AN =Y-Fel—e e h‘["), 1 ~At] )\2)\1 1'—8 1°i __Alt B-A -Y.F. (e d) & S f > > MAQQ (1= 7 (e“"z‘d> A 188 [ro = [~ 0.014 INCHES e 250x% Fig. 15.8. Inner Surface of Core Shell in Contact with Graphite Core of In-Pile Loap 2. 250x. These equations pemnit the calculation of activity if flux is given, or of flux if activity has been measured. The ¥+ F term is the chain production rate at a given flux or power density for the quantity of material under consideration. It was useful to treat the Y. F term in the case of fissioning asa product of fission yield ¥, a standard fission rate F© per gram of uranium (at a given flux), a uranium inventory for the interval j (from salt inventory actually under irradiation, Wi times uranium concentration U,), and the irradiation intensity factors, p. and q., relative to full re- actor power and fully inserted position. Thus for any set of intervals, for example, Z(AA )] — e LU Pty —A. £, —A -t c(1—e ' e P E)) A similar expression can be written for the daughter isotope (such as the 35-day ?°Nb daugh- ter of 65-day °3Zr). The tem on the right may be computed for any given isotope from a knowledge of decay rate, irradiation, and inventory history. The term on the left represents the count to be expected, divided by the saturation value from 1 g of ura- nium at full power. In practice, the total loop inventory was cor- rected for the amount of salt not under irradiation, at the time, in sample lines and the purge tank. The activity withdrawn with each removal and that remaining in the loop were calculated. The total activity (at reference time) of a given isotope produced during the experiment was thus estimated. As mentioned in an earlier section, it appears reasonable to use the activities of certain fission products in the salt samples, in particular !*’Cs, 144Ce, and ?°Zr, as internal standards to esti- mate the flux. A mean flux to the salt of 0.88 x 10'% was thus estimated from activities measured Table 15.4. Comparison of Fission Product Activity Produced with Activity Found in Various Loop Regions All activities in units of 101° dis/min, referred to ORR shutdown, 4-4-67, 0800 147 Isotope QQMO 103Ru lOGRu 132Te 129Te QSNba gszr 131I 140Ba 8981‘ 91Y 137CS 141CE 144Ce Nd Half-life 66h 39.7d 366643 77.7h 334 35d 65 d 8.054 12.8d 50.4d 583d 30y 32.8d4 284d 11.1d Fission vield, % 6.1 3.0 0.38 4.3 0.34 6.2 6.2 2.93 6.35 1.79 5.8 6.0 6.0 5.6 2.5 Total activity formed” 940 8100 1990 1140 970 7400 12,800 5000 16,500 11,500 12,900 108 17,300 3500 61090 Found in salt samples® 2 6 d 9 d 1530 89840 1520 4800 7550 15,600 94 12,100 3830 4400 20 1290 13,3060 1570 S0O00 9370 7900 103 13,500 4330 23960 Found in residual salt 1110 180 220 902 = Found in graphite 385 d o 360 57+ 1001 126 24 410 753 d d 277 68 146 © Found on loop metal 220 d d 160 55+ 1300 77 30 220 122 d d 77 16 42 Found in salt sample line 93 o d 82 5+ 246 104 57 i50 84 d o il4 23 115 Found in hot-gas sample line 0.2 d d 0.4 <01 <1 <1 5 <1 15 d d <1 0.01 <1 Found in cold-gas inlet line <0.1 4d d <1
      0.18 Mev) and thermal- neutron doses of approximately 1 x 102! and 4 x 1020 neutrons/cm? respectively. These specimens were subjected to a temperature of 1190 +18°F for 5500 hr. The first group of vessel specimens, sk Program Semiann. Progr. Rept. Aug. 31, 1265, ORNIL.-3872, pp. 87-92. The surveillance assembly, stringer X1, was also removed. The peak thermal- neutron dose on these is approximately 2.5 x 101° neutrons/cm?. these specimens was 900 to 1200°F, and the time at temperature was approximately 11,000 hr. Initial results of examination of these groups have not re- vealed any unexpected or serious effects. These two groups have been replaced with new sets of specimens which are being exposed in the MSRE environments together with other groups that were not sampled at this time. The estimated temperature range on To review briefly, the reactor core surveillance specimens consist of both graphite and Hastelloy N mounted approximately 3 in. away from, and parallel with, the axial center line of the moderator core. The reactor vessel specimens are Hastelloy N mounted approximately 41/2 in. outside the reactor vessel, The reactor core specimens are exposed to the molten fluoride fuel, and the reactor vessel specimens are exposed to the nitrogen—2 to 5 vol % oxygen atmosphere of the reactor containment cell. The result is that the current and future irradiation 196 Table 16.1. Status of and Future Plans for the MSRE Surveillonce Program® Stringer Pulled Peak Neutron Dose Stringer Inserted Total Sampling Reacior Approximate Core Vesisel Stringer Vesself Cote Vessel No. Power Date Designation” Graphite® Hasteiloy N Designation Hastelloy N Fast® Thermal Tremal Designation Graphite Hastelloy N Designation Hastelioy N (Mwhr) - Heat No.¢ Heat No. Heat No. Heat No. 1 7,820 7-28-68 RS1, RR1, CGB 5081 None None 3.2x 107" 1.3x10%" $.5x 108 RS2 CGB 215458 None None RL1 5085 21554% RLZ CGB 5065 5085 RR2 CGB 5065 5085 2 32,450 5-5-67 RS2 CGB 21545 9.9 x 1028 4.1 x 1020 RS3 CGB 67-302% X4 67-502 21554 AXF-5QBG £7-504 67-504 X1 5065 2.7 % 101?27 x 10!° CGB-L! 5085 By 3 50,000 January 1968 RS3 CGR 67-502 7.1 x 1020 2.910%0 4.1x101° RS54 CGB Expti heat 5 AXF-5QBG 67-504 AXFS3QBG Exptl heat 6 RR2 caB 5065 1.7 x 10%1 7.0 1020 RR3 CGB Expti heat 7 5085 Expt! graphite Expti heat 8 4 70,000 July 19687 RS54 CGB Exptl heat 3 8.1x 102% 3.3 1020 5.8x10'° RS5 CGB Exptl heat 9 AXF-5QBG Exptl heat & Expt! graphite Exptl heat 10 RR3 CGBE Exptl heat 7 8.1x 1G62% 3.3 % 102¢ RR4 CGB Exptl heat 11 Exptl graphite Expti heat 8 Exptl graphite Expil heat 12 5 90,000 March 1969 RL2 cGB 5065 3.3 x 1071 14,1020 7.4x10'° S085 RS5 CGB Exptl heat 9 8.1x 1020 3.3 1020 Exptl graphite Exptl heat 10 RR4 CGB Exptl heat 11 8.1x 1020 3.3 1020 Exptl graphite Exptl heat 12 X2, X3 5065 7.4 % 10t° 5085 ¥4 57-502 4.8 % 101° 67-504 “Planned and compiled by W. H. Cock, H, £, McCoy, A. Taboada, and others, YAll these reactor core specimens have control specimens exposed 1o a static fuel szit under MSRE conditions excep! that there is no neutron radiation. “Graphite grade designations: Grade CGB is the MSRE moderator praphite which is anisotropic. Grade AXF-50QBG is an isotropic graphite. Expt! graphite refers to otber experimental grades of isotropic graphite, “Hastelloy N heat no. refers fo the heat number of 1 standard Hastelloy N composition unless noted other- wise. Lxptl heat refers to modifications of the basic lastelloy N thai are to be determined later. ¥Based on a calonlated fiux supplied by J. R, Engel for nentrons with E > 0.18 Mev. £ . sy ¢ Approximate values for both the reactor vessel wa fExperimental heat 1 which contains an addition of 0.52 wt % Ti. hExperimental heat 2 which contains an addition of 0,42 wt % 21. ‘Impregnated with graphitized pitch. JGrade AXF-30BG grughite brazed to Mo with 60 Pd—35 Ni~3 Cr (wt %) brazing alley. and the reactor vesse! specimens. *Experimental heat 3 which contains additions of (.5 Ti and 2 W wt %. ’Experimenial heat 4 which contains an addition of 0.5 wt % Hf. ™o be concurrent with salt change to one using 2330, L6l 198 (o) S FLUX MONITORS Fig. 16.1. ORNL-DWG &67-10918R STRAP Lo Cod BAND i {R52) REMOVED (RS3) INSTALLED MSRE Reactor Core Surveillance Specimens. (a) Assembled detailed plan view ond {b) Unassembled. The RI.2 and RR2 stringers were not disassembled for the removal of R52 and the installation of RS3 stringers. effects on the Hastelloy N can be monitored by the reactor vessel and reactor core specimens respec- tively. The radiation dose on the structural Hastelloy N, because of its location, is less than that on the Hastelloy N reactor core specimens (see Table 16.1). This is a desirable condition of the surveillance specimens of the MSRE, because Hastelloy N is more strongly affected by the radia- tion than is graphite under the MSRE conditions. The previous reactor core specimen assembly for the first sampling had nonnuclear, mechanical dam- age that broke some of the graphite and bent some of the Hastelloy N specimens.? This assembly was replaced by a slightly modified one, from which the current sampling was made.? The complete assem- bly is withdrawn from the reactor in order to remove any one of the three stringers. This time, the as- sembly appeared to be in the same condition as it was prior to its exposure, except that the surfaces of the Hastelloy N specimens had been dulled. One- third of the assembly, stringer RS2, was removed, and a new stringer, RS3, was joined to the other two stringers (Fig. 16.1). These were returned to ‘w. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1966, ORNL-4037, pp. 97--103, the reactor in a new containment basket, as is our standard practice. This completed our second sampling operation, shown as sampling 2 in Table 16.1. The controls for the reactor core specimens, which are exposed to fuel salt under MSRE conditions ex- cept that the salt is static and there is no radia- tion, were also removed from the controlled test rig. The typical appearance of the reactor core speci- mens and their controls is shown in Fig. 16.2. Visually, the graphite, with the exception of a few salt droplets, appeared unchanged; note that Hastelloy N tensile specimens are reflected in the bright surfaces of the graphite (Fig. 16.2¢) . The graphite specimens were pinned at their tongue-and-groove joints within Hastelloy N straps with 0.030-in.-diam Hastelloy N wire. The move- ment of the graphite column of specimens relative to that of the Hastelloy N tensile specimen rods was sufficiently difficult to deform the pins in the reactor core specimens. These pins had to be drilled out during disassembly. The pins of the control specimens were not deformed, but the con- trol assembly is less complex, since space limita- tions are not as restricted for it as for those in the reactor. Each control stringer is separate from the 199 PHOTO 89482A Fig. 16.2. MSRE Reactor Core Specimens of Grade CGB Graphite and Hastelloy N after Run 11. (a) Controls, stringer CS2, exposed to salt under MSRE conditions except the salt was static and there was no radiation. (b) and (c) Stringer RS2, exposed to the fuel in the reactor core for 27,630 Mwhr operation of the MSRE, 200 others, but all three stringers in the reactor are bound together. The new stringers for the reactor core specimens and their controls are all pinned at the joints with 0.040-in.-diam Hastelloy N in order to eliminate the pin deformation problem. Studies of the deposition of fission products on the graphite and the Hastelloy N are being con- ducted by the Reactor Chemistry Division. For their work, they used the graphite specimens from the bottom, middle, and top of the reactor core specimen stringer, plus matching controls and selected Hastelloy N samples from the stringer and basket. The results of their examinations are re- ported in Chap. 9. The remaining graphite samples are being meas- ured for dimensional changes. Tests of the types outlined in ref. 3 will be made when the dimen- sional measurements are completed. The results of the examination of the Hastelloy N tensile specimens from the core position and from outside the reactor vessel are reported in detail in Sect. 16.2. We have begun to expose samples in the MSRE core that are of interest for future molten-salt re- actors and thus have extended the scope of these studies beyond surveillance of the MSRE. In the specimens just removed, the tensile specimen rods were made of heats of Hastelloy N modified to yield increased resistance to radiation damage. One of the rods had an addition of 0.52 wt % Ti, and the other had 0.42 wt % Z:. Besides evaluating the radiation resistance of these modified alloys, we should also be able to obtain data on their corrosion resistance in the MSRE environment. The alloy modification study was continued in the new stringer, RS3, just returned to the MSRE. One of the tensile specimen rods had an addition of 0.5 wt % Ti and 2 wt % W; the other had an addition of 0.5 wt % Hf. The graphite samples included the anisotropic MSRE graphite (grade CGB), graphitized-pitch impregnated grade CGB graphite, isotropic graphite, pymolytic graphite, and a joint of isotropic graphite brazed to molybdenum with a 60 Pd—35 Ni-5 Cr (wt %) alloy. This test of the brazed joint with radiation and in flowing salt should supplement the data on corrosion of such joints in static salt with- out radiation that are discussed later in this section. Since the radiation dose received in the IMSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNL-3872, p. 89. MSRE is small compared with what will be en- countered in an MSBR, the primary purpose for in- cluding various grades of graphite is to extend the study of fission product behavior with these grades. The principal objective of this program is to en- sure the safe operation of the MSRE, We are, therefore, continuing to retain 65% or more of the space in the assembly for exposure of the MSRE grades of Hastelloy N and graphite (see Table 16.1). 16.2 MECHANICAL PROPERTIES OF THE MSRE HASTELLOY N SURVEILLANCE SPECIMENS H. E. McCoy The results of tensile tests run on the first group of specimens removed from the MSRE were reported previously.? These specimens were removed after 7823 Mwhr of reactor operation, during which they were held at 1190 + 18°F for 4800 hr and accumu- lated a thermal dose of 1.3 x 102?° neutrons/cm?. The creep-rupture tests on these specimens have now been completed, and the results are summarized in Table 16.2. The times to rupture for the surveil- lance specimens and the controls are compared in Fig. 16.3. The rupture life is reduced greatly as a result of the neutron exposure. However, Fig. 16.4 shows that the minimum creep rate is not appreciably affected. Although the reduction in the rupture life is large, it is quite comparable with what we have observed for specimens irradiated to comparable doses in the ORR in a helium environment. This point is illustrated in Fig. 16.5. The superior rup- ture life of heat 5081 is evident. The fracture strain is the parameter of greatest concern, since the rupture life does not appear to be reduced greatly by irradiation at low stress levels. The fracture strains for the various heats of irradiated material are compared in Fig. 16.6. The data indi- cate a minimum ductility for a rupture life of 1 to 10 hr, with ductility increasing with increasing rupture life. A lower ductility of heat 5085 after irradiation in the MSRE is also indicated, although the data scatter will not permit this as an un- equivocal conclusion. The superior ductility of heat 5081 and the least ductility of heat 5085 (heat used for the top and bottom heads of the MSRE) are clearly illustrated. *W. H. Cook and H. E. McCoy, MSR Program Semiann, Progr. Rept. Feb. 28, 1967, ORNL-4119, pp. 95-103. 201 Table 16.2. Postirradiation Creep-Rupture Properties of MSRE Hastelloy N Surveillance Specimens? {rradiated and Tested at 1202°F Stress Rupture Life Rupture Strain Minimum Creep Rate Test No. Heat No. (psi) (hr) () (% /1r) R-230 5085 47,000 0.8 1.45 0.81 R-266 5085 40,000 24.2 1.57 0.031 R-267 5085 32,400 148.2 1.05 0.006 R-250 5085 27,000 25.2 1.86 0.004 R-229 5081 47,000 8.7 2.28 0.20 R-231 5081 40,000 98.7 2.65 0.017 R-226 5081 32,400 474.0 3.78 0.0059 R-233 5081 27,000 2137.1 4.25 0.0009 STRESS (1000 psi) bl RUPTURE TIME (hr) ORNL-DWG &7-7938 “ Fig. 16.3. Comparative Creep-Rupture Properties of MS5RE Hastelloy N Surveillance and Control Specimens Ir- radiated and Tested at 1200°F, We also did some further work to determine the whereas the amount of precipitate in the control cause of the reduction in the low-temperature specimens was much less. Hence, the reduction in ductility reported previously., Using the extraction low-temperature ductility is probably due to the replica technique, we found that the irradiated irradiation-enhanced nucleation on the growth of specimens had extensive intergranular precipitation, grain-boundary precipitates. 202 ORNL-DWG 67-7939 ¥ T : ! [ o b = _A‘ ‘B (=N o QO o @ [6a] i L) cph n- : F . w i ] 5081 5085 | o .. CONTROL o A P ! ‘ SURVEILLANCE @ a 0 4 | C e T : ‘ 11 1 ! ‘ o | ‘ : O, ________________ ‘ JJ‘. ] i . o | AJ 10°° 107> 10 2 10" 10° 10" MINIMUM CREEP RATE {%/hr) Fig. 16.4. Comparative Creep Rates for MSRE Haostelloy N Surveillance and Control Specimens Tested at 1200°F. ORNL.-DWG 67-7940 | 50 e 1] Lo | . 8 40 - . i, . ,‘“.,i__ —_ . Or”" | T,. . ,,. 2 AVERAGE IRRADIATED DATA - @ | o 3 g 30— per ]y T 5 MSRE ORR | | o 5065 | . 20 {— & BOBT -1l 1 i . o 5085 | | ' 5084 | L ; i 0ol —. %,J,J‘ e L P . | ;\!“; T L il 0 o i 107! 10° Te} RUPTURE TIME (hr) Irradiated and Fig. 16.5. Comparative Stress-Rupture Properties at 1200°F for Various Heats of Hastelloy N Irradiated in the MSRE and the ORR at 1200°F. 203 ORML-DWG 67794 ? et TTTTYTTT ] """"" 1Tt [ | ST ’ T T Tt T g ] 6 MSRE ORR T e '. cperalb L bl cmree otk A A 5085 o ’ o 5065 || o 3 ‘ L e b L ceeden bbb o b LD L0 ¢ o 5081 | ‘ | | 2 B 5067 | J - f10] | ! : z 4 ‘l ‘ b | .| Z Q@ ~| 8 O O S8 » 5 i . ; » ' c Z (o) * 17 Tew 0.007 INCHES 500x% Jon G “ 5) Fig. 16.7. Photomicrographs of Surface of Hastelloy N Heat 5085 Exposed to the MSRE Cell Environment for 11,000 hr. (a) As polished. (b) Etched: glyceria regia. The long thermal treatment has resulted in extensive inter- granular carbide precipitation in addition to the oxidation near the surface. o 205 Ihs Q.007 INCHES 500X Ton R~39368 Fig. 16.8. Photomicrographs of the Surface of o Weld in Hastelloy N Heat 5065 After Exposure to the MSRE Cell Environment for 11,000 hr. (a) As polished. (») Etched: glyceria regia. 206 [_.. S T 0.007 INCH 500X Jen [ INCHES Q028 Fig. 16.9. Photomicrographs of the Surface of a Specimen of Titanium-Medified Hastelloy N Exposed to the MSRE Fuel Salt for 4300 hr, {a) As polished. (b) Etched: glyceria regia. 207 Tra O.007 INCHES T Fig. 16.10. Photemicrographs of the Surface of o Specimen of Zirconium-Medified Hastelloy N Exposed to the MSRE Fuel Salt for 4300 hr. (a) As polished. (b) Etched: glyceriaregio. 17. Graphite Studies 17.1 MATERIALS PROCUREMENT AND PROPERTY EVALUATION W. H. Cook The procurement and evaluation of potential grades of graphite for the MSBR remain largely limited to experimental grades of graphite. Gener- ally, most of these were not fabricated specifi- cally for MSBR requirements.! Consequently, they tend to have pore entrance diameters larger than the specified 1 ;1 and gas permeabilities greater than the desired 107 ° cm?/sec for helium. Other properties are reasonably good. The latest graph- ites, discussed below, fall into these classifica- tions. Precursory examinations of five experimental and two specialty grades of isotropic graphite have been made using 0.125-in.-diam by 1.000-in.-long specimens machined from stock and tested in a mercuty porosimeter. The pore entrance diameter distributions are summarized in Fig. 17.1. The in- sets on each plot give the grade designation, the bulk density, and accessible porosity of the speci- mens tested. The data indicate trends rather than absolute values for each grade. Grades AXF-5QBG and H-315A are the specialty grades, and the others are experimental grades. Grade AXF-5QBG is an impregnated version of grade AXF (previous designation, EP-1924), which looked promising in previous tests in that the en- trance diameters of most pores were less than 1 p. The specimens for grade AXF-5QBG, 1-1, and 16-1 (Fig. 17.1) were taken perpendicular and spec- imens 1-4 and 16-4 were taken parallel with the 4 x 6 in. plane of two different plates, each 11,/2 x 4 x 6 in. Individual results on these were plotted to show the variations of properties. The pore 'W. H. Cook, MSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4119, pp. 108—10. 208 sizes of the impregnated materials vary morée than those of the base stock but are around the 1-u size sought. It would be desirable for the pore entrance diameters to be smaller and more uniformly distrib- uted. The impregnation of the base stock reduced its accessible porosity approximately 30%. The experimental grade H17-T174 has a pore distribution similar to that of grade AXF-5QBG. Grades H-335, -336, -337, and -338 were small samples of isotropic graphite that had been impreg- nated and the impregnate graphitized. Grade H-315A was the base stock impregnated to make grade H-335. The similar pore size distributions of H-315A and H-335 illustrate again the well- known fact that graphite with pore entrance diam- eters greater than 1 j¢ is not changed much by con- ventional impregnation techniques.?'* The plot for H-315A is an average of measurements on four specimens machined from different locations of a 41Y -in.-OD x 3%-in.-ID x 11%-in. pipe section; the uniformity of the spectrum of pore entrance diameters was good. The three materials H-336, -337, and -338 appear to have been made from base stocks with pore spectra similar to that of H-315A. The accessible porosities measured for H-337 and -338 are rela- tively low for the small size of test specimens used. Some mechanical properties, specific resistivi- ties, and gas permeabilities are given in Table 17.1 for H-315A and -335 through -338. The me- chanical properties and specific resistivities of these grades are satisfactory for MSBR graphite. The gas permeabilities are high relative to those sought but seem to be typical for grades of graphite 2W. P. Eatherly et al., Proc. U.N. Intera. Cont. Peaceful Uses At. Energy, 2nd, Geneva, 1958 7, 389 401 (1958). 3w, Watt, R. L.. Bickerman, and L. W. Graham, Engi- neering 189, 110—-11 (January 1960). ) 209 AXF—5QBG 11 3 1.89g/cm 11.56 % ORNL —-DWG 6711836 AXF--5QBG 14 195 g/cm3 B8.81 % AXF -5Q86 16---4 192 g/cm3 10.83 % AXF—5QBG 16 -4 3 8 1.84 g/cm b 14.94 % o [ l.-. w =z B 1t a. HAT-Ti74 il § 3 1.80g/cm3 a 13.42 % x ooz o @ H1?;T4?4 & ~ 4B84g/cm3 — L 9.57 % = W = H-335 a 1.83 g/cm?® i2.07 % 0 TR TS — e T T T Ko H-338 1.82 g/cm® 13.24 % 4 — - e ———— e - — H~337 1,98 g/cm3 7.27 Yo O o ——— e ieamas 3 ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, s H-~338 1.93 9/cm3 8. % 0 —— e A e yoo=e=y T e ETIT et 3 .............. L D R T H-315A 1.8% g/em3 170 % () rrmmereemrm———pmnenngmeneerrreeee] )L _—i—]—'—?—“—f"r——r——-r—gl—wjlrj 1 Tt I sy A 2089 10.0 4330 2.0 40 05 04 Q.05 Q02 001 Fig. 17.1. Comparison of the Distributions of the Pore Entrance Diameters for Various Grades of Graphite. PORE ENTRANCE DIAMETER (u) 210 Table 17.1. Summary of Some of the Properties of Isntropic Graphite Bulk Specific Flexural Fracture Modulus of Permeability Graphite Orientation® Density Resistance Strength Strain Elasticity to He lium Porosity - Grade (g/cms) {microhms /em) (psi)b (% b (psi)b (cmz,/sec)c (%) % 10° x 10 * H-315A |-L 1.830(29° 985(8) 4790 p.a9) 1.35¢4 8.9° 11.705" C 8o0(®) 588047 0,49 1.65(4) H-335 ||-L 1.820(1%) 973(13) 4240 0.48(® 1.29¢8) 6.8 12.07 H-336 L 1.823(1%) gs0(! 4900®? 0.49(%) 1.47(0) 3.6 13.24 H-337 |-L 1.967¢13) 901 (1) 7210¢%7 0.70(® 1.82(%) 7.27 H-338 ||-1. 1.92001% 123513 6320 0.54(6 1.81¢9 6.91 a“l I-L” indicates that the length of the specimen “‘C’" indicates that the length of the specimen w bWork performed by C. R. Kennedy. “Work performed by R. B. Evans III. d . . . . The superscript numbers in parentheses indicate was taken parallel with the length of the stock. as taken parallel with a chord in the pipe circle. the number of values averaged. Absence of superscript numbers indicates that the data are from a single specimen. “Measured perpendicular to the length of the stock. Table 17.2. Current Fabrication of Graphite Pipe for MSBR Studies Graphite Grade Manufacturer Type @ — e Remarks oD 1D Total Length 1425-64 A Anisotrapic 3% _ 2% 306° Received 12-23-66 H-337 B Isotropic 5 2%, 36° To be shipped by 10-1-67 or sooner 1 1 b 2 {4 1 /2 768 RBRY 12 C Isotropic 5 223/32 C To be supplied during FY 1968 1 1, “Random lengths, 38 to 51 in. PRandom lengths, 10 to 36 in. “Random lengths, 10 to 48 in.; total lengths not fixed at this time, that approach MSBR requirements. Reducing the permeability below these by orders of magnitude appears to be difficult with conventional tech- niques. This has prompted the backup work on sealing graphite with metal or pyrolytically de- posited graphite that is discussed later in this section. We are receiving potential irradiation samples of anisotropic and isotropic graphites in small quanti- ties from Carbon Products Division of Union Car- bide Corporation, the Chemical Engineering Devel- opment Department of the Y-12 Plant,* Great Lakes *Operated by the Union Carbide Corporation for the U.5. Atomic Energy Commission. Carbon Corporation, Poco Graphite, Inc., Stackpole Carbon Company, and Speer Carbon Company. The first grades of graphite that have been re- ceived or will be received in larger quantities are listed in Table 17.2. Grade 1425-64 is being used in irradiation studies and graphite-to-metal joint investigations. Grades H-337 and BY12 will be used in graphite-to-metal and graphite-to-graphite joint studies and in an engineering test loop. This loop will be used to determine the operating char- acteristics of the proposed MSBR fuel cell, with special emphasis on gas permeability studies to aid in evaluation of the fission-gas behavior in the MSBR. 17.2 GRAPHITE SURFACE SEALING WiTH METALS W. C. Robinson, Jr. Chemical vapor deposition is one of the methods being investigated to decrease the gas permeability of the graphite. The two metals presently being considered for a sealant are molybdenum and ni- obium. The initial objective will be to obtain a helium permeability of 1077 em?/sec or less with a minimum thickness of metal deposit. The depo- sition parameters will be varied in order to deter- mine the conditions which produce an optimum coating. 211 The basic technique involves the deposition of metal on a heated substrate by hydrogen reduction of the metal halide. In this particular case a halide-hydrogen gas mixture is passed over graphite that is contained in a sealed furnace chamber. The metal halides being used are MoF and NbCl . The initial studies are being carried out on a nearly isotropic grade of graphite, designated R-0025, which has a helium permeability of ap- proximately 107" em?/sec and an accessible pore spectrum with maxima at 0.7 and 4 1. Molybdenum is deposited via the reactien IVIOF6 + 3H2 —> Mo + 6HF . Nine runs have been completed using the experi- mental parameters given in Table 17.3. An assembly was built for estimating the helium permeability of the coated samples. This as- sembly, which attaches to a Veeco leak detector, is shown in Fig. 17.2. Two coated and one un- coated graphite sample are shown. Qualitative measurements of the helium permeability have been performed to demonstrate the utility of the as- sembly. The leak detector will be calibrated with known leak sources to make quantitative measure- ments possible. The present qualitative evidence indicates that the helium permeability of these samples can be made less than 107 ° with 0.05 mil or less of molybdenum. Table 17.3. Molybdenum Coatings on R-0025 Graphite Gas Flow Rates Run Numbor (cms/mirl) Temlzera‘rure Pressure Tix:ne S e - o (torrs) (min) MoF 6 H2 M-Mo-1 50 800 700 5 5 M-Mo-2 50 300 700 5 10 M-Mo-3 50 800 700 10 5 M-Mo-4 50 800 700 5 5 M-Mo-5 50 800 700 10 10 M-Mo-6 50 800 800 5 5 M-Mo-7 50 800 800 5 10 M-Mo-8 50 800 800 10 5 M-Mo-9 | 50 800 800 10 10 212 PHOTO 88991 Fig. 17.2. Apparatus for Measuring the Helium Permeability of Graphite Cylinders; It Is Used with a Standard Leak Detector. 17.3 GAS IMPREGNATION OF MSBR GRAPHITES H. Beutler We are exploring the feasibility of impregnating graphites with pyrocarbon to reduce the permeation of gaseous fission products into graphite compo- nents of the MSBR core. It has been estimated that the permeability should be reduced to below 107° cm?/sec (for helium) to prevent diffusion of fission products effectively. So far, we have carried out a number of exploratory experiments which demonstrate that this objective can be at- tained with a gas-phase impregnation technique. Gas impregnation of graphite for a similar pur- pose has been studied by Watt et al. > They found that the gas permeability could be effectively re- duced by passing hydrocarbon vapors (mainly ben- zene) in a nitrogen carrier gas over graphite speci- mens at 1472 to 1652°F. The process relied entirely on diffusion to carry reactants into the pores, and there was no need for a pressure differ- ential across the specimen. With benzene as a reactant, temperatures near 1382°F were required for maximum penetration. Above 1472°F, a defi- nite concentration of deposit on or near the surface SW. Watt et al., Nucl. Power 4, 86 (1959). was found. In view of the low reaction tempera- tures, very long treatment times (up to 800 hr) were required; however, the permeability of tube speci- mens (1 in. OD, 0.5 in. ID, 1 in. long) was suc- cessfully reduced from 8.5 x 1073 to 5.4 x 10~8 cm?/sec. By machining off surface layers it was found that the thickness of the impervious layer was relatively thin (less than 100 y). The most desirable attribute of the starting material to be impregnated was found to be a uniform pore size distribution. In the course of our High-Temperature Gas- Cooled Reactor coated-particle development, we gained considerable experience with the deposition of high-density pyrocarbon outer coatings from pro- pylene over porous buffer coatings. Extensive in- filtration of pyrocarbon into the porous buffer layers was noted, even under fast-deposition rate conditions, if propylene was decomposed at low temperatures (2282°F and lower). We reasoned that the same procedure might effectively reduce the permeability of graphite. We have carried out a few exploratory experi- ments using propylene as an impregnant. The graphite specimens used so far were made from NCC graphite type R-0025, with the following di- mensions and properties: 0.400 in. OD, 0.120 in. ID, 1.5 in. length, 1.9 g/cm? density, and 1 x 101 cm?/sec. Table 17.4. Impregnation Experiment Condition and Results of Gas Specimen number GI-7 Impregnant C‘?'H6 (partial pressure, 32 torrs) Temperature of impregnation 2012°F Treatment time 51/2 hr Weight increase due to 1.2% impregnation Thickness of surface deposit 25 Helium permeability As received ~1 X 103 cm2/sec After impregnation® 1.4 x 10~7 cmz/sec After impregnation and 1.7 x 1077 cm? /sec subsequent heat freat- ment? %Determined at room temperature by helium probe gas technique after 16 hr equilibration. 213 Each specimen was immersed in fluidized coke particles so that it would be supplied uniformly with a mixture of propylene decomposition prod- ucts. The conditions and results of a typical ex- periment are given in Table 17.4. For the evaluation of impregnated specimens, we determined weight and dimensional changes and measured the helium permeability before and after subsequent heat treatments up to 5432°F in argon. We are determining the gas permeability of impreg- nated specimens using a helium gas probe tech- nique in apparatus similar to that shown in Fig. 17.2. For a cylindrical specimen, the permeability co- efficient (K) is given by: (72) F = mass flow rate, moles/sec, 21K In yo/yi where K = permeability coefficient, cm?/sec, ! = length of specimen, cm, yO y. = inside radius, cm, I = outside radius, cm, P - pressure, atm, R = gas constant, cm>-atm (°K)™! mole ™!, T = temperature, °K. The principal source of error using this technique is due to inaccuracy in calibration. More impor- tant, it suffers the disadvantage that it yields a lower value for K than the true value if measure- ments are made before steady-state conditions have been established. We have therefore equili- brated our specimens with helium for at least 16 hr prior to determining helium permeability coeffi- cients. We have, however, noted that the rubber gaskets which we employed for sealing the speci- mens are slightly permeable to helium, and our quoted permeability coefficients are probably slightly too high. The results in Table 17.4 indicate that we suc- cessfully reduced the permeability coefficient from 1x1072%to 1.4 x 1077 cm?/sec after impregnation for 51/2 hr. A subsequent heat treatment to 5432°F did not significantly increase the permeability of the specimen. We noted an increase in specimen PYROCARBON DEPOSIT 214 Fig. 17.3. Micrograph of Gas-lmpregnated Graphite Specimen, GI-5. Graphite NCC R0025. Impregnation, 2012°F /6 hr/C3H6. Bright field. 750x. diameter, which suggested a buildup of a surface layer. Metallographic examination indeed re- vealed a surface buildup of approximately 15 p. We found it difficult to resolve pyrocarbon deposits inside the graphite pores by optical microscopy; but in isolated areas (see Fig. 17.3), there was clear evidence of a buildup of pyrocarbon in in- ternal pores. We machined approximately 4 mils from the specimen surface and found that the per- meability increased (K > 10 —° cm?/sec) rapidly, which indicates that the low-permeability layer is very thin. At present we are optimizing our depo- sition conditions to increase the depth of the im- pervious layer. We are also preparing irradiation specimens for a future HFIR experiment. Although our preliminary results are encouraging, the feasibility of the tech- nique can only be assessed by fast-flux irradiation experiments. Because of the dimensional insta- bility of graphite when irradiated, we must deter- mine whether the low permeability of the graphite will be maintained or whether the pyrocarbon will change dimensions at a different rate and the per- meability increase. SiC TEMPERATURE SENSOR ALUMINUM HOUSING TUBE -~ ORNL-DWG 67—-12716 INSIDE SURFACE COATED WITH COLLOIDAL GRAPHITE TUNGSTEN HEATER iN END POSITIONS 7 / DRSS . gfl N a7 | N N e 5 . A ; s 73 £e kST.AVZUNLESS \ STEEL TUBE RADIAL SPACER “~GRAPHITE SPECIMEN SiC TEMPERATURE 0.400 in. IAM SENSOR Fig. 17.4. Schematic Drawing of the Graphite lrradiation Experiment Inserted in the HFIR. 17.4 IRRADIATION OF GRAPHITE C. R. Kennedy Experiments to irradiate graphite in target-rod positions in the core of the High Flux Isotope Re- actor have begun. The first two experimental con- tainers have been installed in HFIR for a one- cycle (three-week) exposure to determine the accuracy of heating rate and thermal analysis cal- culations. After this experiment has been removed and analyzed and the design parameters have been altered, the long-term graphite irradiations will be started. It should be possible to obtain integrated doses (E > 0.18 Mev) of 4 x 1022 neutrons/cm? in one year. The facility shown in Fig. 17.4 is designed to operate by nuclear heating at 1292 to 1328°F. A uniform axial heating rate will be maintained by adding tungsten susceptors along the axis of the samples to compensate for the axial falloff in nuclear heating. The HFIR control rod design is excellent for constancy of heating rate, and the temperature will vary only about 2% during a re- actor cycle. Irradiation temperatures will be determined using beta SiC located in a center hole in each graphite specimen. The method of temperature determina- tion using the dimensional expansion and anneal- ing characteristics of SiC is that described by Thorne et al.® This procedure has been verified by irradiation of three SiC specimens in the ORR at a controlled temperature of 1400°F. The re- sults indicate that temperatures can be determined within 9°F of the operating temperature. Past graphite irradiations to exposures greater than 1022 (refs. 7 and 8) have been limited to graphite grades that are similar, with only slight variations in the filler material or the coke used in their manufacture. When irradiated at about 1200°F all graphites seem to be characterized by an ini- tial shrinkage and then a very rapid expansion. This rapid expansion corresponds closely to that observed in the axial direction for single crystals, so it appears that the binder has deteriorated and the axial expansion of the individual crystals is controlling the growth. There are, however, indi- cations? obtained from short-term irradiations that there may be potential modifications of the coke materials that could extend the exposure required to cause the binder degradation. ®R. P. Thorne, V. C. H. Howard, and B. Hope, Radia- tion-Induced Changes in Porous Cubic Silicon Carbide, TRG 1024(c) (November 1965). 7J. W. Helm, ‘"Long Term Radiation Effects on Graphite,’’ paper MI 77, Eighth Biennial Conference on Carbon, Buffalo, N.Y., June 1567. 5R. W. Henson, A. S. Perks, and S. H. W. Simmons, ““Lattice Parameter and Dimensional Changes in Graphite Irradiated Between 300 and 1350°C,” paper MI 66, Eighth Biennial Conference on Carbon, Buffalo, N.Y., June 1967. gj. C. Bokros and R. J. Price, ‘‘Dimensional Changes Induced in Pyrolytic Carbon by High-Temperature Fast- Neutron Irradiation,’’ paper MI 68, Eighth Biennial Con- ference on Carbon, Buffalo, N.Y., June 1967. It is the purpose of our studies to develop grades of graphite which will circumvent the breakdown of the binder phase that limits the lifetime expec- tancy of the graphite core. The tailoring of a graphite for use in a molten-salt reactor will very likely require the use of impregnates and/or sur- face-sealing techniques to obtain the required pore size and permeability. Therefore, studies of the 216 effects that these treatments will have on the irra- diation behavior will be made. The studies must include examinations to demonstrate that the core lifetime is not reduced either through the loss of effectiveness of the sealing techniques or through an enhanced potential of breakdown of the binder phase. 18. Hastelloy N Studies 18.1. IMPROVING THE RESISTANCE OF HASTELLOY N TO RADIATION DAMAGE BY COMPOSITION MODIFICATIONS H. E. McCoy Although Hastelloy N has suitable properties for long-term use at elevated temperatures, we have found that the properties deteriorate when it is ex- posed to neutron irradiation. This type of radi- ation damage manifests itself through a reduction in the creep-rupture life and the rupture ductility. This damage is a function of the thermal neutron dose and is thought to be associated with the helium that is produced by the '°B(n,a) transmu- tation. However, the threshold helium countent required for damage is so low that the property detericration cannot be prevented by reducing the 108 Jevel in the alloy. We have found that slight modifications to the composition offer considerable improvement. Our studies have shown that the normal massive precipitate, identified ! as M C, can be eliminated by reducing the molybdenum level to the 12 to 13% range. 2 The strength is not reduced significantly, and the grain size is more uniform and more easily controlled. The addition of small amounts of ti- tanium, zirconium, or hafnium reduces the irradi- ation damage problem significantly. Figure 18.1 illustrates the fact that several alloys have been developed with postirradiation properties that are superior to those of unirradiated standard Hastelloy N. We are beginning work to optimize the compo- sitions and heat treatments of these alloys. We IR. E. Gehlbach and H. E. McCoy, Metals and Ce- ramics Div. Ann. Progr. Rept. June 30, 1967, ORNL.- 4170, ’H. E. McCoy and J. R. Weir, Materials Development for Molten-S5alt RBreeder Reactors, ORNIL.-TM-1854 (June 1957). 217 are also initiating the procurement of 1500-1b com- mercial melts of some of the more attractive alloys. The properties of the modified Hastelloy N in the unirradiated condition seem very attractive. Strengths are slightly better than standard Has- telloy N, and fracture ductilities are about double. 18.2. AGING STUDIES ON TiTANIUM- MODIFIED HASTELLOY N C. E. Sessions Although normal Hastelloy N is not known to suffer from detrimental aging reactions, the ad- dition of titanium to increase the resistance to radiation damage introduces another variable which might possibly affect the aging tendencies. Thus, aging studies are in progress to evaluate the creep and tensile properties of small com- mercial heats of modified Hastelloy N containing varying amounts of titanium. Initial tests on heat 66-548, which contains 0.45% Ti and 0.06% C, have been completed. Spec- imens of this alloy were fabricated by two different procedures and were then annealed at either 2150 or 2300°F. They were aged for 500 hr at tempera- tures of 1200 and 1400°F and tested at 1200°F. The results of these tests are given in Table 18.1. These results indicate that there are no del- cterious effects of aging up to 500 hr on the ten- sile properties of modified Hastelloy N containing 0.5% Ti. There ate significant changes in the strength; the changes in the yield stress do not seem to follow a pattern, but the ultimate strength is consistently increased by aging. The elonga- tion at fracture is increased by aging in most cases, remains unchanged in a few cases, but is never reduced significantly. 218 CRNL-DWG 67-3523R TG | 1 """ NN T i i ) | ‘ } Sl [ ) ‘M o ’ 60 (_‘I T | R MELTS ’ P ‘ i | J ’ | J ‘ “\ ‘| \ |‘ ! by ! V ‘\'l - B | | | 16,05 HE,L) # — i , o0 | wmesw, oz 7 [ ‘ (10,05 Ti ,C) ¢ 46,05 'I'li’ | ’ Saol T L, N S || \}\\l 2,03%0) l(7505HfL “J | e : | o | IRRADIATED AIR MELTS NG L o HH| ] ! o M /{(905TI,L)' E 30 — SR O O ‘ Py Pl | 5 ’ | ’ | ‘ ‘ (13, 1T|—L) | 3 ‘ .“ ‘ ’ l ] 20— IRRADIATION CONDITIONS t (12.5) 7 =650°C | D= 2-5%10%neutrons /ocm? ol Rl _._\___\___i__}_J..l_.. 100 1000 10,000 RUPTURE LIFE (hr) Fig. 18.1. Comparisen of the Postirradiation Creep Properties of Several Hastelloy N Alloys liradiated and Tested at 1200°F, Table 18.1. Effect of Aging at 1400 and 1200°F on Tensile Properties of Titanium-Modified Hastelloy N (Heat 66-548) Tensile Properties® Solution Annealing G_rai;’ Test Condition Yield Ultimate Total Temperature® Size Strength Strength Elongation (°F) (psi) {(psi) (%) x 103 % 103 2300 Large No age 18.5 60.5 50.0 500 hr at 1200°F 18.3 65.5 61.9 500 hr at 1400°F 23.8 68.1 50.0 Small No age 26.6 64.8 33.7 500 hr at 1200°F 22.5 79.1 63.7 500 hr at 1400°F 27.5 81.0 51.4 2150 Large No age 17.7 52.8 37.0 ' 500 hr at 1200°F 17.8 64.0 61.5 500 hr at 1400°F 19.2 62.4 51.0 Small No age 21.7 71.3 53.0 500 hr at 1200°F 21.9 77.7 60.1 500 hr at 1400°F 25.3 77.4 50.5 “Anncaled 2 hr at temperature. bLarge and small grain sizes resulted from the two different fabrication procedures. “Tests were conducted at 1200°F in air using a strain rate of 0.05 min ! These aging studies will be expanded to include longer aging times, alloys with varying titanium and carbon levels, and complete metallographic examination to ascertain metallurgical changes. 18.3. PHASE IDENTIFICATION STUDIES IN HASTELLOY N R. E. Gehlbach The effects of thermomechanical treatments on the mechanical properties of Hastelloy N indicated the need for an electron microscope investigation to understand the nature of the elevated-tempera- ture embrittlement problem. It was noticed that the transition from transgranular to intergranular failure in the temperature range of the ductility dectease indicated probable formation of a brittle grain-boundary product — either a fine precipitate or segregation of impurity elements to grain bound- aries. Preliminary examination of thinned foils by trans- mission electron microscopy showed precipitation occurring in the same temperature range as that of the pronounced ductility change. Thus a detailed study of precipitation in Hastelloy N was initiated to identify the various types, morphologies, and compositions of precipitates and to determine their relationship to the mechanical properties. Several complementary approaches are being used for the purposes of precipitate identification. Thin-foil transmission microscopy is employed for observations of fine matrix precipitate, dislocation structure, and, to a lesser extent, grain-boundary piecipitate. Due to the heterogeneous nature of precipitation in Hastelloy N, sampling problems are encountered with transmission microscopy. This technique is also limited in that the true nature of the precipitates is not apparent and that very thin grain-boundary films are frequently un- detectable. Extraction replication techniques overcome the transmission limitations. Here, conventional metallographic specimens are lightly etched to remove the polished surface and to reveal pre- cipitates and are then coated with a layer of car- bon. Extraction replicas are obtained by heavy electrolytic etching through the carbon film, which dissolves the matrix, leaving the precipitates ad- hering to the carbon substrate. This permits ob- servation of precipitates in the same distribution 219 as that existing in the bulk sample with the ex- ception that grain-boundary precipitates are no longer supported by the matrix and collapse against the substrate as shown in Fig. 18.2. This allows direct observation of the grain-boundary precip- itates on essentially the surface of the original grain boundary, providing information on true pre- cipitate morphology and permitting examination of thin films. Identification of precipitates is simplified with extraction replicas, since interference from the matrix is eliminated. Selected area electron dif- fraction is being used for the ideatification of ciystal structure and the determination of lattice parameters. Semiquantitative compositional analysis of in- dividual particles is being performed using an electron probe microanalyzer accessory on one of the electron microscopes. Thus, by coordinating various approaches available in electron micros- copy, precipitation processes may be studied carefully, Much of our effort has been concentrated on standard Hastelloy N. The microstracture of this material is characterized by stiingers of massive M.C carbides, the metallic constituents being primarily nicke! and molybdenum with some chro- mium.® The carbides are very rich in silicon com- pared with the matrix. On aging or testing in the temperature range 1112 to 1652°F, a fine grain- This precipitate is also M, C with the same lattice pa- rameter (approximately 11.0 A) as the large blocky carbides. Work is in progress to attempt to detect any compositional differences between the two morphologies of M,C carbides. Initial microprobe work using extraction replicas indicates that the grain-boundary carbides are richer in silicon than the large blocky type. All precipitates found in Hastelloy N which has not been subjected to tem- peratures in excess of 2372°F have been M.C carbides. When standard Hastelloy N is annealed at tem- peratures above about 2372°F, the M C begins to transform to an intergranular lamellar product. Autoradiography, using '*C as a tracer, shows boundary precipitate forms (Fig. 18.2). that this product is not a catbide and that the carbon seems to be rejected (Fig. 18.3). The blocky precipitates (FFig. 18.3) are seen to be 3H. E. McCoy, MSR Program Semiann. Progr. Rept. Aug. 31, 1965, ORNL-3872, pp. 94-102. 220 Y— 77654 Fig. 18.2. Extraction Replica from Hastelloy N (Heat 5065) Aged 4 hr at 1600°F. All precipitates are M6C 1000x. carbides. carbides. Figure 18.4 shows the correspondence between the morphologies of the extracted pre- cipitate and that indicated by the autoradiograph. A selected area diffraction pattern and an elec- tron probe microanalyzer trace are also included. This type of precipitate has not yet been iden- tified, but the diffraction pattern does not cot- respond to M .C. A comparison of the microprobe results with standards shows that the noncarbon constituents are about 90% Mo and 10% Cr. How- ever, some M _C carbides are also present after high-temperature anneals and exist in several morphologies. We are concluding our investigation of precip- itation in standard Hastelloy N and will be eval- uating the information generated and attempting to relate our observations to observed changes in mechanical properties. Because of the attractive in-reactor properties of the titanium-modified Hastelloy N, we are ex- panding our studies to include this material. Op- tical metallography shows the alloy to be quite free of precipitate. Initial electron microscopy studies of two experimental heats revealed that a considerable amount of very thin, fine precip- itate was present in the grain boundaries. This was rarely detected in thin-foil transmission mi- croscopy but was readily seen by the extraction replica technique. Selected area electron diffraction studies on the fine grain-boundary precipitates have revealed at least two phases, which exist in many mor- phologies. Although these precipitates have not been positively identified, indications are quite good that Ti O is one phase and that it is present in conjunction with the second type. The other phase has a face-centered cubic structure with a lattice parameter of about 4.27 A and may exist independently of the former. Considering chemical composition, TiC seems to be the most likely com- position of the second phase. However, the re- ported lattice parameter for TiC is 4.33 A, com- pared with our measured value of 4.27 A. TiN and TiB have lattice parameters more closely related to the measured value; however, chemical analyses of the heats involved show the boron and nitrogen contents to be extremely low. AUTORADICGRAPH AUTOR ey AS POLISHED f-82526 2000X 190X Fig. 18.3. Hastelloy N Containing Te Annealed 1 hr ot 2400°F., Autoradiographs made in situ on tightly etched surface using Kodak NTE liquid emulsion; 312-hr exposure. Reduced 38%. The thin film of precipitate in Fig. 18.5a is of the face-centered cubic type with ““Ti, 0" existing as the ‘‘spines’’ and polyhedral particles. The threadlike morphology of the face-centered cubic precipitate is seen in Fig. 18.55. The final objectives of this work are to identify all precipitates present and to correlate their for- mation with the mechanical behavior of the alloy. 18.4. HOT-DUCTILITY STUDIES OF ZIRCONIUM-BEARING MODIFIED HASTELLOY:N D. A. Canonico Qur previous work? has shown that zirconium additions cause weld-metal cracking in Hastelloy N. However, the zirconium addition appears to be very desirable from the standpoint of improving the resistance of the material to embrittlement by neutron irradiation. For this reason, we are continuing studies to determine how we can im- prove the weldability of zirconium-bearing alloys. One tool that we are using in our study is the Gleeble. The Gleeble permits us to assess the hot ductility of the base material as it is being subjected to a thermal cycle simulating that re- ceived by the heat-affected zone (HAZ) of a weld- ment. Specimens are fractured, on heating, at increas- ing temperatures until they no longer exhibit any *Metals and Ceramics Div. Ann, Proge, Rept. fune 30, 1967, ORNTL.-4170. . i o o T nu-?=§#$ L f#( B TR TR )l' - v e A% o . EXTRACTION REPLICA SELECTED AREA ELECTRON DIFFRACTION Y—-8254C AUTORADIO GRAPH ELECTRON MICROPROBE OQUTPUT Fig. 18.4. Hastelloy N Containing 14¢ Annealed 1 hr at 2400°F. Correlation of extraction replication, auto- radiography, selected area electron diffraction, and microprobe analysis for effective precipitate identification. degree of ductility (as measured by reduction in area). This zero ductility temperature (ZDT) is then set as the maximum temperature, and sub- sequent specimens are subjected to the ZDT, cooled at a rate which simulates that experienced by the HAZ, and fractured at a predetermined tem- perature. The ductility exhibited at these tem- peratures is compared with the on-heating duc- tility at the same temperature. The criteria for evaluating the weldability of the base metal are its ZDT and, even more important, its ability to recover its ductility after being exposed to the ZDT. The nominal analyses of the experimental alloys studied in this program are given in Table 18.2. For comparative purposes, a hot-ductility study was conducted on an MSRE grade of Hastelloy N. Its nominal composition is also given in Table 18.2. The zirconium levels in the experimental alloys ranged from 0 (No. 168) to 0.7% (No. 172). The influence of the zirconium on the ZDT is summarized in Table 18.2. It can be seen that the presence of 0.3% Zr lowered the ZDT by 225°F. This is a rather significant decrease; however, as has been pointed out earlier, the ZDT 1is not the sole criterion for evaluating the effect of zirconium. The recovery of ductility upon cooling is equally important. The on-heating and on-cooling results of the hot-ductility study are shown in Fig. 18.6. It can be seen that alloy 168, which contained no zirconium, had both a high ZDT and a satisfactory recovery of its duc- tility. Alloy 169 had a ZDT identical to 168; however, its ductility upon cooling is unsatis- factory. Alloys 170 and 171 both had ZDT’s of approximately 2150°F and both exhibited good ductilities on cooling. Figure 18.7 shows the results obtained from alloys 168, 171, and 174. The results obtained from an MSRE grade of Hastelloy N have also YE~ 9340 YE— 9344 Fig. 18.5. Hastelloy N (Titanium Modified, Heat 66- 548) Annealed 1 hr at 2150°F, Aged 30 min at About 1200°F, Typical precipitote morphologies determined by extraction replice technique. been included. It can be seen in Fig. 18.7a that the Hastelloy N modified by reducing the molyb- denum content (No. 168) is superior to the com- mercial material (MSRE grade Hastelloy N). The addition of 0.5% Zr to the modified alloy lowers the ZDT by 225°F. As is shown in Fig. 18.75, Table 18.2, The Zero Ductility Temperature of the Base Materials Studied Identification Compositioi“(ft %__)__ ZD7T Fiid::;:;n » Number Ni Mo Cr Fe Zr (°F) %) 168 bal 12 7 2350 22 169 bal 12 7 0.1 2350 21 170 bal 12 7 0.3 2125 20 171 bal 12 7 0.5 2125 32 172 bal! 12 7 0.7 2120 5 MSRE grade bal 16 4 2300 3 Hastelloy N Measured at 50°F below ZD'T. the recovery of all four of these alloys is quite good. Excellent recovery is exhibited by alloy 174. Its recovery of ductility was nearly instan- taneous and to a rather high level; however, from the standpoint of weldability, its low ZDT is not acceptable. Further efforts will be made to develop a zir- conium-bearing alloy with a higher ZDT. If we can find a combination of alloying elements that will raise the ZDT and retain the good recovery characteristics, we shall proceed further with the development of this alloy system. 18.5. RESIDUAL STRESS MEASUREMENTS IN HASTELLOY N WELDS A. G. Cepolina Welding inherently leads to large residual stresses. These stresses can lead to dimensional changes and can even be large enough to cause cracks in the weldment. For these reasons we have initiated a study to investigate the stress distribution in and near welds in Hastelloy N. A technique was developed which allowed con- tinuous reading of the stress values, thus making it possible to know, with sufficient precision, the stress gradient adjacent to the weld axis. For this study, a 12-in.-diam, l/z*irl."tllick piece of Hastelloy N base metal was used. Two circular bead-on-plate welds were simultaneously made, one on each side of the plate. The diameter of the REDUCTION IN AREA (%) ION IN AREA (%) BUcCT RE 224 ORNL—DWG 67--11837 100 | ‘ ey e - s [ | EXP HEAT 168 — 0% Zr EXP HEAT 169 — O.( % Zr | | ! | L. b . b . e e - 60 |- - 40 J\.BZ 20 ] S ‘ : L\.,__W ' | O, 100 ! ! ' ! ] l ! EXP HEAT 474 —1% Zr —0.4% Re S TN ! | | | 1400 1800 2200 2600 1400 1800 TESTING TEMPERATURE (°F) 2600 Fig. 18.6. Results of the Hot-Ductility Study Showing the Effect of Zirconium on the Modified Hastelloy N. Symbols: O.H., tested on heating; 0.C., tested on ccoling. ORNL—DWG 67--11838 ALLOY NOMINAL COMPOSITION (wi %) ‘68 B/TIL :\/120 C?r Fe zr Fig. 18.7. The Hot-Ductility Results of Selected 174 BAL 412 7 0.5 Alloys. MSRE GRADE HASTELLOY N BAL 16 7 4 100 —- — TESTED ON HEATING , circular weld beads was 6 in. (see Fig. 18.8). The 8O | welds were made with stationary inert-gas tungsten 60 arc torches while the disk rotated around its center in a vertical plane. This technique was selected 40 in order to minimize the bending effect associated e with the transverse shrinkage. For the radius and thickness ratio of our specimen, it is correct to o & assume that the stress distribution associated 100 - with the weld shrinkage is planar. The perpen- 80 | dicular shrinkage is negligible; that is, there are no stresses perpendicular to the plane of the disk. 60 — With these assumptions, the stress distribution a0 | may be determined by measuring only the tangen- tial strain on the rim of the disk while machining 20 t a series of concentric sections beginning at the Sl l . disk center. 1400 1606” 1800 200 2200 2400 We followed essentia Hy the “boring Sachs”’ TEST TEMPERATURE (°F) method which was set up for pipes. Since plane 225 ORNL—DWG 6711832 ORML-DWG 67 - 11840 o WELD BEAD(ON BOTH $IDES) ——A=in.— THICK PLATE Fig. 18.8. Sketch Showing the Location of the Weld Bead in Relation to the Overall Specimen Geometry. strain and stress problems can be studied with the same equations, the relationships for the pipe will be valid for the disk with the elimination of the terms due to the longitudinal stresses: Fig. 18.9. Pictorial Explanation of the Symhols Used in the Stress Distribution Equations. Ee [ R? O o —O — r 2 R2 ’ ORNL -~ DWG &7-11841 de A, +A | [T o, = E (Ao —A) — — ——p dA 2A ) (see Fig. 18.9), where o, = radial stress, o, = tangential stress, E - Young’s modulus for the disk material, e = total tangential strain measured on the external rim, R, = disk radius, R = internal hole radius, AO = initial disk area, A = hole area. The tangential strain is measured by strain gages that are mounted on the external rim with their axes oriented along the middle of the rim thickness (see Fig. 18.10). We used the gages recommended for stress measurement, type MM £A06 500 BH. These self-compensated strain gages have low transversal sensitivity and are Fig. 18.10. Location of the Strain Gages Used to stable over a relatively long time period. The Measure the Tangential Strain. epoxy used for bonding was EA500, and the pro- tective coatings (GageKote Nos. 2 and 5) further assure their stability. A Budd model P350 strain measuring device was used to obtain the strain values. The metal was removed by a milling cutter. This machining process eliminated the need to remove the electrical connections to the strain gages between readings. Thermal effects were avoided by submerging the specimen in a cutting fluid solution that was continuously circulated during machining. To avoid any errors due to the holders, eccentric clamping was used in order that the pressure could be easily relieved before read- ings were taken. Currently, we are analyzing the results obtained from the first welds and checking the reproduci- bility of results with identically prepared speci- meis. We intend to study the residual stress distribu- tion that results from welds made by various proc- esses and from varying parameters within a given welding process. We initiated the program with bead-on-plate welds. The investigation will be expanded to include the effect of multipass welds in a V-groove joint configuration. Fach weld pass will be deposited under identical welding param- eters, thus permitting us to study the influence of joint geometry. The metal-arc inert-gas process will also be in- vestigated in order to study the influence of this mode of filler metal addition on the residual stress distribution. The effect of postweld heat treatment on the residual stress level will also be determined. The completion of this program should allow us to define the welding process, optimum parameters within that process, and the correct postweld heat treatment that will minimize the residual stress level in Hastelloy N. 18.6. CORROSION STUDIES A. P. Litman We are continuing to study the compatibility of structural materials with fuels and coolants of interest to the Molten-Salt Reactor Program. Nat- ural-circulation loops are used as the standard test in these studies. 226 Two loops are presently in operation, Nos. 1255 and 1258. One loop, No. 10, has recently com- pleted its scheduled circulation time; one loop, No. 12, prematurely plugged recently; and four new loops, Nos. 13-16, will start operation as test salts become available. The latter loops will con- tain candidate MSBR fuel, blanket, or coolant salts. Table 18.3 details the service parameters of these test units. Fuel Salts Loop 1255, constructed of Hastelloy N and con- taining a simulated MSRE fuel salt plus 1 mole % ThF ,, continues to operate without difficulty after more than 5.4 years. Loop 1258, constructed of type 304L stainless steel and containing the same salt as loop 1255, has logged 4.1 years’ circulation time with only minor changes in flow character- istics. To examine the corrosive behavior of the relatively old simulated fuel salt in this loop, ten fresh stainless steel specimens were placed in the hot leg last January. A plot of the weight change for the new specimens at the hottest point in the system and a comparison with earlier data as a function of time are shown in Fig. 18.11. It is clear that very rapid attack occurs in the first 50 ORNL-DWG &7 ~11842 - DATA FOR SPECIMENS DURING LOOP START-UP 1964 DATA FOR NEW SPECIMENS 1967 ZONE = 4 mil/yr UNIFORM ATTACK ZONE Z 2 mil/yr UNIFORM ATTACK i WEIGHT CHANGE (mg/cm?) AN \‘4\ 2000 3000 4000 TIME {hr} 20— —————— 0 1000 Fig. 18.11. Weight Change as a Function of Time for Type 3041 Stainless Steel Specimens Exposed at 1250°F in Loop 1258 Containing LiF-BeFZ-ZrF4-UF4-ThF4 (70- 23-5-1-1 mole %). 227 Table 18.3. Thermal Convection Loop Operation Through August 31, 1967 Maximum A Time Loop AT No Loop Material Hot-Leg Specimens Test Fluid Temperature (°F) Operated LCF) (hr) 1255 Hastelloy N Hastelloy N + 2% Nb® LiF-BeF Z-ZrF 4—UF4*ThF4 1300 160 47,440 (70-23-5-1-1 mole %) 1258 Type 304L Type 304L stainless LiF-BeF?_-ZrF4-UF4—ThF4 1250 180 36,160 stainless steel steel (70-23-5-1-1 mole % 10 Hastelloy N None NaF-KF-BF | 1125 265 8,765° (48-3-49 mole % 12 Croloy 9M*? Croloy 9M¥ I‘L’:lF-KF-BI"T3 1125 260 1,440 (48-3-49 mole %) (plugged) 13 Hastelloy N Ti-modified LiF-BeF Q-UF 4 1300 300 (c) Hastelloy N (65.5-34-0.5 mole %) 14 Hastelloy N Ti-modified LiF—ThF4 (71-29 mole %) 1250 100 () Hastelloy N¢ 15 Hastelloy N Ti-modified NaF—BF3 (50-50 mole %) 1125 300 (e) Hastelloy N9 16 Hastelloy N Ti-modified NaF-BF3 (50-50 mole %) 1125 300 (e) Hastelloy N9 a : d. . Permanent specimens. Removable specimens. PScheduled loop shutdown 5-23-67. “Scheduled to start operation 9-15-67. °Scheduled to start operation 9-30-67. hr of exposure, and the rate of weight loss sub- sequently declines with time. While several per- turbations in the rates are obvious, in general, the rate loss has remained constant at between 1 PHOTO 74314A Fe PLUG and 2 mils per year equivalent uniform attack. - This is many times the corrosion rate observed for Hastelloy N under similar conditions. ~ Ainch Coolant Salts FLUCROBORATE SALT = L COLD LEG Loops 10 and 12, both of which have ceased operation, contained a fluoroborate salt which is a candidate coolant salt because of its low cost and low melting point. Loop 12, constructed of Croloy 9M, ° plugged after 1440 hr circulation due to mass-transfer and deposition of essentially pure iron crystals in the coldest portion of the loop. Similar crystals were found adhering to specimens in the hot leg. A green deposit with " DRAIN PIPE a variable composition of 15 to 17% Fe, 11 to 14% Cr, 2 to 4% B, 1.5% Mn, 10 to 15% Na, and 46% F was noted in the drain pipe (Fig. 18.12). A best Fig. 18.12. Plugging in Croloy 9M Loop 12 Containing Nc:F-KF-BF3 (48-3-49 mole %) after 1440 hr at 1125°F, SNominal analysis 9% Cr—1% Mo—Fe balance. AT = 260°F, 2 NoF-Fe F,-CrF,-BFy DEPOSIT PHOTO 74665A PIPE WALL 3NaF —Crfy PLUG Fig. 18.13. Plug Formed in Hastelloy N Loop 10 Containing N(JF-KF-BF3 (48-3-49 mole %) After 8765 hr at 1125°F. AT = 265°F. 8x. Reduced 23.50%. estimate of the stoichiometry of the deposit is 2NaF-FeF ,~CrF ,-BF ;. Metallographic examination of the hot-leg specimens disclosed only moderate surface roughening. Loop 10, fabricated from Hastelloy N, operated without incident for 8335 hr, at which time the hot-leg temperature increased about 50°F, ac- companied by a simultaneous temperature decrease of the same magnitude in the cold leg. A pertur- bation of this type is usually an indication of plugging. The temperature fluctuations ceased after 1 hr, and no further incidents occurred during the life of this loop. The loop was shut down after its 1l-year scheduled operation. Examination of the loop piping disclosed a partial plug in the lower portion of the cold leg (Fig. 18.13). The plug, which closed approximately 75% of the cross- sectional area of the pipe, was emerald green in color and analyzed to be essentially single Crys- tals of 3NaF—CrF3. 6 Analysis of the drain salt cake from loop 10, Fig. 18.14, revealed extensive increases in the concentrations of Ni, Mo, Fe, and Cr when com- pared with the before-test salt. Nickel was par- ticularly high in the bottom and top portions of 6R. E. Thoma, private communication, June 21, 1967. 229 ORNL—DWG 67-10980 Impurity Analysis: NaF-KF-BF; (48—3—49 mole %) Ni Cr Before test 87 83 After test Top slag 11.15% 1000 Center layer 90 210 Bottom layer 4,472 1500 Lwt. %. Analysis (ppm) Mo Fe 0 H-0 S 1400 400 7 146 3000 300 2,6 4850 1200 1.352 35 4200 1750 1800 5 3540 160 270 3120 1300 2, <5 7300 1500 gggg 2800 <5,19 Fig. 18.14. Drain Salt Cake from Laop 10. the cake, 4.47 and 11.15 wt % respectively. Mi- crometer measurements and metallographic exam- ination of the hot leg of the loop disclosed 1 to 2 mils of metal loss and slight surface roughening. The crossover line to the cold leg, the cold leg, and the crossover line to the hot leg all showed slight increases in wall thickness due to depo- sition of complex surface layers (Fig. 18.15). Chemical analysis of the layers disclosed that they were primarily metallic nickel (60 to 90 wt %) and molybdenum. Iron was present at locations close to the base metal, but chromium was absent in all the corrosion product layers examined. This is reasonable in view of the complex fluoride plug and the observation that in the cold leg two com- plex iron fluorides, 3NaF‘—FeF3 and NaF-FeF ,, were identified. Analysis of this loop is still proceeding, but the mode of attack appears to be dissolution of the container material in the hot leg, followed by temperature-gradient mass transfer. Chromium and iron, the latter to a lesser extent, tend to form complex fluorides, while nickel and molyb- denum plate out in metallic form. The findings to date indicate that corrosion damage was due to (a) an impure salt containing residual HF or H,O from processing or (b) the intrinsic corrosive characteristics of the fluoroborate salt (BF3 is probably the active agent). In any case, while the effect on the Hastelloy N is small in terms of metal loss from the hot section, the formation of metallic layers and complex fluorides in the colder sections is disturbing. Blockage of flow and modification of heat transfer characteristics ORNL-DWG 67-9342 32.5in Q007 inches LOCP 10 SALT: NaF-KF-BFy (48-3:49 mole %) TEMPERATURE:H125°F, A7T=285°F, TIME=8760hr LOCATION QF 3NaF-Cr F3 PLUG-... 0.007 inches Fig. 18.15. Surface Laoyers in Loop 10 after 8760 hr Operation. are, of course, the deleterious results of primary concern. Equipment Modifications During the last six months extensive modifi- cations were made to the thermal convection loop area in Building 9201-3 so as to make it more suitable for advanced studies on the compatibility of MSBR salts with Hastelloy N and modified Hastelloy N (0.5% Ti). All instrumentation has been or is in the process of being upgraded, out- moded furnaces are being replaced, and special facilities have been installed to handle toxic BF gas. Two loops ready to be charged with BF ,- bearing salts are shown in Fig. 18.16. Our future program includes operation of natural-circulation and possibly forced-circulation loops with the prime candidate fuel, blanket, and coolant salts for the MSBR (loops 13-16). A study will be made of the compatibility of graphite—Hastelloy N braze joints in fuel salt. Capsule tests to determine the effect of BF , pressure on the compatibility of the fluoroborate salts with Hastelloy N will also be conducted. 18.7. TITANIUM DIFFUSION IN HASTELLOY N C. E. Sessions T. S. Lundy Titanium additions to Hastelloy N improve the resistance of this alloy to damage by neutron ir- radiation. However, we have some concern about how the addition of titanium will influence the cotrosion resistance. Evans et al. 7 have shown that the primary corrosion mechanism of standard Hastelloy N in pure fluoride salts is the leaching of chromium by the reaction UF, +Cr— UF, + CtF, . They demonstrated that the process was con- trolled by the diffusion of chromium in the alloy. Titanium will tend to undergo a similar reaction, since the standard free energy of formation of TiF , is even more negative than that for CrF, 7R. B. Evans III, J. H. DeVan, and G. M. Watson, Self-Diffusion of Chromium in Nickel-Base Alloys, ORNL-2982 (Jan. 20, 1961). 231 PHOTO 74946 Fig. 18.16. Hastelloy N Thermal Convection Loops Ready for Charging with BF ;-Bearing Salts. ORNL-DWG 67-11844 m 0.500in. L—— *—-‘ 0.250in. Fig. 18.17. Geometry of Diffusion Specimen. (at 1110°F, AF®for TiF, is ~90 kcal per gram- atom of F and AF®for CrF, is —77 kcal per gram- atom of F).® Since protective films do not seem to form in fluoride systems, we would assume that both the chromium and the titanium will be re- moved as rapidly as these elements can diffuse. Our previous studies indicate that the corrosion rate of the standard alloy is acceptable, and the question of paramount importance is whether the addition of titanium will accelerate the corrosion tate. To answer this question we are measuring the rate of titanium diffusion in modified Hastelloy N. The following technique? is being used. Dif- fusion samples, 0.625-in.-diam cylinders (Fig. 18.17), were machined from small commercial heats of modified Hastelloy N (heat 66-548). They were heat treated 1 hr at 2400°F to establish a stable grain size and impurity distribution. The radio- active **Ti isotope was deposited on the polished face of the sample using a micropipet. The iso- tope was supplied in an HF-HC1 acid solution; therefore, additions of NH,OH were made after depositing the isotope to neutralize the solution. The sample was then heated in vacuum for 0.5 hr at 932°F to decompose the mixture to leave a thin layer of **Ti. Each sample was given a diffusion anneal for appropriate times in flowing argon at precisely controlled temperatures. Sections were then taken on a lathe at 0.001-in. increments, and the activity of the turnings was measured using a single-channel gamma spectrometer with an Nal(T1) scintillation crystal detector. At lower diffusion anneal temperatures, a hand-grinding technique 8A. Glassner, The Thermodynamic Properties of the Oxides, Fluorides, and Chlorides to 2500°K, ANL-5750. 9_]. F. Murdock, Diffusion of Titanium-44 and Vana- dium-48 in Titanium, ORNL-3616 (June 1964). 232 ORNL-DWG 67-11845 TEMPERATURE {°C) _9 125C 1200 1150 1100 1000 1C . - 5 —_ 2 <« @ < -0 NE 10 L I =z 5 wl O m L e 2 O = o —1 & 1° o L L o 5 2 ..... 10—12 65 70 75 80 85 10,000 /7 s Fig. 18.18. Diffusivity of Titanium in Modified Hastelloy N. was used to obtain smaller increments since the penetration distances were less. From a plot of the specific activity of each section against the square of the distance from the original interface, a value of titanium diffusivity was obtained for the diffusion anneal temperature of that specimen. To date we have determined the diffusivity of titanium at five temperatures from 2282 to 1922°F. The results of these measurements are shown in Fig. 18.18. Although the temperature range over which we have data is considerably higher than the proposed reactor operating temperature, we can compare the rates of titanium diffusion with that of chromium to get some ideas of the rela- tive mobilities of the two constituents. At 2012°F the diffusion rate of chromium in an Ni—20% Cr alloy is reported to be approximately 8 x 10— 1! cm?/sec.’® From our data at 2012°F the diffu- sivity of titanium in modified Hastelloy N is 3.9 x 10— ! e¢m?/sec, which is a factor of 2 lower than for chromium at this temperature. Thus on a very rough basis there does not ap- pear to be too much difference in the rate of dif- fusivity of these constituents at these higher tem- peratures. However, we cannot conclude at this 10P. L. Gruzin and G. B. Federov, Dokl. Akad. Nauk SSSR 105, 26467 (1955). time what the effective diffusivities of titanium wiil be in the alloy at 1100 to 1400°F, because at these lower temperatures short-circuit diffusion paths become an important factor in the material | transport. Since we currently have no estimate of this latter contribution to the net diffusivity for titanium, we must extend our diffusion measure- ments to lower temperatures before concluding what the expected behavior would be under reactor opetating conditions. 18.8. HASTELLOY N-TELLURIUM COMPATIBILITY C. E. Sessions The compatibility of Hastelloy N with fission products in the MSRE is of concern, since the strength and ductility of the structural material could be reduced after prolonged exposure at ele- vated temperatures. Consideration as to which of the products might be detrimental to the strength of the alloy revealed that tellurium was a poten- tially troublesome element. Tellurium is in the same periodic series as sulfur, a known detrimental element in nickel-base alloys. To evaluate the possible effects ot tellurium on Hastelloey N, several tensile samples were vapor plated with tellurium and then heat treated in quartz capsules to allow interdiffusion of the tel- lurium with the alloy. Also, several samples were 233 vapor coated with tellurium and then coated with an outer layer of pure nickel in order to reduce the vaporization of the tellurium during subsequent heat treatments. After heat treatment of the coated samples, the specimens were tensile tested at either room tem- perature or 1200°F using a strain rate of 0.05 min~!. Table 18.4 lists the test conditions and results for 12 samples of Hastelloy N. No effect of the tellurium coating on the ductility of Hastel- loy N was found at either test temperature. At 12007 F, the ductility following the various treat- ments ranged from 20 to 34% elongation, which is within the range normally obtained in the ab- sence of tellurium. At room temperature the duc- tility was in the range 52 to 57%, which again is normal. Metallographic examination of the specimens was made after testing to evaluate the interaction of tellurium with Hastelloy N. A representative area on the shoulder of one sample is shown in Fig. 18.19. Irregular surface protuberances are evident in a localized region of the edge of the Hastelloy N. The gray phase around the protu- berances is tellurium metal that remained on the sample after mechanical testing in air at room temperature. The roughness of the Hastelloy N specimens resulted from corrosive interaction of the vapor or liquid tellurium during the heat treat- ment. Al higher magnifications it is evident that a slight amount of grain-boundary penetration of the tellurium into the Hastelloy N had taken place. Table 18.4. Tensile Properties of Tetlurium—Hastelloy N Compatibility Studies® Maximum ) Annealing Test Yield Total Sample Coating Annealing Time Temperature Strength Elongation Technique Temp:rature (hr) (°F) (psi) @) (CF) Tellurium in 2012 24 75 49,100 57 capsule with 2012 24 1200 31,300 22 specimen 2012 150 75 49,800 57 2012 150 1260 37,000 20 Vapor coated with 1652 100 75 53,200 52 tellurium and 1652 100 1200 36,700 27 nickel Vapor coated with 1472 100 75 52,800 54 tellurium 14772 100 1200 40,500 34 "Tested at a strain rate of 0.05 min —1 234 Y-79588 100X T Fig. 18.19. Hastelloy N Tensile Specimen Following Exposure to Tellurium Vapor for 150 hr at 1832°F. 100x. Y-79930 . 5 & # .0 = 1 e 1 A 1 * e ‘j;_‘ % 2 * * - -3 ' o . i * . ZJ é = T Ol . “|S . o Sls O - - » o2 Choen N X - 5 7 o , * o - a. * - T = o *; o, - T o . ,__/’? - * - Fig. 18.20. Hastelloy N Sample Showing Slight Intergranulor Penetration by Metallic Tellurium. 500x. Figure 18.20 shows an area near the edge of the sample, where grain boundaries contain a tellurium- tich phase. Electron microprobe examination of this region confirmed the presence of tellurium and indicated that it was preferentially associated with the large carbide precipitates within grain bound- aries. The penetration of the tellurium into the alloy was only about 0.005 in. following the 1832°F heat treatment. No penetration was found for samples heat treated at 1472 or 1652°F. Elemental scans of the microprobe within the tellurium phase shown in Fig. 18.19 indicated 235 that manganese and chromium were also present. These elements were preferentially sepregated into bands within the tellurium matrix and un- doubtedly resulted from the leaching of these ele- ments from the Hastelloy N during the 1832°F heat treatments. The results of these tests indicate that liquid and/or vaporized tellurium metal interacts with Hastelloy N to a minor extent under the conditions tested, which included up to 150 hr in contact at 1832°F. No reduction in room-temperature or elevated-temperature ductility was found. These tests will be extended to longer exposure times. 19. Graphite-to-Metal Joining 19.1. BRAZING OF GRAPHITE TO HASTELLOY N W. J. Werner Studies were continued to develop methods for brazing large graphite pipes to Hastelloy N. At the present time, three different joint designs are being tested for circumventing the differential thermal ex- pansion problem associated with joining the two ma- terials. Concurrently, two different brazing tech- niques are under development for joining graphite to graphite and to Hastelloy N. In addition to wetta- bility and flowability of the brazing alloy on graph- ite, the brazing technique development takes under consideration the application-oriented problems of service temperature, joint strength, compatibility with the reactor environment, and braze stability under irradiation in neutron fluxes. Joint Design The first joint design, which has heen reported previously,! is based on the incorporation of a transition material between the graphite and Hastelloy N that has an expansion coefficient be- tween those of the two materials. Molybdenum and tungsten are two applicable materials, and, in ad- dition, they possess adequate compatibility with the reactor system. The transition joint design is illustrated in Fig. 19.1. The design incorporates a 10° tapered edge to reduce shear stresses arising from thermal expansion differences. The second design is the same as the first except that the transition portion of the joint is deleted and the graphite is joined directly to the Hastelloy Lysr Program Semiann. Progr. Rept. Feb, 28, 1966, ORNIL-2936, p. 140, 236 Fig. 19.1. Transition Joint Design with 10-deg Tapered Edges to Reduce Shear Stresses. From left to right the material s are graphite, molybdenum, and Hastelloy N. ORNL-DWG 67 -1184¢ e e ] 0000 CLEARANGCE HASTELLOY N~ GRAPHITE Fig. 19.2. Schematic Hlustration of a Direct Huastelioy N—to—Graphite Jaoint, N. This joint is, of course, more desirable from the standpoint of simplicity. In addition, the graph- ite is in compression, which is highly desirable for a brittle material. The amount of compressive strain induced in a 3% -in.-0D by ¥,-in.-wall graphite tube using this design is approximately 1%, which is postulated to be an acceptable compressive strain for high-density, low-permeability graphite. The third design, which is also a direct graphite— to—Hastelloy N joint, is shown schematically in Fig. 19.2. Once again we have the graphite in compression; however, in this case allowance is - made for keeping the joint in compression even if the graphite should shrink under irradiation. Brazing Development We are continuing wotk on the development of alloys suitable for joining graphite to graphite and to structural materials, We are currently looking at several alloys based on the corrosion-resistant Cu-Ni, Ni-Pd, Cu-Pd, and Ni-Nb binary systems. Quatemary compositions were prepared containing a carbide-forming elemeat plus a melting-point de- pressant. Preliminary discrimination between the various alloys was obtained through wettability tests on high-density graphite. Poor flowability was obtained with the Ni-Nb and Cu-Pd alloys. In the Pd-Ni system, the carbide-forming elements and melting-point depressant were added to the 70-30, 60-40, and 50-50 binary alloys. In the Cu-Ni sys- tem, the carbide-forming elements and melting-point depressant were added to the 80-20 and 70-30 binary alloys. Most of the alloys seemed to wet graphite well at temperatures ranging from 2102 to 2192°F. Concurrent with the brazing development work, we are investigating the radiation stability of the brazing alloys. We are cutrently irradiating four batches of Hastelloy N Miller-Peaslee braze speci- mens in the ORR. The specimens will receive a dose (thermal) of approximately 2 x 1020 neutrons/cm? at 1400°F, 19.2. COMPATIBILITY OF GRAPHITE- MOLYBDENUM BRAZED JOINTS WITH MOLTEN FLUORIDE SALTS W. H. Cook The salt-corrosion studies of joints of grade CGB graphite brazed to molyvbdenum with 60 Pd-35 Ni--5 Cr (wt %) have continued. The specimens are ex- posed to static LiF-BeF -ZrF ,-ThF -UF (70- 23.6-5-1-0.4 mole %) at 1300°F in Hastelloy N. We 237 have reported previously?'3 that there was no visible attack on the braze after a 5000-hr exposure, but there was a coating of palladium on the braze and some Cr3C2 on the graphite. A 10,000-hr test has now been concluded with similar results. All salt-corrosion tests of this series were sealed at room temperature with a pressure of approximately 4 % 107 ° torr by TIG welding. A thermal control for the 10,000-hr salt test was made in which the test components and test history were the same ex- cept that no salt was present. The results are shown in the microstiuctures of the two joints in Fig. 19.35 and ¢. The diffusion of the palladium out of the brazing alloy to form a nearly pure palladium coating on the surfaces of the braze oc- curred both in the control (the one exposed to a vacuum) and the one exposed to the salt. The coating formed in the vacuum may be mote uniform. Formation of the coating in the vacuum eliminates the salt as an agent in its formation. The more probable explanation is that the palladium is diffusing to the surface of the braze metal. The thickness of the coating appeared to be a function of time in the 5000-hr test, but this time depend- ence does not seem to continue for as long as 10,000 hr. There is some possibility that the palladium coating may help prevent corrosion of the brazing alloy by decreasing ot preventing exposure of the alloy to the salt. The chemical analyses of the salts remained essentially unchanged, as shown in Table 19.1, with the exception that the chromium content of the salt in the 10,000-hr test rose sharply relative to the others. This is higher than one would expect with Hastelloy N in these types of tests. This particular test series for this brazing alloy will be terminated by a 20,000-hr test which is in progress. Another corrosion test of this brazing alloy in a similar joint configuration is being made in the MSRE core surveillance assembly, where the joint is being exposed to radiation and flowing fuel salt. Other potential brazing alloys will be subjected to similar tests as they are developed. The most promising alloys will be more rigorously tested in dynamic salts and irradiation fields. 2W. H. Cook, MSR Program Semiann. Progr. Rept. Aug. 31, 1966, ORNL-4037, pp. 115-—-17. w. H. Cook, MSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4419, pp. 11115, 238 o Pd COATING =g PHOTO 89483 0.030in, - Fig. 19.3. Microstructure of Joints of Grade CGB Graphite Brazed to Molybdenum with 60 Pd—35 Mi—5 Cr (wt %). (a} As brazed, (b) after 10,000-hr exposure at 1300°F to vacuum, and (c) after 10,000-hr exposure at 1300°F to LiF- BeF2-ZrF4-ThF4-UF4 (70-23.6-5-1-0.4 mole %). The exposed surfaces are te the right. Etchant: Table 19.1. A Summary of the Results of the Chemical Analyses of LiF-Ber-ZrF4-ThF4-UF4 Hastelloy N to Grade CGB Grophite Joints Brozed with 60 Pd—35 Ni—-5 Cr (wit %) ? 10% oxalic acid. 100x., (70-23.6-5-1-0.4 Mole %) Salts Before and After Various Periods of Exposure in Test Period Test No. (hr) Pd Cr Fe Control 0 <0.005 0.047 0.018 A2361 100 <0.005 0.055 0.015 A2362 1000 <0.005 0.056 0.012 A2363 5000 < 0.00006 0.053 0.004 A2365 13,000 <0.0003 0.4000 0.0040 Ni Mo C 0.013 <0.001 < 0.01 0.133 0.016 <0.003 0.048 0.038 0.006 <0.003 0.049 0.026 0.003 <0.005 0.20 0.22 0.0605 <0.0003 0.0012 0.0748 ?The salt components remained constant within the tolerances of the chemical analyses. Part 6. Molten-Salt Processing and Preparation M. E. Whatley We have continued development work on the more critical parts of the fuel processing system for an MSBR. The basic concept of an on-site plant which will continuously process 14 £t of fuel salt per day through a fluorination operation to recover the ura- nium, a distillation operation to separate the carrier salt from the less-volatile fission products, and a recombination of the purified carrier salt and uranium remains unchanged. The area of work which has re- ceived most attention during this period is the dis- tillation operation. Relative volatility values have been more accurately determined and extended to additional fission product poisons. The technique of using a transpirational method for relative vol- atility measurement is being exploited. We have fabricated the experimental still which will be used 20. Vapor-Liquid Equilibrium J. R. Hightower at the MSRE to demonstrate distillation of radioactive salt, and it will be tested soon. The alternative process, removal of rare-earth fission products by reductive extraction, still appears promising; there- fore, this work is continuing. The problem of heat generation in the various components of the process- ing plant is being elucidated by a new computer code which will provide concentrations of individual fis- sion products with adequate accuracy for this purpose. We now plan to recover the uranium from the MSRE in early 1968, with a cooling time of only 35 days; for this procedure some modification of the process- ing cell will be necessary. The new loading of the MSRE in 1968 will use ?33U. Plans for the prepara- tion of this fuel are reported here. Data in Molten-Salt Mixtures L.. E. McNeese Relative volatilities of several rare-earth trifluorides having the desired composition was melted in a and alkaline-earth fluorides with respect to LiF have been measured in an equilibrium still at 1000°C.. The measured relative volatilities are in close agreement with those predicted by Raoult’s law and are low enough that a simple distillation system will work well for removing rare-earth fission products from the MSRE fuel stream. ' The equilibrium still used for these measurements is shown in Fig. 20.1. The still consists of a 1Y in.-diam still pot, a 1-in.-diam condenser, and a trap through which condensate flows prior to its return to the still pot. For each experiment, a salt charge 239 graphite crucible. The mixture was then cooled to room temperature, and the resulting salt ingot was loaded into the 1% -in.-diam section of the still. The still was welded shut, the condenser section was insulated, and the still was suspended in the furnace. Air was purged from the system by re- peatedly evacuating the system and filling to at- mospheric pressure with argon, and the pressure was set at the desired level. The furnace tempera- ture was raised to 1000°C and the condenser tem- perature set at the desired value. During runs with a given component dissolved in LiF, the operating ORNL OWG 66-8393 1-in. NICKEL PIPE | g TO VACUUM P PUMP C/> T q /p d 1'% -in. c/ NICKEL PIPE —___ - d q ¢ VAPCRIZING LIQUID CONDENSED f }zfli- SAMPLE o AIR-WATER MIXTURE “NICKEL TUBING Fig. 20.1. Molten-Salt Still Used for Relative ¥Yola- tility Measurements. Table 20.1. Relative Volatilities of Rare-Earth Trifluorides, YF3, Ban, Ser, Z.rF4, Ber at 1000°C with Respect to LiF and Relative Volatility Relative Volatility Compound in LiF—BeF2-REF in LLiF-REF Mixture Mixtures CeF, 3.3x107* 4.2 x 1074 LaF 1.4 x 1074 3x107* NdF <3 x10~" 6.3 x 1074 PrF 1.9x 1073 2.3 x 1074 SmE <3x10"* YF, 3.4 x 1077 BaF, 1.1 x 1074 SIF,, 5.0 x 107° ZrF 0.76 to 1.4 BeF, 4.73 240 pressure was 0.5 mm Hg and the condenser outlet temperature was 855 to 875°C; during runs with the LiF-Ber mixture, the pressure was 1.5 mm Hg and the condenser outlet temperature was 675 to 700°C. An experiment was continued for approximately 30 hr, after which the system was cooled to room temperature. The still was then cut open in order to remove the salt samples from the still pot and condensate trap. The samples were analyzed for all components used in the experiment. Experimentally determined relative volatilities of six rare-earth trifluorides, YF3, Ban, Ser, Ber, and ZrF with respect to LiF (measured at 1000°C and 1.5 mm Hg in a ternary system having a molar ratio of LiF to Bel', of approximately 8.5) are given in Table 20.1. The mole fraction of the component of interest varied from 0.01 to 0.05. The relative volatilities of the fluorides of the rare earths, Y, Ba, and Sr are lower than 3.3 x 107 *, with the ex- ception of Pr, which has a rclative volatility of 1.9 x 1073, The relative volatility of Z:F, varied between 0.76 and 1.4 as the ZrF, concentration was increased from 0,03 to 1.0 mole %. The average value of the relative volatility of Bel") was found to be 4.73, which indicates that vapor having the MSBR fuel carrier salt composition (66 mole % LiF~34 mole % BeF,) will be in equilibrium with liquid having the composition 91.2 mole % LiF - 9.8 mole % BeF . Relative volatilities are also given for five rare- earth trifluorides in a binary mixture of a rare-earth fluoride and LiF. These measurements were made at 1000°C and 0.5 mm Hg using mixtures having rare-earth fluoride concentrations of 2 to 5 mole %. Except for Prk (and possibly Smk',) the relative volatilities for the rare-earth fluorides are slightly lower when BeF, is present. If the molten salt solutions were ideal, the rela- tive wolatility of the respective fluorides could be calculated using Raoult’s law and the vapor pres- sure of the fluorides and L.iF. The relative vola- tility of a fluoride with respect to I.iF would be the ratio of the fluoride vapor pressure to that of LiF. In Table 20.2, experimentally determined rela- tive volatilities are compared with calculated rela- tive volatilities for several fluorides for which vapor- pressure data are available, Since the experimental errors in measuring rela- tive volatilities or vapor pressures could be as large as the discrepancies between the experimental and the calculated relative volatilities (except in the case of SrFQ) shown in Table 20.2, one can infer 241 that the behavior of the rare-earth fluorides, YFS, The relative volatilities of the fluorides of the and BaF2 in both LiF and the LiF-BeF_ mixture rare earths, yttrium, barium, and sfrontium are low studied can be approximated by assuming Raoult’s enough to allow adequate removal in a still of law. The deviation of SiF, and, in the binary sys- simple design. Zirconium removal, however, will tem, LaF, from this pattern is unexplained. be insignificant. Table 20.2. Comparison of Experimental Relative Yolatilities with Calculated Yalues Calculated Measured Component Relative Relative Volatility aexperimental Relative Volatility iexperimentai Volatility in Ternary System a’calc ulated in Binary System acafcula ted NdF 3.0x10"* <3x107* <1 6 x107* 2.0 CeF, 2.5 x 107* 3.3x10"4 1.32 4.2 x 107* 1.68 BaF 1.0 x 10~* 1.1 x107* 0.69 YF, 5.9 x 104 3.3 x107° 0.56 LaF, 4.1 %1077 1.4 x 10”4 3.4 3x107? 7.3 SrF 6.8 x 1078 5.0 x 1077 7.4 21. Relative Volatility Measurement by the Transpiration Method F. J. Smith The transpiration method for obtaining liquid- vapor equilibrium data is being used (1) to cor- roborate data obtained by the equilibrium still technique and (2) to determine relative volatilities for compounds of interest that have not yet been studied. The apparatus being used closely re- sembles that described by Cantor.! The initial experiments were made with LiF- BeF2 (86-14 mole %) over the temperature range 920 to 1055°C. Plots of the logarithms of the apparent vapor pressures of LiF and BeF, (based on the assumption that only I.iF and BeF, were present in the vapor phase) vs 1/7T were linear, although some scatter was evident in the data points for Bel',. The relative volatility of BeF, with respect to LiF was 4.0 + 0.2 over the tem- perature range investigated. This is in reason- able agreement with a value of 3.8 reported by Cantor’! for LiF-BeF, (88-12 mole %) at 1000°C. C. T. Thompson 242 L. M. Ferris Recent experiments with LiF-BeF (90-10 mole %) gave a relative volatility of 4.7 (BeF, with re- spect to LiF), which is in good agreement with the average value of 4.71 obtained by Hightower and McNeese? using an equilibrium still. Experi- ments with LiF-BeF -CeF, (85.5-9.5-5 mole %) gave vapor samples in which the concentration of CeF, was below the limit of detection by spectio- chemical methods. The relative volatility (Cel, with respect to LiF) was less than 1 x 1073, in agreement with the reported® value of 3.33 x 1074, 'Reactor Chem. Div. Ann. Progr. Rept. Dec. 31, 1966, ORNL.-4076. 2M. IZ. Whatley to D. E. Ferguson, private communica- tion, June 12, 1967. M. E. Whatley to D. E. Ferguson, private communica- tion, July 20, 1967, 22. Distillation of MSRE Fuel Carrier Salt L. E. McNeese W. L. Carter J. R. Hightower Equipment for demonstration of vacuum distil- lation using MSRE fuel salt has been built and assembled in its supporting framework. Installa- tion of the unit is in progress in Building 3541, where experiments using nonradioactive salt are to be carried out. During construction, a series of radiographic and ultrasonic examinations were made over areas of the vessels that will be sub- jected to 850 to 1000°C temperatures, and numerous dimensional measurements were taken between points on the still, condenser, and condensate receiver, Similar data will be obtained after non- radioactive operation and will be compared with initial data in evaluating vessel integrity for sub- sequent radioactive operation. The assembled unit was stress relieved by an B-hr anneal at 1600°F in an argon atmosphere in order to preclude dimensional changes during operation due to stresses arising during fabrication. Specimens of candidate materials of construc- tion for vacuum distillation equipment were in- stalled in the still so that each is exposed to molten-salt vapor and liquid in the region of high- est temperature. The test specimens are attached to the drain tube and include Hastelloy N, grade AXF-50BG isotropic graphite, Mo-TZM, Haynes { ORNL-—DWG 67-—-5B80%A 1967 1968 J A $ o N D J F M Al M J MOUNT AND CHECK INSTRUMENTATION - 3] 4 o W ) = A 2% a2 W HEATERS DELIVERED—{oR] w S S|« r b O « & w a . Z T PREPARE 3544 W = © a. > | : \\;\ NN FABRICATE STILL @ T ARSI RN N .‘2 &l = o = sl g - z _ z| G| z - o = - Lt - i < - 1 = H oo £l o “ =45 2 - S| B 1 BN ~ 9 = <3 . « N g B § 0 - Bty N 2y N oy & 3 Z 3 5 & D 3 5 2| EZ = < X g W o v L N Do d Bo|a " & o o0 o ot = o oo == W g o Q o o o Fig. 22.1. Schedule of Activities Associated w 243 ith Distillation of MSRE Fue!l Carrier Salt. alloy No. 25, and an experimental alloy (alloy No. 82) having the composition 18 Mo—0.2 Mn--0.05 Cr—bal Ni. Activities associated with the system during the past two months and scheduled activities through completion of radioactive operation are shown in Fig. 22,1, Still fabrication has been completed, and the thermal insulation and electric heaters have been received. Four faulty heaters out of a total of 41 were rejected, and replace- ments have been ordered. Instrumentation work The first of several 48-liter salt charges for nonradicactive operation is being is 90% complete, prepared by Reactor Chemistry Division personnel. Experiments using nonradioactive salt will begin 244 in early October and will continue through Feb- ruary 1968. The still will then be inspected for radioactive operation and installed at the MSRE site in the spare cell adjacent to the fuel processing cell. The equipment will then be used for distilling 48 liters of radioactive MSRE fuel salt (decayed 90 days) from which uranium has been removed by fluorination. It was determined that the roof plugs in the spare cell do not provide adequate biological shielding for 90-day-old salt; the plugs have been redesigned and will consist of 30-in,-thick barytes concrete rather than the present 24-in. ordinary concrete. 23. Steady-State Fission Product Concentrations and Heat Generation in an MSBR and Processing Plant J. S. Watson L. E. McNeese W. L. Carter Concentrations of fission products and the heat V_/V, = core volume/total fuel volume, generation associated with their decay are re- T, = processing cycle time for isotopes quired for design of systems for processing the with atomic number Z (sec), fuel and fertile streams of an MSBR. A method t = time (sec). for calculating the steady-state concentration of fission products in an MSBR will be described, The reactor was assumed to be at steady state and heat-generation rates in the fuel stream will with respect to all fission products. At steady be given. Changes in heat-generation rate re- state, sz’A/d; - 0; so the differential equation sulting from removal of fission products in both reduces to the reactor and the fuel-stream processing system will be shown. N TKPY +Nz,_.q-1 (’AO-Z,A-I(VC/Vt)JrNZ-I.A )\Z-I,A Several simplifying assumptions were made in Z.4 A, a0, SV V)T ’ calculating the concentration of fission products in the fuel stream of an MSBR. The differential which defines the steady-state concentration of equation used to define the concentration of a each isotope. This equation was solved for each given isotope was isotope using an 1BM 7090. The calculations used 2357 yields given by Blomeke and Todd;' yields v, . KPY + N 4 V. for 233U were obtained by distributing ‘‘mass 7 P z,41% 2,40 v yields” reported by England? among the isotopes f in each mass chain so that each iscotope had the + N Y N same fraction of the mass yield as reportied for z-1,4 2-1.4 Zoa 2,4 235U, Decay chains were simplified and approxi- . y s implifie PP v N mated by ‘‘straight’’ chains considering only beta - c;')c:rz'A 7‘3 NZ,A + f'A , decay. Where isomeric transitions or other t z complications were noted, personal judgment was where used to approximate the real process with straight beta decay. N = number of atoms of an isotope with Z,A atomic number Z and mass number A, P = power (w), 1]. 0. Blomeke and Mary F. Todd, Uraniumn-235 Y = primary yield of isotope of atomic Fission=-Product Production as a Function of Thermal . Neutron Flux, Irradiation Time, and Decay Time. 1. number Z and mass number A (atoms/ Atomic Concentrations and Gross Totals, ORNL-2127 fission), (part 1), vol. 1. P = tlux (neutrons sec™ ! Cm‘z), 2. R England, Time-Dependent Fission-Product o _ s : I : Themal! andResonance Absorption Cross Sections ¢ A7 cross section of isotope with aton;lc (Data Revisions and Calculational Extensions), WAPD- number Z and mass number 4 (cm*), TM-333, addendum No. 1. 245 Effective ‘‘one group’’ cross sections were cal- £ P culated from results from OPTIMERC (a reactor analysis code) for the MSBR reference design, The fuel salt was assumed to be completely mixed, and capture terms were reduced by a factor equal to the ratio of the core volume to the total fuel salt volume. Heat-generation rates, both at the instant of removal from the reactor and after various decay periods, were calculated from the steady-state fission product concentrations in the fuel using CALDRON, a fission product heat-generation and decay code, Heat-generation rates obtained with 233U vields were in good agreement with rates calculated by a modified (to treat a steady-state reactor with continuous processing) version of PHOBE, a code written by E. D. Arnold and based on experimental data for decay of fission products, Agreement with a reputable code indicates that the simplifications in the present calculations were reasonable. In the MSBR, some fission products will be re- moved from the fuel by the gas purge or by plating on solid surfaces, and other fission products will be removed in processing (fluorination) before the 246 salt reaches the still. Removal of fission products by plating and gas stripping was taken into account by assigning appropriate residence times, T, for volatile or noble elements. After withdrawal from the reactor, the fuel salt was assumed to be re- tained in a hold tank for 12 hr, after which specified fractions of elements having volatile fluorides were removed. The remaining fission products were then allowed to decay for an additional 12 hr be- fore entering the still. Figure 23.1 shows fuel salt heat-generation rates calculated for various times after removal These calculations were made 2220 Mw (thermal), 2 from the reactor. with MSBR design conditions: 3.7 % 10** neutrons sec™! cm™?, core volume of 9400 liters, fuel salt volume of 25,400 liters, and 4.5 % 10° sec (52 days) processing cycle. The uppermost curve represents no fission product re- moval in the reactor or in the fluorinator. The next lower solid curve represents Kr and Xe re- moval from the reactor with a 30-sec residence time, and the lowest solid curve represents re- moval of Kr and Xe (30-sec residence) along with removal of Mo, Tc, Ru, Rh, Pd, Ag, In, Nb, Te, and Se with a mean residence time of 1.8 x 10° T 1 [ ] ‘ i T [ oo | \ a ’ | 0 i | ‘ o | { ] 0" | o 4#5“ = NO REMOVAL IN REACTOR VOLATILE - . - |l : l| P i s 1 | r == FLUCRIDES REMOVED IN PROCESSING -1 707 & | } Lo by i A 1 1 o I - . ERE I . .. tr . ' el A R T”]L,L ’ T ! .} L T T _ .~ Kr—Xe REMOVED IN REACTOR | } o | | o 5 | N =k » | - o 10 e o I : C N o I ==} 1 0l TSad T Kr-Xe AND PLATING ELEMENTS . | - e , | gf: REMOVED IN REACTOR Lo L = I Kr-Xe REMOVED IN REACTOR |71 | = RN o S | = | VOLATILE FLUCRIDES REMOVED| . | e b o "7 IN PROCESSING | l | % MO REMOVAL ST T - 17 B Kr-Xe AND PLATING ELEMENTS REMOVEGC | > i Lol Tl = L ) IN REACTOR VOLATILE FLUORIDES REMOVED | i [l i Lo = - ;== IN PROCESSING - ' | S = { i ' o ‘ i | ! ‘ : 5ot D P - 4 Cobee Lo © 1 Dl oo [ T b b Z ; $o | o - ' i : 5 Gl e £ ol e " FUEL = 23yF, ' T . POWER PER REACTOR = 556 Mw (thermal) == = | -~ FUEL VOLUME (iotal} = 224 #2 L - PROCESSING CYCLE TIME = 52.08 days o : P oo L | i 1 ! | | i ; 3 ‘ i i i N Lo Ly 5 10" 2 5 10° 2 TIME AFTER REMCVAL FROM REACTOR {days) Fission Product Heat Generation Rate in M$BR Fuel After Removal from the Reactor. 247 sec (50 hr). These latter residence times are rough estimates, and as better information becomes available from figsion product behavior in the MSRE, more reliable estimates of the residence times for these elements will be possible. This lowest solid curve in Fig. 23.1 is shown only to supggest the general range of conditions under which the fuel processing plant may operate, The dashed curves in Fig. 23.1 were obtained from the same reactor calculations, but some fis- sion products were assumed to be removed in the fluorinator. The elements Xe, Kr, Br, I, Mo, Tc, Te, and Se were considered to be completely re- moved, and 15% of the Ru, Rh, Nb, and Sb were removed. After fluorination, however, these elements may ‘‘grow’’ back into the system. HEAT GENERATION IN A MOLTEN-SALT STILL Buildup of heat generation in a molten-salt distillation system was calculated using the heat- generation data given in Fig. 23.1. The reactor and that part of the processing system prior to the still were assumed to be at steady state, although the transient associated with buildup of fission products in the still was considered. TFuel salt containing fission products remaining after fluorination was fed to a distillation system, and complete retention of fission products was as- sumed. The fuel salt was assumed to have been held up 24 hr in processing prior to entering the still (12 hr before fluorination and 12 hr after), The heat-generation data from Fig. 23.1 was fitted by the method of least squares to the rela- tion 15 Kt Ho=Y 4de 7, i=1 where H(t) == heat-generation rate at time ¢ (Btu hre ! £t d), A k, = constants, t - time after salt leaves fluorinator (days). The rate of heat generation in the still is then given as A —k. t Q(t)r:Ff H(x)dY—Fz (I—-e 1), 0 i=1 i where 1) = heat generation in still after operating for a time ¢ (Btu/hr), F - fuel salt processing rate (ft3/day), t = length of time still has operated (days). Calculated heat-generation rates in the still are shown in Fig. 23.2 for a processing rate of 15 ft*/day for several assumed removal efficiencies (the same assumptions noted for Fig. 23.1) in processing steps prior to the still. The still system will be near steady state in two to three years, and heat generation rates as high as 3.0 Mw can be expected. Removal of fission products by gas stripping, plating on metal surfaces, and formation of velatile fluorides during fluorination will reduce this rate to about 2.2 Mw, DRNL~DWG 67-97384 10" & Er T T T T T ST P R e P T L~ REACTOR POWER 2224 Ww (thermal) | |T— AT “:ijiiiiFi,“ T .- PROCESSING CYCLE 8208 Davs 1]l L ’ : L Al * 11 | Cbeemmmemen L LD """" T 77T '::““ e o e T | 3 | | | 2,0 e ' L b D HO REMOVAL 1N REACTOR SR = — g I /}4,:,6{ - VOLATILE FLUCRIDES REMOVED IN PROCESSING - 1 @ T JTiT RN N AN Z o LT T K - Xe REMOVED I REACTOR LT g ,,,,;fi‘/,;Fi [H - xo» ATHLE FI UOR\D“S RCWOVED IN PROCESSING....\.... P VA AT I | \ b Lt ! o A T Ir(r -Xe AND PL :\rmc H FNFN[: REMOVED IN REACTOR & ‘ & . 7 VOIIATILIFJF\ L(‘RIEF) REMOVED 11 PRO(FS"S NG : | by J L L Lol 1 i o REMOVALE AT SR ST - ) ST LT CUTTRITIY | . I i } ,,,,,,,,,,,,,, 1144 1 e dibo J\ L \ , L] L] Ll —— ‘ ‘ 4 : 1 il 10— oL L L v it oLl S 4of ! 13 10" 10° THIE (days) Fig. 23.2. Fission Product Decay Heat in MSBR 5till and Accumulation Tank. 24. Reductive Extraction of Rare Earths from Fuel Salt L. M. Ferris C. One alternative to the distillation process for decontaminating MSBR fuel salt involves reductive extraction of the rare earths (and other fission products) after the uranium has been recovered from the salt by fluorination.' ™ The reductant currently being considered is ‘Li dissolved in molten metals such as bismuth or lead which are immiscible with the salt. Studies in the Reactor Chemistry Division''?'* have defined the equi- librium distribution of several rare earths between Li-Bi solutions and LiF-BeF, (66-34 mole %) at 600°C. The results of these studies indicate that the rare earths can be effectively extracted in such a system. The apparent stoichiometry of the extraction reaction is such that, under the experimental con- ditions employed, the lithium concentration in the metal phase should not change detectably, even if the rare earth were quantitatively extracted. However, in these, and similar, experiments, usually 50 to 75% of the lithium originally present in the Li-Bi solution apparently was consumed.? The mechanism by which lithium is ‘‘lost’ in these systems has not yet been elucidated. In order to properly evaluate the practicality of the reductive extraction process, the behavior of lithium in these systems must be clearly defined. Knowledge of its behavior is especially important in designing a multistage extraction system, be- cause the extent to which the rare earths are ex- 'rR. B. Briggs, MSIR Program Semiann. Progr. Rept. Aug. 31, 1966, ORNL-4037. M. W. Rosenthal, MSR Program Semiann. Progr. Rept. Feb. 28, 1967, ORNL-4110, ’p. E. Ferguson, Chem. Technol. Div. Ann. Progr. Rept. May 31, 1966, ORNL.-3945, 4W. R. Grimes, Reactor Chem, Div, Ann, Progr. Rept. Dec. 31, 1966, ORNL-4076. i . Schilling 248 J. F. Land tracted (at equilibrium) is directly related to the lithium concentration in the metal phase,!-?* Consequently, mote experiments have been con- ducted, with the primary objective being the de- termination of the cause of the apparent lithium loss. As discussed below, these experiments did not produce a definite answer to the question, and more study will be required. In each experi- ment a rare-earth fluoride, usually Equ, was added to the salt to provide a basis for compari- son with the results of earlier studies and to allow us to obtain preliminary information on the effect of temperature on the distribution of europium be- tween the two phases. Experimentally, solutions of lithium in bismuth (usually about 7 at. % Li) and of EuF, in LiF- BeF (66-34 mole %; mole fraction of EuF |, about 10=%) were prepared at 600°C in separate mild steel or graphite crucibles, using pure argon as a blanket gas. Then, the salt was transferred to the crucible containing the Li-Bi alloy, and the system was equilibrated at high temperature in an argon atmosphere, Filtered samples (stainless steel samplers) of both phases were removed periodically for analysis. No detectable ““loss’’ of lithium occurred during preparation of the Li-Bi alloys in either mild steel or graphite crucibles, When the system was not agitated (by sparging with argon), the time required to achieve equilibrium was generally about 24 hr, in bismuth when the system is quiescent has been observed by others.! The rate-controlling step may be the dissolution of Li Bi, a high-melting, insoluble compound that probably forms rapidly when mixtures of the two metals are heated. After about 24 hr at 600°C, the lithium concentra- tion in the solution, as determined by chemical The low rate of dissolution of lithium analysis of filtered samples and thermal analysis of the system, usually reached the expected value, When the system was sparged with argon, equilibrium was reached in 1 to 2 hr or less. As was the case in other studies,’ a significant (30 to 70%) decrease in the lithium concentration in the metal phase occurred during equilibration of salt and Li-Bi solutions at 500 to 700°C. Several possible causes of this phenomenon were considered: (1) dissolution of lithium in the salt, (2) reduction of BeF from the salt by lithium with formation of LiF and berylium metal (which is insoluble in both the salt and bismuth), and (3) reaction of the lithium with water that was in- advertently admitted to the system. Neither of the first two mechanisms seems likely, based on the experimental evidence. If either lithium or beryllium metal were present in the salt in the amounts expected from the lithium ‘‘loss,”” dis- solution of the salt in hydrochloric acid should have resulted in the evolution of more than 10 cc (STP) of hydrogen per gram of salt. However, it was found that samples of both filtered and un- filtered salt generally gave less than 0.2 cc of H, per gram. This finding, along with thermo- 249 dynamic considerations,? appears to rule out either of these mechanisms. The inadvertent admittance of water into the system (as a contaminant in the salt and/or blanket gas, or during sampling) seems to be a more reasonable explanation for the consumption of lithium. Although this hypothesis has not yet been confirmed, several observations have heen made which support this mechanism, The salt after equilibration is invariably permeated with an ingoluble material which tends to concentrate at the salt-metal interface. The presence of this phase does not appear to be related to the presence of a rare earth, and, on the basis of chemical analysis, is not the result of corrosion of the crucible or samplers, High oxygen concentrations (about 0.5%) have been detected in salt samples, and, in a few instances, the presence of BeO in the system has been confirmed, These results, if caused by the presence of water, are consistent with a mechanism in which the water reacts first with BeF , to form insoluble BeO and gaseous HF which, in turn, reacts with lithium to form LiF that dissolves in the salt, Accordingly, the LiF concentration in the salt would increase and the Table 24.1. Distribution of Europium Between LiF-BeF, (66-34 Mole %) and Lithium-Bismuth Solutions L.ithium Approximate Amount Concentration in Temperature of Europium in Experiment Sample Metal Phase o Metal Phase D (mole %) (7o) CES75 1 0,85 608 525 1.20 CES66 3 2.6° 602 81° 5.4° CES66 2 3.4% 583 93P 17.1% JFL64 3 3.97 500 92°¢ 14.8° JFL64 1 4.0P 605 84° 6.7° JFL64 2 4.0° 700 88¢ 9.1¢ JFL64 1 4.84 605 84¢ 6.7 CES66 1 5,00 583 kL 16.8” JFL64 2 5.19 700 88° 9.1°¢ mole fraction of Eu in metal phase “Defined as D = mole fraction of Eu in salt phase PEmission spectrographic analyses. “Neutron activation analyses. A lame photometric analyses. 260 BeF concentration would decrease. The changes in salt composition calculated from the amount of lithium consumed were too small to be detected by chemical analysis; however, in some instances x-ray diffraction analyses of salt samples that had been cooled to room temperature revealed the presence of LLiF, a phase that would be expected if the LiF/BeF | mole ratio in the salt were higher than its original value of about 2. Ob- viously, more work, conducted under very care- fully controlled conditions, will be required to confirm or refute the tentative hypothesis that water is the cause of the apparent loss of lithium in reductive extraction experiments. Despite significant changes in the lithium con- centration in the metal phase during an experiment, it was possible to determine the distribution of europium between the two phases at various tem- peratures. This distribution is expressed as a ratio, D, defined?* as mole fraction of Eu in metal phase mole fraction of Eul", in salt phase ' The values of D obtained in this study are given in Table 24.1 and are compared in Fig. 24.1 with those obtained by Shaffer, Moulton, et al.? at 600°C. Agreement between the two sets of data is reasonably good. It also appears that tempera- ture has no marked effect on the equilibrium dis- tribution of europium between the two phases. An emission-spectrographic or a neutron-activa- tion method was used to analyze for europium in both the salt and metal phases. The data reported are from experiments in which the europium balance was 90 to 110%. FEach sample of the metal phase ORNL-DWG 67-11848 20 TEMPERATURE © (°c) + 500 583 c02 5 A 605 608 700 maie fraction of kuin merai phase moig frection of Eu in sait phase 05 05 ! 2 5 10 LITHIUM CONCENTRATION IN METAL PHASE (at %) Fig. 24.1. Distribution of Europium Between LiF- BeF2 {66-34 male %) and Li-Bi Solutions. Data points, this study; solid line, data of Shaffer, Moulton, et al. (see Reactor Chem. Div. Ann. Progr. Rept. Dec. 31, 1966, ORNL-4076). was analyzed for lithium by both an emission spectrographic and a flame photometric method. In some instances, the values obtained were markedly different; this is readily apparent in the data obtained from experiment JFI1.64 (Table 24.1). 25. Modifications to MSRE Fuel Processing Facility for Short Decay Cycle R. B. In order to keep reactor downtime to a minimum during the next planned shutdown, it has been pro- posed that the flush and fuel salts be processed with a minimum decay time. Problems anticipated because of the short decay are due to the much larger amount of '3[ and to the radiation level at the UF _absorbers because of coabsorbed ??Mo. The processing schedule, with pertinent radiation information, is given in Table 25.1. Design of modifications to the MSRE fuel processing facility is in progress to permit safe operation at these increased activity levels. These consist of: Lindauer 1. A deep-bed backup charcoal trap downstream of the existing charcoal canisters. These traps will be tested under simulated operating conditions to demonstrate at least 99.998% re- moval of 1*11, 2. An activated alumipa trap downstream of the SO_ system for excess fluorine disposal. This will prevent fluorine from releasing iodine on the charcoal beds in case of a malfunction of the SO2 system. 3. Shielding and revised handling procedures for removal of the loaded UF6 absorbers from the absorber cubicle, Table 25.1. MSRE Processing Schedule and Pertinent Radiation Levels ISII Uranium Decay Required Removal to Be Radiation Dose Rate Operation Time Curies in Charcoal Volatilized 2 ft from f\bsorber (days) in Salt Traps® (k) (millirems /hr) (%) Hz—HF treatment 6 600 99,95 of flush salt Fluorination of 18 210 99.88 7 620 flush salt HQ-HF treatment 28 23,000 99.990 of fuel salt Fluorination of 35 12,000 99,997 230 170 fuel salt AFor 0.3 curie release. 251 26. Preparation of 2*3UF,-’LiF Fue! Concentrate for the MSRE J. M. Chandler Plans are now being made for refueling and op- erating the MSRE with 273U fuel early in 1968,; approximately 40 kg of ***U as ?**UF -"LiF (27 and 73 mole %) eutectic salt will be required. This material must be prepared in a shielded fa- cility because of the high 22?U content (240 ppm) of the ??3U. The following sections describe the process, equipment, and status of the fuel prep- aration process. PROCESS The process for preparing eutectic-UF -LiF (27 and 73 mole %) salt for the original MSRE charge started with UF, and LiF.' Since **°U is avail- able only as the oxide, the process was modified. The UO3 is charged to a nickel-lined vessel, and LiF is added on top of it. The charge is heated, under helium, to 900°C to melt the LiF, which wets the oxide and thermally decomposes the UO, to an average composition of UO, - The charge is treated with H, to reduce the uo, | to UO2 and then hydrofluorinated with HZ-HF (approximately 10:1 mole ratio) to convert the UO2 to UF4. The progress of the reaction of UO2 to form UF, is followed by the changes in the liquidus tempera- ture of the binary mixture as the melting point of the charge decreases from the initial melting point of 845°C for the LiF to 490°C for the UF -LiF (27 and 73 mole %) eutectic product. Hydrogen fluoride utilization and water production are also indicative of the progress of the reaction. After 'MSR Program Semiann. Progr. Rept. Feb. 28, 1965, ORNL-3812, pp. 149-50. 252 E. L. Nicholson a final H, sparge to reduce impurities, such as iron and nickel fluorides, to metals and to sweep out dissolved HF, the salt is sampled, filtered, and stored in a product container. This process has been satisfactorily tested on a small scale, and the results indicate that the processing time will be about 15 days/batch. Tentative plans are to make three 12-kg 233U batches of salt, which will be used for the major additions to the barren salt in the MSRE, and one 7-kg 233U batch, which will be loaded into 60 en- riching capsules. EQUIPMENT AND OPERATIONS The schematic equipment flowsheet for salt preparation is shown in Fig. 26.1. The cans of UO,, each containing about 500 g of uranium, will be transported in a shielded cask from the *33U storage facility in Building 3019 to the TURF building (7930) and introduced into cell G. All operations in cell G will be done remotely. By the use of manipulators, the cans will be loaded into the alpha-sealed can opener and the dumpet box, which discharges the oxide into the heated nickel- lined 8-in.-diam by 3-ft-high reaction vessel and routes the empty cans to storage for subsequent removal from the cell. The "LiF is charged to the reaction vessel via the dumper. Process gases are supplied from outside the cell, and instrumentation is provided to control gas flows and processing temperatures. Waste gases are filtered to remove entrained 233U and scrubbed with KOH to remove unreacted HF before dis- posal to the radioactive off-gas system. After sampling, the product is transferred via a heated 253 ORNL-DWG 67-- 3644 A GAS KOk ANALYSIES e . CAS SAFETY Ta ] OFF=GAS =1 goqopck TRAR [ OHH v X ES’EUOS 1l [ ALUMINUM LiF e WASTE Y et CANS CHARGING GAS ANALYSIS HYDROGEN FLUORIDE ™7 HYDROGEN —i> 233 - UF, - LiF PRODUCT STORAGE G-in. DIAM. ----- CAN DPENER AN DUMPER SMETY CAN STORAGE FUHTER } SALT GAS FLOW -- RIEAGCTION VESSEL 8-mn. DIANM. Fig. 26.1. MSRE 233U Fuel Salt Preparation System. nickel line through a sintered metal filter into heated 4-in.-diam by 3-ft-high nickel product storage vessels. The capsule-filling operation is not illustrated but will consist in transferring salt from a product vessel to an array of 60 en- riching capsules in a furnace. Product vessels and enriching capsules will be stored in cell G until needed at the MSRE. STATUS Engineering design is estimated to be 95% com- pleted, and fabrication of the equipment is 85% completed. The project has been delayed because the TURF building is not completed; consequently, the cost-plus-fixed-fee (CPFF) contractor and the laboratory forces have been delayed in starting the installation of equipment. Assuming that the build- ing will be available on September 15, 1967, and that the CPFF contractor and ORNL forces each require about one month for their portions of the job, the full-scale cold run should start December 1, 1967; hot runs should start January 20, 1968, and be completed in mid-April 1968, 255 OAK RIDGE NATIONAL LABORATORY MOLTEN-SALT REACTOR PROGRAM AUGUST 33, 1967 M. W. Rasenthal, Director R R. B. Briggs, Associate Director D P. R. Kasten, Associate Director R MSBR DESIGN STUDIES COMPONENTS & SYSTEMS DEVELOFMENT‘I PHYSICS MATERIALS FUEL PROCESSING DEVELOPMENT CHEMISTRY MSRE OPERATIONS E. 5. Bettis R Dunlap Scott R A. M. Perry* R J. R, Weir M&C M. E. Whatley cT W. R. Grimes* RC P. N. Haubenreich R H. E. McCay M&C E. L. Nicholson T C. E. Bettis* GE MSRE PHYSICS HASTELLOY N STUDIES M5BR PROCESSING O, W. Burke™ |1&C W. L. Carter’ CcT B. E. Prince R D. A, Canenice* MAC J. 0. Blomeke* cT F.H. Clark* 18C PUMP DEVELOPMENT A G. Cepolina” MAC W. L. Carter €1 REACTOR CHEMISTRY COORDINATION NUCLEAR AND MECHANICAL ANALYSIS C. W. Collins R MSBR ANALYSIS R. E. Gehlbach* M&C L. M. Ferris cT W. K. Crowley” £ A. G. Grindeli R O, L. Smith R A P. Litman* MBC 8. A. Hannaford cT H. F. McDuffie* RC C. H. Gobbard** R J. R. Engel R 5. J. Ditte* 1L P. G. Smith R "n T Ken R T. 5. Lundy* MAC J. R. Hightower LT R. E. Thoma** RC €. H. Gabbard™* R W. P. Eatherly uce L. V. Wilsen R T C. E. Sessions MsC 1. C. Mailen T =. F. Blankenship RC J. A Watts R A, G. Grindell R D. D. Owens R MSBR EXPERIMENTAL PHYSICS G. M. Slaughter* M&C L. E. M.cNeese T R. B. Lindaver* cT J. P. Nichols cT MSRE ON-SITE CHEMISTRY GRAPHITE STUDIES E senilli e o Liewellyn” e B b g " ey T R. E. Thoma'* RC OPERATION R. L. Moore* 18C H. Beutler* MLC F. J. Smitl ~ 5 noma H. A. Netms® CE CONSULTANT W. H, Caok MAC J. 5. Watson g 5. 5. Kirsfis RC R. H. Guymon kR INSTRUMENTATION AND COMTROLS L. C. Qokes™ 14C COMPONENTS & SYSTEMS . . €. R. Kenned M&C J. Beams T w Pkl cE T W. Keriin uT ¥, C. Robimeon® MaC V.L. Fowler T PHYSICAL AND INORGANIC CHEMISTRY 2 Crowley R — L. C. Ockes 18C - - * . . v > R. C. Robertson® R Dunlap Scott 1. F. Lend cT C.F. Baes, Jrt RC T Huds:n R g K. Adams. l&g J. R. Rose GE GRAPHITE-METAL JOINING R. D _Fr’ayne ET C: J.- Bar!a’r\“ ’ RC A, I. Krakoviak R C- 2 :u::n ~ ;ic W. C. Stoddar™ %EC COMPONENTS W. Castleman* ML €. T Thompsan T J. Breunstein® RC R. R. Minue R L R:df::“: &G d' ? Tallackson . R. B. Gatloher R K. V. Cook* MAC MSRE PROCESSING G. D. Brunton* RC . Richardson R R. W. Tucker JAC - Terry 8 R J Kedl R 3. P. Hommond* M&C 5. Cantor” RC R. Stefty R L. E. Basler 1%L E' L W!:Hs . R T. H. Mouney RC W. J. Verner* M C R. B. Lindauer** cr J. H. Shatier* RC T. Amwine R J. Campbell* 1ac Y. Wilson R AN Smith R e 7 R. A. Strehlow* RC W. H. Duckworth R 5. E. Ellis* 1&C W. C. George R 2 E Comes R TECHNICAL SUPPORY PREPARATION OF 2*3UF 7LiF FUEL FOR MSRE R. L. Thomo** RC J. D. Emch R e E Rickwood* 18 H. M. Poly § S . . i - E €. A. Gifford R M. D. Alien* MaC E. L. Nicholson T G e oo Re T o ; R. W. Cunningham* MaC J. M. Chandler cT - L. Bamberger RC C- C. Hurtt R REMOTE MAINTENANCE 1. C. Feimer MAC 5. Mann cr FoiQoss RC 1. C. Jor don R - —~ - . nedmoern - T R. Blumberg R -Jr #. Goer 1 AL A . Rolm t‘ L. Q. Gilpatrick* RC H. J. Millar R F. 0. Horvey MAC W. F. Schotfer < 2 CHEMICAL PROCESSING J. R. Shugart R D. M. Moulton RC L. E. Penton N . R. Shug, V. G. Lane M&C E. L. Youngblood cT — E. J. Lawrence MAC J. D. Re¢man'* RC W. E. Ramsey R R. B. Lindeuer** cT L K. A. Remberger™ RC R. Smith, Jr. R CONSULTANT E. H. Lee* M&C DR S N RC J. L. Stepp R - . - JeQrs - . F. N. Pecbies uT L C- Monley MaC H. H. Stane* RC 5. R. West R N W. J. Mason* M&C E. W B. McNabb MEC ¥W. K. R. Finnell RC J. E. Wolfe R RA P‘Ldge"* VEC 3. F. firch RC S. J. Davis R N. 0. Pleasant MEC _C E. R°!’°"s EE R S Pultiam® MAC W. P. Teichert R. G, Shooster* mé&C W. H. Smith, Jr.* M&C L. D Stack* M&C DESIGN LIAISON G. D. Stohler™ M&C C. K. McGlathlan R 0 R Tinch MaC IRRADIATION EXPERIMENTS R Stoter . E. M. Thomas M&C E. G. Bohlmann* RC F. J. Underwood GE 1. i. Woodhouse* M&C E. L. Compere RC H. £. Savage RC J. M. Baker RC — J. R. Hort RC MAINTENANCE B. L. Jahnsan RC B. H. Webster R o J. W. Myers RC AE Gill R. M. Waller RC - F. Gillen R L. P. Pugh R ANALY TICAL CHEMISTRY DEVELOPMERT AC ANALYTICAL CHEMISTRY DIVISION J. C. White* AC CT CHEMICAL TECHNOLOGY DIVISION . D DIRECTORS DIVISION & DEVELOPMENT GE GENERAL ENGINEERING DIVISION RESEARCH M 1£C INSTRUMENTATION AND COMTROLSS DIVISION A, S. Meyer AC M&C METALS AND CERAMICS DIVISION R. F. Apple AC R REACTOR DIVISION C. M. Boyd AC RC REACTOR CHEMISTRY DIVISION J. M. Dote AC UCC UNION CARBIDE CORP H. W. Jenkins uT UT UNIVERSITY OF TENNESSEE G. Maomantav* uT * PART TIME ON MSRP D. L. Murlmrlg" fig ** DUAL CAPACITY ; R Maller o _ L. Whiting 1. P. Young" AC ANALYSES L. T. Corbin* AC G. Goldberg” AC F. K. Heacker AC C. E. Lamb* AC P. F. Thomason* AC 0 ONO LA LN ErELMOCOE0VOOONEIAMICONMPAIS-TDOMOITVINN-NATAIOX W4T >rPr==X . K. Adams . M. Adamson . G. Affel . G. Alexander . F. Apple . F. Baes M. Baker J. Ball . J. Barton F. Bauman E. Beadll Bender . Bettis Bettis Billington . Blanco . Blankeriship . Blomeke OCmmwem . Blumberg L.. Boch . G. Bohlmann . J. Borkowski . E. Boyd Braunstein Bredig Breeding . Briggs . Bronstein . Browning . Bruce . Brunton . Burger Canonico Cantor PITTAAMAUET = > . W. Cardwell . L. Carter . I. Cathers . Chandler . Collins . Compere Cenlin . Cook . Corbin . Cottrell INTERNAL DISTRIBUTION 257 59. 60. 61. 62. 63. 64. 65. 66. 67. 68. 69. 70. 71. 72. 73. 74. 75. 76. 77. 78. 79. 80. 81. 82. 83. 84. 85. 86. 87. 88. 39. 90. 91. 92. 93. 94, 95. 96. 97. 98. 99. 160. 101. 102. G. S. . L. Crowley . L. Culler . M. Ddle . G. Davis . W. Davis . J. DeBakker ME-PIMIONDIPRARPETINRNETO-IPIUEMEZEOP ZAL P ETEME OXATMNOPZLOIPTIOAMOBIITLT>VEIMATAAIT>Y MO ORNL-4191 UC-80 — Reactor Technology A. Cristy J. Cromer (K-25) H. DeVan J. Ditto . Donnelly . Dunwoody Dworkin . Dyslin . Eatherly . Edlund (K-25) Engel . Epler . Ergen . Ferguson . Ferris . Fraas . Friedman Frye, Jr. . Gabbard Gall . Gallaher . Gilliland . Goeller Grimes . Grindell . Guymon . Hammond . Hannatord . Harley . Harman Harrill . Haubenreich . Heddleson . Herndon . Hibbs {Y-12) . Hightower . Hitl . Hise 103. 104. 105. 106. 107. 108. 109, 110. 111-115. 116. 117. 118. 119. 120. 121. 122. 123. 124. 125. 126. 127. 128. 129. 130. 131 132. 133. 134. 135. 136. 137. 138. 139. 140. 141, 142. 143. 144. 145. 146. 147. 148. 149. 150. 151. 152. 153. 154. 155. 156. 157. — % ——— 2OmMP>r0O0; IqufiénUmmIEZOp%?fifim@%fi&?pmp;fipg;ppgIa>mm