ORNL=3369 UC-80 — Reactor Technology MO LTEN-SALT REACTOR PROGRAM SEMIANNUAL PROGRESS REPORT FOR PERIOD ENDING AUGUST 31, 1962 OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. DISCLAIMER Portions of this document may be illegible In electronic image products. Images are produced from the best available original document. Printed in USA. Price: $2.75 Available from the Office of Technical Services U. S. Department of Commerce Washington 25, D. C, LEGAL NOTICE This report was prepared as an account of Government sponsored work. Neither the United States, nor the Commission, nor any person acting on behalf of the Commission: A. Makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of any information, apparatus, method, or process disclosed in this report. As used in the above, ‘“person acting on behalf of the Commission’’ includes any employee or contractor of the Commission, or employee of such contractor, to the extent that such employee or contractor of the Commission, or employee of such contractor prepares, disseminates, or provides access to, any information pursuant to his employment or contract with the Commission, or his employment with such contractor. | B A ORNL-3369 Contracf No. W-7405-eng-26 MOLTEN-SALT REACTOR PROGRAM SEMIANNUAL PROGRESS REPORT For Period Ending August 31, 1962 R. B. Briggs, Program Director Date Issued DEC 12 1962 OAK RIDGE NATTONAL LABORATORY Oak Ridge, Tennessee . operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION e - THIS PAGE WAS INTENTIONALLY LEFT BLANK iii SUMMARY Part 1. MSRE Design, Engineering Analysis, and Component Development 1. MSRE Design No significant changes in design concept or in detail of any component or system were made. Design work was essentially completed, and a design report giving all engineering calculations and analyses of the systems is being compiled. The remaining design work involves primarily the incor- poration of results of development work into the final drawings. Work on the major building modifications is proceeding as scheduled, - and the estimated completion date is October 15, 1962. A contract was - (w awarded for construction work required outside the building, and this work too is to be completed soon. It consists of construction of the exhaust filter house, the cooling towers, the underground piping, and the inlet filter house. Materials procurement is on schedule with the exception of the core graphite, delivery of which is now expected November 1, 1962. Fabrication of components is approximately 50% complete and on schedule. The basic instrument criteria and the control philosophy were estab- lished. The layout of the instrument and controls system was established, and the design of control-panel interconnection facilities is under way. Designs for 30 of the 42 instrument panel sections required in the MSRE system are complete. Fabrication of these panels by the ORNL shops is under way. ) A freeze-valve test facility was completed and placed in operation. ./' . A The selection of detectors for the process and personnel radiation monitoring system is approximately 80% complete. Procurement of a data-handling system was initiated. Specifications were prepared and procurement initiated for 80% of the components required in the process instrument system. 2. MSRE Reactor Analysis Conceivable reactivity accidents were analyzed to permit evaluation of reactor safety. Most of the calculations were done using Murgatroyd, a digital kinetics program, but some analog analyses were also used. None of the accidents analyzed led to-catastrophic failure of the reactor. Undesirably high temperatures were predicted, however, for circumstances associated with extreme cold-slug accidents, premature criticality during filling, and uncontrolled rod withdrawal. iv Criticality and flux calculations were revised to correspond to the latest fuel composition and the detailed design of the core. Calculations were done for a core model consisting of 19 regions with different volume fractions of fuel, graphite, and INOR-8. The clean critical concentration for the reactor fueled with salt containing no thorium was calculated to be 0.15 mole % uranium (93% UZ3%), High-energy fluxes were also calculated. At a reactor power of 10 Mw, the maximum flux was about 1.3 X 1013 neutrons/cmZ.sec for neutrons having energies greater than 1 Mev; for neutrons having energies greater than 0.1 Mev, the maximum flux was about 3.7 x 102 neutrons/cmZ-sec. _ Fluxes and power densities from the Equipoise calculations were used, together with the detailed flow distribution in the MSRE core, to predict the gross temperature distributions in the fuel and graphite. At 10 Mw, with fuel entering the core at 1175°F, fuel in the hottest channels left the reactor at 1261°F. Under the same conditions the peak graphite tem- perature was 1296°F, A low-temperature region existed through the center of the core where the fuel velocity was above average. The nuclear average temperature, defined as that uniform temperature which results in the same reactivity as the actual distribution, was cal- culated for the fuel and for the graphite. Perturbation theory was em- ployed, using coefficients from the Equipoise calculations. At 10 Mw, with the fuel inlet at 1175°F and the mixed-fuel outlet temperature at 1225°F, the nuclear average temperatures for the fuel and graphite were 1213 and 1257°F, respectively. Importance-averaged temperature coefficients of reactivity were cal- culated for the case of fuel containing no thorium. The results were A hs x 1075 and -T7.27 x 1072 (8k/k)/°F for the Tuel and the graphite, regpeetively. Power coefficients of reactivity were calculated from the nuclear average temperatures and reactivity coefficients. This quantity, defined as the reactivity adjustment required as the power is raised, depends upon which temperature is used as the control parameter. If the mean of the fuel inlet and outlet temperatures were held constant, the power coef- ficient of reactivity would be -L4.7 x 10™* (8k/k)/Mw. If the fuel outlet temperature were held constant, the power coefficient would be only -1.8 x 107% (dk/k)/Mw. Effective delayed-neutron fractions were calculated, taking into account the delayed-neutron energies and the spatial source distributions in a one-region model of the core. For a total yield of 0.0064 delayed neutron per fission neutron, the effective fraction in the static core was 0.0067. With fuel circulating at 1200 gpm, the effective steady-state fraction was 0.0036. ' Preliminary tests were made of a new IBM-7T090 program for the analysis of simulated power transients, based on a simplified space-dependent ki- netics model. This program calculates fuel and graphite temperature Iz i 7 A\ LY &\ [ A e I“ & L distributions during transients and uses the nuclear-average-temperature concept to relate temperature changes to the reactivity. 3. Component Development Measurements were made on INOR-8 freeze flanges for 5-in.-diam sched- 4LO pipe under several different steady-state and transient conditions with both an octagonal and an oval nickel ring gasket. The following conclusions were drawn from these measurements: the flange joint design appears to be satisfactory from an accumulated thermal stress standpoint after 43 cycles from room temperature to 1300°F and higher; the thermal distortion of the flanges is small in the operating range and is not of a nature to cause the flanges to separate at the seal; thermal cycling does not cause the gas leakage rate to increase above an acceptable level; the flange Joint will not form a frozen plug in the pipe under no-flow conditions if the pipe heaters are kept on; the thermal and distortion properties of the INOR-8 flanges are superior to those of the Inconel flanges tested previ- ously; and thermal cycling of one flange did not affect its ability to seal with a new flange, A full-scale conceptual model of the control rod and drive unit was assembled and tested under operating conditions for over 24,000 cycles. Difficulty with & bushing was solved by replacement with bail bearings. Over-all operation was satisfactory. Evaluation of a removable 5-in. pipe heater insulated with hardboard and mineral wool indicated excellent performance except for the tendency to produce dust after high-temperature operation. Nuclear activation analyses of insulating materials indicated that the dust would present a problem of air-borne activity during maintenance. Another unit, con- structed of foamed fused-silica, was better with respect to dusting but had an excessive heat loss. An improved fused-silica pipe heater, incor- porating some canned hardboard is being fabricated. An all-metal reflec- tive type of unit was procured for evaluation. The prototype cooling bayonets for removing fuel afterheat from the drain tanks were thermally shock tested 384 times without failure. Reactor- grade thermocouples were still intact after 6 cycles, whereas other types of thermocouples failed in fewer cycles. Detail drawings of essentially all the components and instrumentation of the sampling and enriching system were finished, and construction of a mockup was started. The sampling-capsule access-chamber mockup passed the hydrostatic test requirements, and hot cell removal of a sample from the capsule was demonstrated. The full-scale MSRE core model was operated at 85°F with water to make preliminary measurements of the velocity distribution, heat transfer coefficients, and solids-handling characteristics. The velocity distri- bution in the lower core wall-cooling annulus was found to be uniform around the circumference, indicating that the flow into the lower head was also uniform. The heat transfer coefficients in the lower head agreed Vi with the predicted values, except near the axis, where the measurements were irregular because of the random flow. Preliminary measurements of the solids distribution in the lower head indicated that solids which do remain in the core tend to collect near the drain line at the center. It appeared, however, that they would not plug the drain line. The engineering test loop was revised to include a graphite container and INOR-8 pipe and was placed in operation with a newly made zirconium- free coolant salt. Information was obtained of the operating character- istics of the frozen seal in the graphite container. It was found that a salt mist existed in the pump-bowl gas space that led to some difficulties with frozen salt deposits, but no such mist was found in the drain-tank gas space. After 1500 hr of operation with the empty container, graphite / was installed, vacuum pretreated at 1200°F, and prepared for flushing with salt that had been treated with HF to lower its oxide content. Design and construction of equipment for opening and closing freeze- f'lange joints was completed, and procedure testing was started. A peri- scope was purchased for use with this equipment. The design of a portahle shield facility for semidirect maintenance operations is 60% complete. Tt is a motorized unit with the appropriate openings and lighting for specific operations. Six prototype brazed joints were fabricated for quality evaluation "using the equipment and general techniques required for remote operation. A similar joint was removed from a loop for metallurgical examination after 6780 hr at 1250°F and exposure to fuel salt for 123 hr. The design of the overflow-line disconnect to the fuel pump was com- pleted, and a carbon-steel prototype was built and successfully rough- tested as a demonstration of the concept. An INOR-8 model was constructed: for finel testing. The design drawings for the coolant pump, the drive motors of the fuel and coolant pumps, and the lubrication stands were approved. The assembly of the rotary element and the fabrication and assembly of the hot test stand for the prototype fuel pump were completed and testing was started. Acceptable INOR-8 castings of impellers and volutes for the fuel and coolant pumps and dished heads for the pump tanks were obtained. Fab- rication of all components required for the reactor pumps was initiated. A prototype model of the two-level conductivity-type liquid-level probe being developed for use in measurement of level in the MSRE storage tanks was constructed and is being tested. Developmental testing of an element for continuous measurements of the molten-salt level in the reactor pumps was continued. The maximum variation recorded during a four-month test was 0.1 in., and there has been no significant degradation in performance characteristics to date. Investigations of single-channel thermocouple alarm switches for use in freeze-flange and freeze-valve monitoring continued. A commerical system is undergoing field .tests. 0 vii Development of the mercury-jet-commitator thermocouple-scanning system is continuing. Modifications of the switch were made to reduce spurious noise generation. An alarm discriminator was designed and con- structed. A complete system was assembled and tested. Testing of mechanical attachments for use on radiator tubes in the MSRE was resumed after modifying the test apparatus. The differential between indicated inner and outer wall temperatures under simulated opera- ting conditions was reduced to 9°F when the thermocouple was insulated with Fiberfrax paper and Thermom cement was applied between the thermo- couple and the tube wall. Six Inconel-sheathed MgO-insulated Chromel-Alumel thermocouples are being tested in 1200 to 1250°F air. The observed drift in all units after 5000 hr of operation was less than *2°F, Ten MSRE-prototype wall-mounted thermocouples were exposed to fast temperature transients on the drain-tank bayonet-cooler testing facility. One thermocouple failed immediately. The remaining nine units have with- stood repeated raplid temperature changes. Results of tests of thermocouples installed on an MSRE-prototype freeze valve indicate that either 1/8-in.-OD duplex or 1/16-in.-0D single- conductor thermocouples will measure wall temperatures on the freeze valve within the required accuracy and will withstand the temperature transients inherent in this application. No significant difference in the perform- ance or durability of the units was noted. A six-circuit radiation-resistant thermocouple-disconnect assembly was fabricated and tested in a remote-maintenance facility. The discon- nect is simpler in construction than the assembly described previously and is more compatible with MSRE requirements. Part 2. Materials Studies 4., Metallurgy Tests were completed for determining the corrosive effect of CFa vapor on INOR-8 immersed in fuel salt at 1112 and 1292°F. Results of these indicate that CF4 vapor is effectively nonreactive toward INOR-8 at 1112°F but that minor attack may be promoted by CF; in the presence of fluoride salts at 1292°F. A final design and a welding procedure were established for the tube- to-tube sheet joints of the MSRE heat exchanger. The final design pro- vides for ultrasonic inspection. In order to demonstrate the welding and inspection procedures, a 96-joint test assembly was successfully fabri- cated and inspected by radiographic and ultrasonic Lamb-wave techniques. Brazing procedures for remotely preparing INOR-8 pipe joints were demon- strated by the fabrication of six joints. Only minor unbonded areas were found in these by ultrasonic inspection. L viii Additional mechanical properties data were accumulated for INOR-8, including thermal fatigue data at 1250 to 1600°F. Stress relieving was found to improve the elevated-temperature stress-rupture properties of weldments. Evaluation studies of the mechanical properties of heats of INOR-8 procured for MSRE construction were started. Grade TS-281 graphite, considered representative of MSRE material, was found to be more porous than experimentally made samples of similar graphite; however, it met design requirements. Specimens of TS-281 were permeated to 0.2% of the bulk volume in the standard permeation tests. Oxygen contamination was purged from R-0025-grade graphite with the de- composition products of NH4F.HF at temperatures as low as 392°F. Exposure of INOR-8 specimens to this environment at temperatures varying from 392 to 1300°F resulted in minimum attack at 752°F. Sintering studies were begun to develop procedures for fabricating Gdz03 and Gdz03-Alz0sz cylinders for MSRE control rods. The shrinkage characteristics and bulk density changes as a function of green density were determined for Gdz0s pellets that were fired at 1750°C in hydrogen. 5. In-Pile Tests Two molten-salt-fueled capsule assemblies, ORNL-MTR-LT-L and 47-5, were built to study the formation of CF4 in the.gas over fissioning MSRE fuel containing submerged graphite. In assembly 47-4, four capsules con- taining a graphite core submerged in fuel and two withvgraphite crucibles - containing fuel were irradiated. Assembly 47-4 was irradiated from March 15 to June 4, and it is now undergoing postirradiation examination at ORNL. The maximum measured tem- perature of the capsules was 1400 # 25°F during approximately 1500 hr of- full-power steady-state reactor operation.. Sixty. of the 121 recorded tem- perature changes of ~30°C or more included decreases to the solidus tem- perature of the fuel salt. Two capsules containing submerged graphite cores were installed in assenbly L47-5 to permit gas sampling during irradiation. Additional sealed capsules designed to provide a range of oxidation-reduction levels by altering the accessibility of chromium metal to the fuel were included. The range of the ratio of INOR-8 surface area to graphite surface area in contact with the fuel is O to 46:1., Assembly L47-5 is scheduled to be ir- radiated from September 17 to December 10, 1962. Postirradiation examination of assembly L47-4 revealed that elemental fluorine at pressures as high as 35 atmospheres was generated in some of the rapidly frozen salt used in tests of the effect of radiation on evo- lution of CF4 from fissioning fuel and graphite. The fluorine, undoubtedly of radiolytic origin, appears in the frozen salt after shutdown and is possibly the main causative agent of the CF4 in the capsules that contained large quantities of CF4. Since capsules of different configuration in the same lrradiated assembly showed widely divergent behavior, the undesirable products may be strongly dependent on some incidental effect, such as the ix freezing history of the capsule. Present indications are that the fluo- rine problem will not constitute a severe threat to the reactor operation. Tests are in progress (ORNL-MTR-47-5) to obtain direct confirmation of the evidence for a negligible concentration of fluorine in the molten fuel. 6. Chemistry A detailed examination of the U50°C contour in the LiF-BeFs-ZrF, ter- nary system suggests that 67 mole % LiF, 29 mole % BeFz, and 4 mole % ZrF4 should be a suitable solvent for UFs in a simplified fuel that contains no ThFy. A fuel based on this solvent, LiF-BeFp-ZrF.-UF, (66.85-29-4-0.15 mole %), has an estimated melting point of 445°C. This composition has a density of 2.15 g/cm® at the operating temperature. _ The eutectic 1n the system LiF-UF4 appears to be useful as a concen- trate (27 mole % UF4; mp, 527°C) for fuel makeup. The amount and nature - of segregation on freezing of MSRE-type fuels are tolerable. Experiments to demonstrate the reaction of CF, with MSRE systems gave positive indications, mainly at 800°C. Presumably, they would provide more conclusive results at higher temperatures than the 600 to 800°C range. Studies of the cleanup of graphite and of fuel continued and efforts to improve analytical aspects of MSRE fuel technology were directed toward oxygen determinations in fuel and in graphite and toward homogenization of radioactive samples of fuel. T. Fuel Processing Work on the detailed design of the MSRE fuel-processing system was started. The system provides for oxide removal from the salt and uranium recovery. A simplified flowsheet was prepared. 8. Engineering Research on Thermophysical Properties of Salt Mixtures The liquid-state enthalpy and the viscosity of a LizsC0a-NasCOz-KCOa (30-38-32 wt %) mixture proposed for use in out-of-pile studies relating to the MSRE were experimentally obtained. Unusual scatter in the data warranted only a linear fit of the enthalpy data. The derived heat ca- pacity over the temperature span 475 to T715°C is therefore given by the constant value 0.413 cal/g-°C. The kinematic viscosity was found to vary from 16.5 centistokes at 460°C to 3.1 centistokes at T15°C. An estimate of the mixture density as a function of the temperature allowed prediction of the absolute viscosity. THIS PAGE WAS INTENTIONALLY LEFT BLANK xi CONTENT'S ------------------ PART 1. MSRE DESIGN, ENGINEERING ANALYSIS, AND COMPONENT DEVELOPMENT 1. MERE DESIGN .....ciitiiieiniiierennnennnnns Design Status ....... . i, Reactor Procurement and Installation ...... Major Modifications to Building 7503 ... Construction QOutside Building 7503 ..... Procurement of Materials ............... Fabrication of Components .............. Reactor Instrumentation and Control Systems ------------------ ------------------ ------------------ ------------------ Layout of Instrumentation and Control System ............. Design Status ........ ... i, Procurement Status ........ i v 2. MSRE REACTOR ANALYSIS .....cvvivinnenennnn. Nuclear Accident Analyses .........ccieiuunn Calculational Procedures .........oeveee, ------------------ ------------------ ------------------ ------------------ Results of Fuel Pump Power Failure Analysis .............. Effects of Cold-Slug Accidents ......... Results for Filling Accidents .......... ------------------ Effect of Uncontrolled Withdrawal of Rods ................ Effect of Graphite Movement ............ Results of Concentrated Fuel Addition .. Reactor Statics Calculations .............. Criticality Calculations ............ ... Flux Distributions ............ ... Power Density .......... ... i, Steady-State Fuel and Graphite Temperatures Temperature Distributions .............. Average Temperatures .............c..... Reactivity Coefficients ................... Temperature Coefficients ............... Power Coefficient .......... ... ... ... ... Effective Delayed-Neutron Fractions ....... Reactor Kinetics Code Development ......... 3. COMPONENT DEVELOPMENT ...........c.oieuinine... Freeze-Flange Development ................. Freeze-Flange-Seal Test ................ Thermal-Cycle Test Loop ................ Freeze Test in Thermal-Cycle Test Loop . ------------------ oooooooooooooooooo ------------------ ------------------ ------------------ ------------------ ------------------ ------------------ ------------------ ------------------ ------------------ ------------------ ------------------ ------------------ Test of a Freeze Flange in the Prototype-Fuel-Pump Testing Facility ....... ..o, MSRE Control Rod ...ttt nnnennnan ------------------ ------------------ iii Heater Tests ....... Pipe Heaters .... ......................................... ----------------------------------------- Activation Analysis of Insulating Materials .............. Drain-Tank Coolers . ----------------------------------------- Sampling and Enriching System ....... .. it Detail Design ... ----------------------------------------- Mockup of Sampling and Enriching System .................. Core Development ... Model Operations ----------------------------------------- ----------------------------------------- Velocity Measurements .........ciiiiiiiireninrinninnnnas Measurement of Heat Transfer Coefficients ................ Experiments on the Behavior of Solids ........... ... ... MSRE Engineering Test Loop (ETL) . vviiiiiteerinennnnneenns Frozen-Salt-Sealed Graphite-Container Access Joint ....... Salt Composition HF Treatment .... ----------------------------------------- ----------------------------------------- Cold-Finger Collections in Cover Gas ...... . ..ot iiivnnnn ETT, Graphite Treatment ...ttt ittt tiontianeennnros MSRTE Maintenance ... ----------------------------------------- Placement and Removal of Freeze-Flange Clamps ............ Miscellaneous Flange-Servicing Tools ..........coiviiivenn Remote Viewing .. ----------------------------------------- Portable Maintenance Shield ........c.ii ittt inntnnenenns Brazed-Joint Development .. ...ttt i i Brazed-Jdoint Fabrication .....c. it ittt ineteineneeoonns Brazed-Joint Salt Y= v Mechanical-Joint Development ......... ... .. i, Pump Development ... ----------------------------------------- Fuel Pump Design and Fabrication .......... ... . Coolant Pump Design and Fabrication ............ ... v Design and Fabrication of Lubrication Stands for Pumps ... Lubrication-Fump IEndurance Test .............. ... v Drive Motor Deoign and Fabrication i..:i:iivisssisisssivessas Test Pump Fabrication ..........ciiiiiiiiiiiiiiiiennns Test Facility Construction ......... ... BT Hydraulic Performance Tests ........cc i, PKP Pump Hot Endurance Test ... i, Test Pump With One Molten-Salt-Lubricated Bearing ........ Instrument Development ......... ... ittt innnnanns Single-Point Liquid-Level Indicator ............ . ciiuie.n.. Pump-Bowl Liquid-Level Indication .............ciiiie..n. Single-Point Temperature Alarm System .................... Tenperaturc DeamIeT . ...ttt it it i i ettt i e e Thermocouple Development and Testing ........... ... . ..., PART 2. MATERIALS STUDIES METALLURGY ......... Corrosion Effects of ----------------------------------------- CF L o e s Welding and Brazing Studies ........iiiii ittt Heat Exchanger Fabrication ............ it nans Remote Brazing .. Welding of INOR-8 ----------------------------------------- ---------------------------------------- - 46 46 47 48 49 49 49 49 49 50 51 52 53 53 54 56 56 58 59 59 60 60 60 61 61 62 62 62 62 62 63 63 63 63 63 65 66 66 66 66 68 68 ran V% 87 87 88 88 92 92 xiii Mechanical Properties of INOR-8 ... .... . ciiiiiinnen.. Properties of Reactor-Quality INOR-8 ............... Thermal Fatigue of INOR-8 .......cciiiiiiiiinnnnnnn Graphite Studies ... ..t i i it i e i Evaluation of Grade TS-281 Graphite ................ Effects of Thermal Decomposition Products of NH,F-HF on Graphite-INOR-8 Systems ................ Fabrication of Gd;03 and Gds03—Al1,03 Pellets .......... IN-PILE TR S .. ittt et ettt et et e v s ssaansaenonnsensnns ORNL-MIR-47 Molten-Salt Irradiation Experiments ....... Experiment ORNL-MIR-47-4 ... ... iuiiiinnnnennennnnnn. Experiment ORNL-MIR-47-5 ... ... ..t iniiiniinnnan. Postirradiation Examination of Experimental Assembly ORNL-MIR =4 7=3 .. ittt ittt it tins e iaane e, Chemical Analyses of Irradiated Fuel for Corrosion Products . ... . i e Distribution of Radicactivity and Salt in Irradiated Graphite ... .. i e Postirradiation Examination of Experimental Assembly ORNL-MIR -4 74 i ittt et ettt et et s e Description of Capsule ..........ciiiiin e, Fabrication and Loading of Capsules ........... ... Fuel o e i i e e e Graphite ... . i i i i e e e e e Decay Bnergy ... e e e e Irradiation and Temperature Data ................... Production of Xenon and Krypton .................... GasS ANa Ly Se S vt ittt e i et e e e e Discussion of Results ........ ... ., Phase Equilibrium Studies .......c.cuiii ittt nnnnnens The System LiF-BeFo-ZrFs; . ... i iiiiiiaen.. The System LiF-BeFp-ZyF 4-ThF4-UFs ..o iivn... Li¥-UF, Eutectic as a Concentrated Solution of UF, for Fuel MaKkeup ..... vt ittt it nnnaeeennns Fractionation of LiF-BeF,-ZrF,-ThF,~-UF, on Freezing Crystal Chemistry ........ ... i, Oxide Behavior in Molten Fluorides — Dehydration of LiF Physical Properties of Molten Salts ................... Density of LiF-BeFo-2rF4-UFys ...ttt Estimation of Densities of Molten Fluorides ........ Graphite Compatibility ....... ... i . Removal of Oxide from Graphite Moderator Elements .. Behavior of CF, in Molten Fluorides ................ Analytical Chemistry . ... ittt ten it an Determination of Oxygen in MSRE Fuel Mixtures ...... Determination of Oxygen in Graphite ................ Homogenization of Radiocactive MSRE Fuel Samples .... oooooo ------ ...... oooooo ------ ------ ------ ------ ------ ------ ------ ------ ------ oooooo ------ ------ oooooo ------ 95 95 95 96 96 97 98 100 100 100 102 105 105 106 106 108 108 10% .109 110 110 110 112 115 117 117 117 118 119 121 121 122 123 123 123 125 125 128 130 130 131 131 FUEL PROCESSING Xiv --------------------------------------------- ENGINEERING RESEARCH ON THERMOPHYSICAL PROPERTIES OF SALT MIXTURES .« v otvvee ettt et e e e e e e e e e e e e Enthalpy Viscosity ---------------------------------------------------- --------------------------------------------------- 133 135 135 136 PART 1. MESRE DESIGN, ENGINEERING ANALYSIS, AND COMPONENT DEVELOPMENT ~ THIS PAGE WAS INTENTIONALLY LEFT BLANK 1. MSRE DESIGN Design Status Design work on the MSRE was essentially completed. There remains only final revising of system layout and design drawings and completion of the electrical design. The design report, in which all engineering calculations and analyses of the systems are presented, is presently be- ing compiled. Much of the design work remaining to be completed involves incorporation of development work into final design drawings. An example of such a design is the removable salt-pipe heaters and insulation. Final test evaluations of two different concepts of such removable insulated heater sections are nearing completion. Design drawings of these units must be completed after these tests indicate which type will be used in the system. Another item that awaits final test evaluation before the design can be completed is the control rod drive. A model incorporating all the principles has been operated successfully, and testing of an actual pro- totype unit will permit the design to be finished. Such a prototype 1is being ordered. All layouts for the three major areas, reactor cell, drain tank cell, and coolant cell, are complete, The variocus penetrations into these cells have also been designed. The layout of interconnecting gas, water, elec- trical, and instrument lines between these cells is nearing completion. Considerable effort was devoted to the design of the containment ves- sel, and all parts of this design were completed. Several methods of ef- fecting a closure with the seal membrane that welds to the steel shell were analyzed before an acceptable concept could be detailed. The pe- riphery of the containment vessel just below the closure flange was re- inforced with a circumferential steel ring to bring the stresses within limits at design-point pressure. This pressure was dictated by the maxi- mum credible accident. The vapor-condensing system, which ensures toler- able pressures within the cell in the event of the maximum credible acci- dent, was designed. Since the molten-salt system employs electrical heat to bring all components to the operating temperature, the electrical distribution for this heater network is rather elaborate. This electrical system is about 40% designed. The heater control centers are complete. Conduit and cable-tray layouts are 6U% complete. 'I'ne design ot moditications to existing auxiliary diesel power controls are 60% complete. All electri- cal designs are scheduled to be coumpleled Ly March 1, 19G3. Design drawings for the charcoal-adsorber system for the offgas are finished. Layout drawings of offgas piping in the reactor cell, drain tank cell, and special equipment room are 95% complete. - Very little remains to be done to bring the design of the entire gas system to completion. 4 The design of the jigs and fixtures for the assembly of the initial and replacement components for the test cell was completed. The compo- nents, piping, and flanges are supported by adjustable supports that du- plicate or resemble the test cell structure and are positioned precisely by optical tooling. Design work on the fixture for the assembly of the drain tanks and flush tank is in progress. Completion of the items mentioned above will complete the design of all major equipment and facilities, including the maintenance control room, the remote crane control, the television viewing devices, and the remotely operable lifting tongs. The design of the work shield, protec- tive environment enclosure (standpipe), tools, and operating procedure for replacing the 7/8 in, -0D graphite sampler is complete. The graphite sampler is identical in principal to the sampler illustrated previously,1 which was modified to accommodate the control rods and to raise the work shield 3 ft above the lower shield beams for working convenience. There were no changes in design concept or significant changes in detail of any component or system of the MSRE within this report pericd. A plan drawing of the major systems showing arrangement of vurlous com- ponents in the MSRE building is shown in Fig. 1.1. Reactor Procurement and Installation Major Modifications to Building 7503 The Kaminer Construction Corporation, which has the prime contract for the major modifications to Building 7503, has completed approximately 90% of its contract. The estimated completion date is October 15, 1962, as scheduled. The status of the work in progress is outlined in the fol- lowing sections, Reactor Cell, Stress relieving of the reactor cell vessel is com- plete, and installation of the sleeves between the inner and outer ves- sels is progressing. This work will be completed when the expansion joints (bellows) are received. The support beams for the lower roof plugs were concrete-filled and put in place. The frames for the lower roof plugs were fabricated and are being fitted to the cell. These plugs will be poured soon. The up- per roof plugs were poured but were not delivered because of lack of storage space. The seal membrane is on-site. Drain Tank Cell. All the concrete was poured in the drain tank cell, including the upper and lower roof plugs and the support shield beams. Plug welding of the stainless steel liner was completed, as shown in Fig. 1.2. The lower roof plugs should be installed and ready to weld in the seal membrane when it is delivered. 1"MSRP Prog. Rep. March 1 to Aug. 31, 1961," ORNL-3215, p 17, Figo lo 126 : 5 -~ — UNCLASSIFIED i3 EXI7 S % ORNL-LR-OWG. 74970 LUBE OIL L ! g INSTRUMENT 9 i BUILDING NORTH PANEL (o°REE) i | ‘—¢——4 l EUEL PUMP LUBE OIL { /.‘ LUBE oIt LUBE QIL INSTRUMENT . ! / POWER PANELS PANEL | F1REWA / B = | i N ‘ ‘ | cooLa Mo ‘ | i LUBE O 2 . ‘ PACKAGE\ Sl £ | , : e 7 £7 A 2 ‘ £LEV. 2400 £LEY 8400 o 5 , 3 : 7 | LRI E: h = § N -, o A EXIST. SPECTROMETER / 3 W3S FEACTH TUNNEL FLOOR £L.833-3" . EXIST. ACCESS / ! 3 HATCH. N | FiLey % )| ExisT oucT BasKSIM ew g E!fl:: pE!ZL 0 BE waLL o F/L?f»i/7 pucr RUN* & P : ! N R SLEEVES-R, IR NeyrroN. |, ) e FOR_INSTRUMENTATION \ SHELL INSTRIMENT /] Ar N 5 AUX, INSTRUMENTE) s sl 1 ELECTRICAL SERVICE 2 £ 5 AREA-FLOOR EL.84070) ’ REACTOR -~ QE§AEL A 30" BUTTERFLY — 7 o YALVES e 75 i | TRANSMITTER | | ALt & . e I ! (= 7 sAulE > 4800 b Y Joner MR e i L NEWIO" 2 SLEEVE 2 (BELO i L 5 7 b s P (p— = —3+—— | ——n " i 3 & / 4 ( ; o steEves Ry, 2 R A —} 3 A ConTAck FOR HEATER LEADS . MAINTIENANCE| |, i - L A S i e 30"EXHAUST Ducr PLAQTICE i | ey = A S i | K A N\ W STE d| ! eouipMENT al =~ IS A j f TREATMENT STORAGE i [ | fecmsaow ||| N / . i RAD/ATOR ELEV, 8310, /A joé Tl R & DUCTWORK | | JUNET, HATCH, | | K387 REYIRYD N EXCHINGE < e S s o A ¢ roR NEW 247D SLEEVE Re- LEAK DETECTION. CELL =3 s 1 00L i /4 \} DRAIN TAKK ety DRAIN TANK CELL S REACTOR 3 FLo 81420 i oo | = N XK= - === — 7 &9 2, 5 A 1 i =i £ ! i . ox 'r-eml. s q z 0 N ¥ DECONTAMINATION AUEL by S ¢ ?t\v‘"’@ L4l CE p srorace | | 14 /, 2 . ' ! FLOOR £1.832°C - // [~ %K 748 SPARES| Spuzi Mz “ r “ : S 1 { ? =S ) (R . N \HEW v \ Negacroz 7 f I N ‘)’ - N 8 SLEEVE Q) S 7 n | K, { . DRAIN ¥ DA FLUS, RN, =0 Mié;% e s = . TANK et | TANK K62 | Tk 7 M Bk 7 " - - - - b \, N Np? / \\‘; / % A N, L ! & SO - EXIS]. VENT By 2TACKE, £LocK 4 b, 7o o HOUSE / S S T o DRAIN LINE oe'% Y/ r,‘_‘; %) ; 3 TRANSFORMER 2N e = e = = 2 0 _____ ‘ e CONDENS, i - ! | PENrAOUSES | 3 73 / WALL (ABOVE) | s < a5 y i S o ST SR 2 SWITCH _HOUSE ; wa = i Wy 2oLET S e 1 0 % w0 U _FIN FLR ELR40°0" XPAN3, i s B ADJUSTABLE muih rr;;)y(e::wu Al F:Z 3 ?: ORY STACKED BLOC % E‘F’; | . SIORAGE TANK = - | hitvd| If, . COOLING : i 5 & t . / WATER I T /] PUMBPS > S AL FLow FAN-10000 CFM i - v i 7 ¢ A 7 ! // ) —1 =) N ELOWFAH | T == | # f 10,000 CEM E".fi’g CAHIHe ) Q Ex157 co"? _— > ¥ AXIVANE EANS N = < y / 5 il £L00R EL.B38O° | iy EX/ST BLOWER HOoUSE Fig. LI Plan View of MSRE. UNCLASSIFIED PHOTO 58774 Fig. 1.2. Fuel Drain Tank Cell. Maintenance Control Room. The grade beams for the malntenance con- Lrol room were poured. The [orms [or pourlng up Lo elevallon 862 'L are being installed, and the concrete will be poured soon. Radiator Cell. The sheet metal work in the radiator cell is approxi- mately 50% complete. It is expected that this work will be completed on schedule. Sheet-Metal Liner for Crane Bay. All framing for the crane-bay liner was fabricated and 40% of the framing was installed. The sheet metal liner was fabricated and is ready to install. Building Ventilation Duct Work. The duct work is 90% complete. The remaining work consists of the installation of dampers, which are sched- 1led to be delivered soon. Reactor Cell Exhaust Duct. The 30-in. exhaust duct is installed to the end of the Kaminer portion, including the butterfly valves. Hot Waste Storage Cell. The storage tank is installed and 95% of the piping complete. Hot Storage, Fuel Transfer, Decontamination, and Spent Fuel Storage Cells. The stainless steel liner is being installed in the hot storage cell. Work in the other three cells is complete, except for miscella- neous concrete patching and stacking of shielding blocks in the spent fuel storage cell. Electrical System. Electrical conduit installation in the crane bay is 90% complete. Wiring for the air compressors and controls is complete, except for some miscellaneous control wiring, which is awaiting delivery of control switches. Air and Water Piping. The compressed air and water piping is approxi- mately 98% complete. Auxiliary Line Penetrations into Reactor Cell, O0Oil, water, and other service lines were fabricated, and installation is approximately 50% com- plete, Electrical Service Area, All concrete was poured, and ladders and hatchway covers were installed. Work in this cell is 80% complete. Construction Outside Building 7503 The R. S. Hixson Construction Company of Oak Ridge was awarded the con- tract for the construction work outside Building 7503. Work was started ap- proximately May 1 and is expected to be completed approximately September 15. The exhaust filter house is complete except for the installation of filters, which are expected to be delivered approximately September 15. The offgas exhaust stack is installed. The exhaust blowers, duct, and dampers will be delivered by September 1 and installed by September 15. The cooling towers were installed, and the piping was completed, ex- cept for connections to the pumps. The pumps are to be delivered soon. The underground piping is complete. The inlet filter house is complete, and the steam coils, dampers, duct, and filters will be delivered soon. Procurement of Materials With the exception of the graphite for the MSRE core, materials pro- curement is approximately on schedule. The graphite is now promised for November 1, 1962, and efforts are being made to improve this delivery date. The status of procurement is outlined in the following sections. INOR-8. Delivery of all plate, rod, weld rod, and pipe is complete, except for some miscellaneous pieces. Partial delivery was made of pipe fittings and tubing; delivery is approximately 75% complete. Delivery of forgings, that is, freeze flanges, heat-exchanger tube-sheet blank flanges, O-ring gaskets, and other special forgings, is approximately 80% complecte, Graphite. The National Carbon Company has formed, baked, and graphi- tized all the base-stock graphite bars. The individual lots are at vari- ous stages of the multiple impregnation treatment. All bars received the first impregnation. There are about 80 graphite bars completely through all the required impregnations; however, the fabrication of sound base- stock graphite bars is the critical part of the fabrication. The delivery of the completely machined graphite bars was tentatively rescheduled from September 1 to November 1, 1962. Auxiliary Heat Exchangers. A contract was awarded Berlin Chapman Company for the fabrication of three stainless steel auxiliary heat ex- changers. Stainless Steel Piping. Contracts were awarded for the manufacture of type 304 stainless steel pipe, tubing, and fittings for all auxiliary piping systems. Most of this material has been manufactured and is be- ing inspected at the vendors' mills. Contracts have not been awarded t'or gspecial f'langes and gaskets. Reactor Cell Support Steel. Steel for the reactor cell equipment supporting structures is being furnished by O'Neal Steel Company. Special pipe spring supports are being manufactured by Bergen Pipe Support Company. Fabrication of Components The fabrication of MSRE components in the Y-12 Machine Shops is es- timated to be approximately 50% complete and on schedule. Delivery of the first component is expected approximately October 1, and all compo- nents are expected to be delivered by January l. The status of component fabrication is described in the following sections. Reactor Vessel. The reactor vessel (Figs. 1.3 and 1.4) is being fab- ricated and is approximately 45% complete. Delivery is expected approxi- mately January 1, 1963. The stainless steel thermal shield for the reactor vessel is being fabricated in the Union Carbide Nuclear Company Paducah machine shop and is approximately 35% complete. Delivery is expected approximately October 1, 1962. Procurement was initiated for approximately 75 tons of steel balls that will be poured into the thermal shield for gamma shield- ing. Ilecnt Exchanger. The INOR=C primary heat exchanger is approximately 65% complete (Fig. 1.5), and delivery is expected approximately January 1, 1963. A vendor, Wall Colmonoy, is under contract to furnace-braze the tubes to the tube sheet with a gold-nickel alloy. Radialor. The salt-to-alr radiator is approximately 12% complete. The headers for this component are being fabricated in the Union Carbide Nuclear Company shops at X-10 and Paducah and are approximately 95% com- plete. Delivery is expected approximately January 1, 1963. UNCLASSIFIED PHOTO 38994 Fig. 1.3. 1Inside of Reactor Vessel Showing Flow Straightening Vanes. Radiator Enclosure. The enclosure for the radiator (Fig. 1.6) is approximately 65% complete, and delivery is expected approximately January 1, 1963. The door-lifting mechanism was fabricated in a local Oak Ridge Machine Shop and is complete. Delivery is complete on all motors, clutches, brakes, and electric heaters for this component. Salt Storage Tanks (Total of 4). The several INOR-8 fuel and cool- ant storage tanks (Fig. 1.7) are approximately 45% complete. Delivery of these components will begin approximately October 1, 1962, and be com- Pleted by October 1, 1963. Fuel and Coolant Pumps. Fabrication of the fuel and coolant pumps (Fig. 1.8) is approximately 75% complete. Delivery of the first unit is expected approximately January 1, 1963. Miscellaneous Fabrication. The ORNL and Paducah machine shops are fabricating miscellaneous parts for the major reactor components. In addition, certain equipment supports, assembly jigs, and panel boards are being fabricated in the K-25 shops. The sampler enricher, charcoal beds, 10 UNCLASSIFIED PHOTO 38932 TR A Fig. 1.4. Flow Distribution Nozzle [or Reactor Vessel. UNCL ASSIFIED PHOTO 38530 Fig. 1.5. Some Completed Components of INOR-8 Primary Heat Ex- changer,. i UNCLASSIFIED PHOTO 38995 Fig. 1.6. Partially Completed Radiator Enclosure. 12 UNCLASSIFIED PHOTO 38680 () “0‘ ®, 0’. LX) .0 Fig. 1.7. Coolant Salt Drain Tank. 13 UMCLASSIFIED PHOTO 38989 Fig. 1.8. Volute and Upper Half of Fuel Pump Bowl. 14 salt piping, and other miscellaneous parts are being fabricated in the ORNL machine shop. A contract was awarded to Taylor Engineering Company for the fabrication of seven special equipment supports and four special vessels. Reactor Instrumentation and Control Systems Layout of Instrumentation and Control System The layout of the instrumentation and controls system proceeded closely in accordance with the scheme outlined in the previous report.2 Drawings showing the exact location of all control panels were completed, including drilling details for floor penetrations in the main and aux- iliary control areas. The relay cabinet was moved to a new position, and three panels for control of the diesel generators were added to the auxiliary-board panel arrangement. Requirements for wire ways between the main control area and field-mounted equipment were determined, and detail design is now under way. 'These wire ways consist of three sepa- rate sets of 3- by 24-in. open trays. Thermocouples, signal cables, and control circuit wiring will be run in separate trays. Cable from isolated signal transmitters will be run in conduit to the nearest tray. Details for wire terminal boxes between field-mounted electrical equipment were also determined. Design Status Instrument Application Diagrams. Instrument application flow dia- grams for the fuel sampler-enricher system, the fluorination system, and the containment system were completed and approved for construction. Nine diagrams previously approved were revised in accordance with recent de- sign changes. A tabulation of all instruments shown on the flow diagrams, giving identifying numbers, location, function in process, and a brief description, was completed. These will be revised periodically to in- corporate design changes. Control Panels. The design of 8 of the 12 modular instrument panel sections required for the main board was completed and fabrication of these panels is presently under way in the ORNL panel shops. The three remaining panels, as well as the control console, will be completed when all the nuclear instrument requirements are known. A complete design for one panel section includes instrument-mounting-hole cutout drawings, pneumatic tubing diagrams, and electrical wiring diagrams. 0 20 40 60 . 80 100 120 TIME (sec) Fig. 2.1. Power and Temperatures Following Fuel Pump Failure, with No Corrective Action. Ls 19 control rod action and closing of the radiator doors following the fuel pump power failure are given in Fig. 2.2; the control rods were considered to decrease reactivity at the rate of 0.07% per second 1 sec after power failure, and heat removal from the radiator was stopped after about 30 sec. For these circumstances the fuel salt temperature rise was small, and the secondary coolant temperature did not become too low. Effects of Cold-Slug Accidents The consequences of cold-slug accidents were estimated by calcu- lating the reactivity transient associated with passage of a cold fuel slug through the core and introducing this transient into a Murgatroyd calculation by considering a series of reactivity ramps. The temperature rise calculated by Murgatroyd for a constant inlet temperature was then superimposed on the temperatures that would be produced by the cold slug without nuclear heating. Results are shown in Fig. 2.3 for two extreme UNCLASSIFIED ORNL-LR-DWG 67579 1300 CORE OUTLET . _— GRAPHITE MEAN u — ] I 0 1 — ° 200 CORE FUEL MEAN (VS g CORE INLET '_ 1200 GRAPHITE A / g \ __'_--_,__ / ./_--_——_—_‘___--_- = N 7 i ac \ /"——"—-__/ w a \ _ /a = 1100 e \ / — 20 ft3 COLD SLUG EL —==30 f PRESSURE RISE {psi) o \ 300 \ POWER ZOuU : \ ' 100 \Q 3 POWEFR (Mw) TIME (sec) Fig. 2.3. Bystem Behavior Associated with PaSSagé of Cold Slugs at 900°F and a 1200-gpm Flow Rate with the Core Initially at 1200°F. cases: 20 or 30 ££t3 of fuel at 900°F was assumed to be pumped at 1200 gpm through the core; the reactor was initially critical at 1200°F and at low power with no circulation prior to introduction of the cold slug. Peak temperatures in the core were estimated to reach about 1600°F. The core pressure rise was not appreciable in either case. 21 Results for Filling Accidents Criticality could be reached prematurely while the core was being filled if the fuel were abnormally concentrated, or if the fuel tempera- ture were abnormally low, or if the control rods were withdrawn too far. The power excursions that would result from the continued addition of re- activity associated with such filling accidents were computed by Murgatroyd, assuming no heat removal from the fuel and using a fuel-graphite heat trans- fer coefficient appropriate for a static core. In the first case it was postulated that the fuel was concentrated by selective freezing in the drain tanks of phases containing no uranium (or thorium). With 39% of the fuel frozen,4 filling the core with the remain- ing liquid at 1200°F at a rate of 1 ft3/min would result in a 55-Mw power excursion about 60 sec after initial criticality. No significant pressure surge would result. Filling the core with salt having the normal fuel concentration and a temperature of 1200°F would produce an excursion similar to the one de- scribed above if all rods were fully withdrawn. Filling at 1000°F with the fuel concentration and rod positions corresponding to criticality at 1200°F would produce a much smaller excursion. Rod insertion would be quite effective in preventing the power ex- | cursions in these accidents because the rates of reactivity increase would be very low. If no corrective action were taken (e.g., insertion of rods or draining of core), temperatures in the core would exceed 1800°F, in the worst case, by the time the core became completely full. Effect of Uncontrolled Withdrawal of Rods Simultaneous withdrawal of all three rods could produce a large power excursion and undesirably high temperatures if these rods were withdrawn rapidly with the reactor initially subcritical or at very low power. In the worst case examined, the rate of reactivity addition was assumed to be 0.08% 6k/k per second and the initial power was taken to be 1 watt. A pover excursion to 730 Mw and a pressure surge of 19 psil resuited. More serious, the calculated mean fuel temperature rose from the initial 1200°F to a high of 1600°F, and local temperatures in excess of 2000°F were esti- mated. . Effect of Graphite Movement If a graphite stringer were to break in two near the center of the core and the upper half could float up, the reactivity would increase 0.004% 6k/k for each inch of stringer replaced by fuel. If the entire central stringer were replaced with fuel, the reactivity increase would be only 0.13% 6k/k, No power or temperature excursion of any consequence would result, even in the latter event. “This is the highest fraction that could be frozen and still leave enough liquid to fill the core. 22 Results of Concentrated Fuel Addition An upper limit on the disturbances which could result from a fuel addition of 120 g of U®35 yas calculated. For this case it was postulated that the added fuel passed through the core as a flat pancake-shaped ele- ment under normal flow-rate conditions. If this were possible and the reactor were initially at 10 Mw, the power would rise to a peak of 120 Mw, and the fuel mean temperature would rise 90°F. The pressure surge would be about 2 psi. Smaller disturbances would result at lower initial powers. Reactor Statics Calculations The reactor statics calculations reported here are based on the cur- rent reactor design. In this connection, a two-dimensional, 19-region nuclear model of the reactor was utilized to represent the various reactor regions. Because of the limitation to two dimensions, the control rod thimbles were considered as a single INOR-8 cylindrical annulus concentric with the core center line. A vertical half-section of the reactor model showing the relative sizes and positions of the various regions is shown in Fig. 2.4. The boundaries of the regionc and the volume fractions of fuel, graphite, and INOR-8 in each are summarized in Table 2.1. A fuel salt mixture containing x mole % UF4, 70 mole % LiF, (25 - x) mole % BeFs, and 5 mole 9% 7rF4 was considered in the calculations. The isotopic compositions of lithium and uranium were based on material com- positions that might reasonably be available for use in the reactor. The Table 2.1. Nineteen-Region Core Model Used in Equipoise Calculations for MSRE (See Fig. 2.4 for graphical location of rcgions) Rodiun (in.) Height (in.) Composition (val &) , Region Region Represented Inner Outer Bottom Top Fuel, (raphite INOR-8 A 0. 29.56 Th.92 T76.04 0 0 100 Vessel top B 29.00 29.56 -9.1h4 Th .92 0 o) 100 Vessel sides C 0 29.56 -10.26 -9.1k 0 0 100 Vessel bottom D 3.00 29.00 67 b7 T4 .92 100 ) 0 Upper head E 3.00 28.00 66.22 67 .47 93.7 3.5 2.8 F 28.00 29.00 0 67 .47 100 0 0 Downcomer G 3.00 28.00 65.53 66.22 94 .6 5.4 0 H 3.00 27.75 64.59 65.53 63.3 36.5 0.2 I 27,75 28.00 N A5 53 0 0 100 Core can J 3.00 27.75 5.50 64,59 22.5 7.5 0 Core K 2.94 3.00 5.50 T4 .92 O 0 100 Similated thimbles L 0 2.94 2.00 64,59 25.6 T4 .L 0 Central region M 2.94 27.75 2.00 5.50 22.5 TT+5 ) Core N 0 27.75 0 2.00 23.7 76.3 o) Horizontal stringers 0 0 29.00 -1.k1 0 £A.Q 15.3 1.7.8 P 0 29.00 -9.1hk -1.k1 90.8 0 9.2 Bottom head Q 0 2.94 66.22 T4 .92 100 0 0 R 0 2.94 65.53 66.22 89.9 10.1 0 S 0 2.6h 64.59 65.53 43.8 56 .2 o 23 UNCLASSIFIED ORNL-LR-DWG 736140 Y I P PO B | 0O 2 4 6 8 10 INCHES Fig. 2.4. Nineteen-Region Core Model for Equipoise-3A Calculations. See Table 2.1 for explanation of letters. calculations described below were for an isothermal system at 1200°F with no reactivity controlled by the control rods, no fission-product poisons, and no chemical impurities in the fuel. Criticality Calculations The initial criticality calculations were made with Modric, a 34-group, multiregion, one-~dimensional, neutron-diffusion program for the IBM-7090. Because this program is limited to one dimension, two types of geometry calculations were made. One was based on slab geometry, with region and material specifications corresponding to an axial traverse through the main part of the core; a radial buckling term was used in this calculation. The other was a radial calculation, with region and material specifications corresponding to a radial traverse through the reactor midplane; an axial buckling term was used in this calculation. The two calculations agreed within 1% in predicting a critical uranium concentration of 0.1% mole %. In addition, these calculations provided the two-group nuclear constants for each of the regions as required for two-dimensional, two-group calcu- lations. Flux Distributions The spatial distributions of the fast and thermal fluxes and adjoint fluxes were calculated for the 19-region, 2-dimensional model using the Equipoise-3A program. The results for a reactor pover of 10 Mw are pre- sented in Figs. 2.5 and 2.6. The radial distributions (Fig. 2.5) are for an axial position that corresponds to the maximum in the thermal flux, which is at a position very close to the core midpiane. The axial dis~ tributions® (Fig. 2.6) are given for a position 7 in. from the core center line; this radius corresponds to the maximum value of the thermal flux. Information about the space-energy distribution of the fast flux was obtained by combining the Equipoise-3A results with the results of the two Modric calculations. Figure 2.7 shows the radial distribution, near the core midplane, of the neutron fluxes for energies greater than 0.1 and 1.0 Mev, and Fig. 2.8 shows the axial distribution of the same fluxes 3 in. from the core center line. This radial position corresponds to that of the control-rod thimbles and the small grapnlte test specimeris. The Tluxes are normalized to a power level of 10 Mw. Power Density 'he spatial distribution of the power density is determined by the total fission rate. About 87% of the total fissions in the MSRE are caused by thermal neutrons. As a result, the spatial distribution of the fuel power density is essentially the same as that of the thermal-neutron flux. The maximum fuel power density at 10 Mw with no fuel permeation of the graphite is 33.6 w/cm®. The corresponding maximum graphite power density is 0.83 w/cm®. ' SThe reference plane for axial measurements was the bottom of the horizontal graphite bars on which the vertical stringers are supported. ~y UNCLASSIFIED ORNL-LR-DWG 73611 (x10'3) ,4 ~C UNCL ASSIFIED ORNL-LR-DWG 73612 FAST FLUX | (x10'5) I o Q {.') I y _ P : ~ FAST FLUX £ \ 3 g Co \ 3 S 4 = P N | f‘* / N\ 'E’: 8 -~ SLOW FLUX N E 1o N e : o =~ ™~ 2 SLOW FLUX e SLOW ADJOIN'I\\ L. 8 o iR ) T — 5 TN 2 Bt W 7 SLOW ADVOINT S ] T e 3 d "\ o« S, ':33. 6 N = FAST ADJOINT S« > —_feeees 0.1 Mev) N - 5 3 2 25 ' G (£ > 0.1 Mev! N 3.0 e . E z > = c o g S 2 as r 2.0 A z ] . = 2 [+ 4 lCL) 2.0 \ /\ = g v S s z g \ & 0.4 . - . 4 14 t ol P LE > Mev) Q 3 & $lE>iMew) w 1.0 A\ o / \ 2 / x 1.2 / Vs ™, 0.5 4 \ 0.5 // \ 5 | . ,//// 0 5 10 5 20 25 30 -0 0 1c . 20 30 40 S0 80 70 80 RADIUS (in.) : AXIAL POSITION (in.) Fig. 2.7. Radial Prcfiles of High-Energy Fig. 2.3. High-Energy Neutron Fluxes 3 in. Neutron Fluxes in Plane of Maximum Flux Based on from Core Center Line Based on Modric Calculations. Modric Calculations. : - ¢ 1 k!l o¢ 27 - Steady-State Fuel and Graphite Temperatures Fluxes and power densities obtained from Equipoise calculations were used to estimate fuel and graphite temperature distributions and average temperatures for steady-state reactor operation at 10 Mw.® Temperature Distributions The shape of the temperature distribution in the reactor is deter- mined by the shape of the power-density distribution and the fuel flow . pattern. In the main part of the core, where the fuel flows in discrete channels among the graphite stringers, local temperature variations across individual fuel channels and graphite stringers are superimposed on the gross distributions. The gross distributions, which have the greatest effect on the reactor average values, are described below. They are based on 10-Mw reactor operation and a total flow rate of 1200 gpm. In describing the gross temperature distributions, only that part of the core which generates the major fraction of nuclear energy needs to be considered, that is, the central core associated with the regions designated J, L, M, and N in Fig. 2.4t and Table 2.1. The fuel temperature in any chan- nel at a particular elevation is a function of the radial position of the channel, the inlet fuel temperature, the flow rate in the channel, and the total amount of heat produced in or transferred to the fuel as it passes from the inlet to the elevation in question. Figure 2.9 shows the axial variation of fuel temperature in the hottest channel for the power dis- tribution calculated for the MSRE. The channel inlet temperature is some- what higher than the reactor inlet temperature because of the heat produced in the entrance regions through which the fuel must pass in reaching the main part of the core. Figure 2.10 shows the radial variation in fuel temperature at the midplane of the core. The relatively lower temperatures in the central and outer regions of the core are caused by the higher fuel velocities in these regions. The fuel velocity in the central channels is expected to be about three times that in most of the remaining channels. Less extreme variations are expected in the outer channels. The over-all temperature distribution in the graphite depends on the fuel temperature, the graphite power density, and the resistance to trans- fer of heat produced in the graphite to the fuel in the adjacent channels. The internal resistance of the graphite to heat transfer produces some elevation of the graphite temperature above that of the fuel, but most of the temperature difference results from the Poppendiek effect.” Figure 2.9 shows the axial distribution of the mean temperature in a graphite stringer immediately adjacent to the hottest fuel channels. The calculated axial 8J. R. Engel and P. N. Haubenreich, "Temperatures in the MSRE Core During Steady-State Power Operation,' ORNL-TM-378 (in preparation). 7H. F. Poppendiek and L. D. Palmer, "Forced Convection Heat Transfer in Pipes with Volume Heat Sources Within the Fluids," ORNL-1395-C, Nov. 11, 1953 UNCLASSIFIED CRNL—LR— JWG 73615 1300 UNCLASSIFIED o ORNL-LR-DWG 73616 /’_\ 1290 / \ / 1280 / : 1280 GRAPHITE 1270 GRAPHITE 1260 / ‘ - 1260 E N / / | \ | . 4250 1240 / yd , 1240 \ 1220 - / 1230 \ ' 1220 _ - \ y 1200 / ,/ | \JEL \ / - 1210 i \ / | 1200 i ;‘ 100 ™ | : \ \ 190 by \ 5 TEMPERATURE (°F) TEMPERATURE (°F) 1160 1180 0 10 20 30 40 52 60 70 0 5 10 15 20 2 30 DISTANCE FROM BOTTOM OF CORE fin.) RADIUS (in) Fig. 2.9. Axiel Temperature Profiles in Hottest Fig. 2.10. Radial Temperature Channel of MSRE Core (7 in. from Core Center Line). Profiles in MSRE Core Near Midplane. 8¢ 29 distribution is somewhat steeper than may be expected in the MSRE because axial heat transfer in the graphite was neglected in the calculation. Figure 2.10 shows the radial variation of the graphite temperature at the core midplane. The distribution in the plane of maximum graphite tempera- tufes would have the same general shape but would be displaced upward about 12°F, The graphite temperature distributions in Figs. 2.9 and 2.10 were calculated on the basis that 6% of the reactor power is produced in the graphite by gamma and neutron heating; no other heat sources in the graph- ite were considered. The graphite-fuel temperature differences will be greater than the indicated values if fuel salt soaks into the graphite and produces a source of fission heat in the stringers. If 2% of the graphite volume were uniformly filled with fuel salt, the maximum difference between the mean temperature in a graphite stringer and that in the adjacent fuel channel would increase from 62.5 to T4.5°F. In addition to the above results for the main core regions, the mean fuel temperatures and temperature rise were also calculated for a number of peripheral regions, based on 10-Mw reactor operation and an average fluid temperature rise of 50°F in passing through the core. The increase in fuel temperature as the fuel passes through a region and the associated mean temperatures for the various regions are summarized in 'lable 2.2. Average Temperaturcs A useful concept in relating the actual fuel and graphite temperature distributions to the total reactivity is the nuclear average temperature. This quantity is defined as follows: At low power, reactor criticality is assumed to be established at isothermal temperature conditions in fuel and graphite. Then, as the power is raised, the fuel nuclear average tempera- ture, T,, is defined as the equivalent uniform fuel temperature that gives the sam® total reactivily level as that associated with the actual fuel temperature distribution, with the graphite temperaturg held constant. Similarly, the graphite nuclear average temperature, T , is defined as the uniform graphite temperature that gives the same react%vity level as that given by the actual graphite temperature profile, with the fuel temperature held constant. A reactor nuclear average temperature could be defined in an anglogous manner; however, it appears more convenient to consider Tf and T individually in the manner indicated below. Using first-order, two-group perturbation theory, the relation de- rived between the nuclear average temperatures and the actual distributions was8 8B. E. Prince and J. R. Engel, "Temperature and Reactivity Coefficient Averaging in the MSRE," ORNL-TM-379 (in preparation). where J Reactor 30 T(r,z) G(r,z) dv G(r,z) av Reactor * »* e 3 >* G(r,z) = G 8.8 + G B8 +G,P.0 + 008, - Table 2.2. Fuel Temperature Rise and Average Temperature Asdoelated with Varivus Reaclor Regions at a Power Level of 10 Mw Average Temperature Rise in Passing Average Temperature Region” Through Region in Region (°F) (°F) D 2.4 1224 .7 B 0.8 1223.1 F 1.3 1175.6 G 0.5 12224 H 0.6 1221.9 J 43.0 (b) L 22.0 (b) M 1.0 (v) N 0.3 (v) 0 0.3 1177.1 P 0.7 1176.6 4] 1.7 1200.9 R 0.3 1199.9 S 0.2 1199.7 aSee Table 2.1 and Fig. 2.4 for location and com- rosition of regions; only regions which contain fuel are considered here. bActual temperature distribution was computed in this region; see Figs. 2.9 and 2.10. (1) 31 In Eq. (1), T is either Tf* or T*, and the integration takes place_over the reactor volume. In thé defin¥tion of G(r,z), B, o, ¢i, and @, are the fast and slow fluxes, and the fast and slow adjoint fluxes, respective- ly; the G; 4 are constant coefficients for each region of the reactor in which the composition is uniform and are related to the macroscopic two- group nuclear constants. The values for ¢ and ¢* are the unperturbed fluxes and adjoint fluxes calculated at isothermal conditions in the reactor. In Eq. (1), the tem- perature derivations associated with the function G are taken with respect to Ty in obtaining the fuel-temperature-weighting function, and with re- spect to Tg in obtaining the graphite function. The temperature-weighting functions were evaluated from the MSRE, using the Equipoise two-group fluxes, adjoint fluxes, and nuclear param- eters. The results are shown in Figs. 2.11, 2.12, and 2.13. These corre- spond to radial and axial traverses, which pass through the thermal flux maximum in region J of Fig. 2.4. Discontinuities in the weighting func- tions occur as the effective fuel volume fraction changes from region to region, and thus reflect the change in local neutron multiplication, as well as the variation in nuclear importance from region to region. UNCLASSIFIED ORNL-LR-DWG 73617 0.9 | S 0.8 N\ \ FUEL AN 0.7 N GRAPHITE\ \ 0.6 \ 0.4 \ %\ 0.3 \ c.2 \ RELATIVE NUCLEAR IMPOR™ANCE - o 5 10 15 20 25 30 RADIUS (in.) Fig. 2.11. Relative Nuclear Importance of Fuel and Graphite Tem- perature Changes as a Function of Radial Position in Plane of Maximum Thermal Flux. 32 UNCLASSIFIED ORNL-LR-DWG 73618 1.0 [ I | I N EQUIPOISE 3A RESULT / \ 08 sin2 ["/79.4 (2 + a1 +) / Y8 os : / \ I =z . Il b | E I 2 i = o4 By - / \ 3 < / \ | 1] d 0.2 \ l ” \ 2 / W = \ \ Lé' 0 Jl - \"--._ :j ‘ | je———— MAIN PORTION OF CORE - (i} | L ) I -o'2 | -0.4 -06 : -20 -10 0 10 20 30 40 50 60 70 80 z, AXIAL POSITION (in.) Fig. 2.12. Relative Nuclear Importance of Fuel Temperature Changes as a Function of Position on an Axis Located 7 in. from Core Center Line.: UNCLASSIFIED ORNL-LR-DWG 73619 1.0 I T I Py —— EQUIPOISE 3A RESUL/. \ na L — —5in2 ["/79.4 (z+ 4.71} ] \ 8 | / \ s Z 06 |_ x (o] a Z 54 . N T \ 3 /| \ > 02 N\ w \ v |-—MAIN PORTION OF CORE m-mmrmormmeserm o m —0.2 —04 -10 0 10 20 30 40 50 60 70 80 2,AXIAL POSITION (in.) Fig. 2.13. Relative Nuclear Importance of Graphite Temperature Changes as a Function of Position on an Axis Located 7 in. from Core Center Line. 33 Use was made of these temperature-weighting functions to compute nuclear averages of Tf and T¥ for the conditions of a 10-Mw power level, with reactor inlet and outle% temperatures of 1175 and 1225°F, respectively. The results obtained are given in Table 2.3, considering three levels of fuel permeation of the graphite. Both the bulk average and the nuclear average temperatures are given. Table 2.3. Average Temperatures in Reactor Vessel at 10 Mw as a FMunction of Fuel Permeation of Graphite Fuel Temperatures Graphite Temperatures Fuel Permeation » (°F) (°F) (% of graphite volume ) Nuclear Bulk Nuclear Bulk Average Average Average Avcrage 0 1213.1 1198.0 1256.6 1226.3 0.5 1258.4 1227 .4 2.0 1263.7 1230.7 Since axial heat transfer in the graphite was neglected, any heat produced in the graphite was considered transferred to the fuel immedi- ately adjacent to it. As a result, the gross fuel temperature distri- bution was unaffected by fuel permeation. It was assumed that the shapes of the temperature-weighting functions were also unaffected so that the different degrees of fuel permeation did not produce any change in the fuel nuclear average temperature. The changes in the graphite nuclear average temperature resulted from changes in the temperature level of the graphite. Reactivity Coefficients Temperature Coefficients Consistent with the definition of the nuclear average temperatures, importance-averaged temperature coefficients of reactivity for graphite and fuel were calculated. The equations for the coefficients were based on first-order perturbation theory and involved the function G(r,z) given previously. Results of calculations for fuel containing no thorium, based on per- turbation theory are compared below with those based on a one-region homogeneous-reactor model employed in previous studies.® The temperature coefficients of reactivity, in units of (6k/k)/°F, were: S'MSRP Prog. Rep. March 1 to August 31, 1961," ORNL-3215, p. 83. 34 For Fuel For Graphite Perturbation theory | -4 .46 x 107° -7.27 x 107° Homogeneous reactor model -4.13 x 10*S -6.92 x 10-5 The differences in the results from the two methods are small and well within the probable error associated with the value of the temperature coefficient of expansion of the salt. The coefficients from the homogeneous model do depend strongly, however, on the correct choice of effective re- actor size. These calculations also account for the effect of the salt temperature on fast-neutron leakage; taking this into account has the net result of increasing the fuel coefficient slightly relative to the graphite coefficient. TPowel Cocfficient The power coefficient of reactivity is defined here as the amount of reactivity reswlting from nuclear heating that must be compensated by con- trol rod motion per unit increase in power. 'This may be calculated from the relations between nuclear average temperatures, fuel inlet and outlet temperatures, and reactivity coefficients. At a 10-Mw thermal power, these relations are T = ° o Tp; + 50°F (2) T = T, - 11.9°F (3) f = “fo ? 1 = T+ 31.6°F | (4) . g fo ? -F ¥ ‘ - * . Ak = w(4.45x 10 5) AT, = (7.27 x 10 5) aTg , (5) where Ak = reactivity change associated with changes in nuclear average temperatures, Tfi = fuel inlet temperature, Tfo =- fuel outlet temperature, i Tf = fuel nuclear average temperature, ¥ . Tg = graphite nuclear average temperature. In obtaining power coefficients of reactivity, three cases were con- sidered: (1) the fuel outlet temperature was held constant as the power was raised to 10 Mw; (2) the mean of the inlet and outlet temperature, T,., was held constant; and (3) the outlet and inlet temperatures were lowed to vary to maintain criticality with no control rod motion. The 35 initial temperature at zero power was assumed to be 1200°F; the reactivity change associated with increasing the power to 10 Mw was calculated for the three cases considered. The results are presented in Table 2.k4. Table 2.4. Temperatures at 10 Mw and Associated Power Coefficients of Reactivity m ¥ * lIIfo Tfi Tf Tf Tg POYe? Condition : Coefficient (°r) (°F) (°F) (°F) (°F) [(% ok/ik)/Mw] Toy = constant 1200 1150 1175 1188.1 1231.6 -0.018 T’f = constant 1225 1175 1200 1213.1 1256.6 -0.0LT Teys Tpy varied 118%.9 1134.9 1159.9 1173.0 1216.5 Effective Delayed-Neutron Fractions Fuel circulating at 1200 gpm spends 9.4 sec in the core and 16.4 sec in the remainder of the loop. Conseguently, a large fraction of the longer lived delayed neutrons will be emitted outside the core and effectively lost. Furthermore, the distortion of the source distributions that will result from the transport of the precursors will increase the leakage probability for the neutrons emitted in the core. The reduction in the contribution of B;, the delayed neutrons of group i, to the chain reaction would have a strong effect on the reactor kinetic behavior. The computer programs that have been developed for MSRE kinetics calculations (Murgatroyd and Zorch) do not treat the delayed neutrons rigorously. Instead, they use the neutron and precursor equations for a noncirculating system, with values of B adjusted to allow for losses resulting from circulation. Thus use of these programs requires previous evaluation of effective B values. Some calculations were made to determine the effect of the energy and spatial distributions of the delayed neutrons on the effective values of B: in steady-state operations.lo For each group, the spatial distribution of the delayed-neutron source was calculated from the precursor production distribution (which was assumed to follow the fundamental mode of the neu- tron flux), taking into account the transport of the precursors prior to their decay. The core was represented as a single region, with uniform fuel volume fraction and velocity equal to the averages for the actual MSRE core. Figures 2.14 and 2.15 show calculated source distributions along a channel through the core. (The densities are per unit volume of fuel salt, normalized to one fission neutron, and apply to a channel where the flux 1%p | . Haubenreich, "Effective Delayed Neutron Fractions in MSRE, " ORNL-TM~380 (in preparatlon) (x107%) 2.8 T 2.4 E = >- = W z 2.0 w o w (&) @ 3 - 3 {.i3 = o (144 }_ = w {.2 = [&] w > a m L 0.8 V3] 2 = < o 0.4 w . 0 Fig. 2.14. Axial Distribution of Delayed- Neutron Source Densities in an MSRE Fuel Channel. "JNCLASSIFIED ORNL-LR-DOWG 7:362) I GROUP 74, (sec) 4 56 2 4 2.3 5 6 (SEE FIG 2.i5 FOR GROJP 4 DISTRIBUTION) . S !, //' R v _ / 6 7 / 0.4 0.6 0.8 1.0 RELATIVE HEIGHT Fuel circulating at 1200 gpm. UNCLASSIFIED ORNL-LR-DWG 72621 (10793 6 2 2 5 / \ > / \ g) /'S"ATIC \ Wl S 4 / A ¢ / 7\ 3 / / \ & 3 / ‘ \ 3 / /. \ uzJ / / CIR(?ULATING \ o 2 / \ 3 [/ \ 2 |/ / \ [a] ¢ LT \ = | A \ 2|/ \ 0 0 0.2 0.49 - 0.6 0.8 1.0 RELATIVE HEIGHT Fig. 2.15. Effect of Fuel Circulation on Axial Distribution of Source Density of Group 4 Delayed Neutrons. Fuel circulating at 1200 gpn. 9¢ 37 equals the radial average.) The ratio of the nuclear importance of a de- layed neutron to the importance of a prompt neutron was taken to be the ratio of the appropriate nonleakage probabilities. The nonleakage proba- bility for neutrons of a particular group was calculated by representing the source distribution by the series S,2) = ) ) Sigy JoUp/R) sin (am/m) (6) m=]l n=l and assigning to the m,n term a nonleakage probability given by P, = < (7) and £ (2F (1) where S.(r,z) is the source spacial distribution for delayed neutrons of ith group, S;,, orc cxponsion coefficients, j, is the mth root of the zero order Bessel function, Jg5, R is the radius of the reactor, H is the height of the reactor, L is the thermal diffusion length, and T. is the Fermi age for delayed neutrons of the ith group. The age, Tio wds reduced from that for prompt neutrons to allow for the lower initial energy of the delayed neutrons. (When the fuel is not circulating, this effect enhances the importance of the delayed neutrons by reducing the nonleakage probability.) Results of the calculations are summarized in Table 2.5. Reactor Kinetics Code Development Preliminary results were obtained from a new reactor kinetics program, Zorch, in which approximate account is taken of the space dependence of fuel and graphite temperatures.l! This is an improved version of the Murgatroyd program used in previous studies. In the Zorch calculation, the reactivity was related more realistically to the temperature profiles by using the concept of nuclear average temperatures. During an excursion, the axial pover-density shape was assumed to be that of the fundamental mode, that is, sin nz/H. The axial temperature profile was recalculated at each time step, taking into account energy production, transfer, and transport. In the radial direction, the magnitudes of the temperatures 11c. W. Nestor, Jr., "Zorch, An IBM-7090 Program for Analysis of Simulated MSRE Power Transients with a Simplified Space-Dependent Kinetics Model," ORNL-TM-345 (in preparation). Tablz 2.5, Deiayed Neutrons in MSEE Group - Total 1 2 3 L = 6 Half-life, sec 55.7 22,7 .22 2.30 0.61 0.23 B, (see foot- 2,11 x 107% 14.02 x 127* 12.5% x 107%* 25.28 x 1¢"* 7.40 x 107%* 2.70 x 10°* 6k x 107* note a) _ ¥* B;, static (see 2.23 x 107% 14.57 x 19™* 13.07 x 107* 26.28 x 1¢™* 7.65 x 10”* 2.80 x 10”* 67 x 10°% footnote ) - - 6;,'circulating 0.52 x 107% 3.73 x 10~* k.99 x 107% 16.98 x 10™* 7.18 x 107* 2.77 x 107* 36 x 107* ®precarsor yield per fission neutron. bEffective fraction of nesutrons in a given delay group. 8¢ UNCLASSIFIED UNCLASSIFIED ORNL—LR—DWG 73623 ORNL-LR-DWG 73544 1500 1450 //“\ / \ TIME e /\\\ 3.0sec § T / N 2.0sec = 1300 g / \ {.0sec s Osec ///(/’—_ 12€0 / 1200 % // % INITIAL POWER =10Mw 1150 0] 0.2 04 0.6 08 1.0 [o0) Q1 Q2 Q3 oy Qs Qa6 Q7 Qa8 Q97Z/H 1.0 FRACTION OF DISTANCE ALONG CHANNEL Fig. 2.16. Output of Calcomp Plotter: Fuel and Graphite Fig. 2.17. Fuel Temperature in Temperature Distributions in Hottest Channel with Reactor at Hettest Channel Following One Dollar 10 Mw. Step Insertion of Reactivity. 6¢ 40 and power density varied with time; however, the power density shape was assumed to remain the same as initially. Results of calculations for the MSRE with fuel containing 1% thorium were obtained for a test case in which one dollar of reactivity was added instantaneously (Beff was assumed to be 0.0034). The initial power level was 10 Mw. The maximum rise in the fuel nuclear average temperature was 96°F. (This was about 4°F higher than the maximum rise in the fuel mean temperature computed in the Murgatroyd program.) Also the rate of rise of the nuclear average temperature was higher than the rate of rise of the mean reactor temperature. The maximum power level calculated with the Zorch program was T8 Mw, as compared with 107 Mw from the Murgatroyd calculation. As an option in the Zorch program, the temperature profiles in fuel and graphite at specified time steps may be plotted using the Calcomp digital plotter. The temperature profiles at t = O, obtained for the test case, are shown in Fig. 2.16. The change in the fuel temperature with time following the reactivity insertion is indicated in lig. 2.17. 41 3. COMPONENT DEVELOPMENT Freeze-IFlange Development A pair of INOR-8 freeze flanges for 5-in.-diam sched-40 pipe was installed in the freeze-flange-seal testing facility® for measurement of thermal distortion and gas leakage during high-temperature operation. Following the successful completion of this series of tests, each of two flanges was mated with new INOR-8 flanges and placed into actual salt service. ‘lhe several tests are described below. Freeze-Flange-Seal Test The flanges mounted in the seal-testing facility were thermally cycled in air, and precise measurements were made of flange dimensions and of helium leakage at the ring-joint seal. This preliminary testing consisted of 10 cycles from a cold condition to a hot eguilibrium con- dition. The equilibrium temperatures were varied at 1O0CF intervals from 800 to 1400°F. Heat was provided by 12 silicon carbide heating elements in the bore of the flange and by removable pipe-heaters2 along the ex- terior of the flange adaptor, beginning 2 l/h in. from the flange faces. At a bore temperature of 1400CF, the radial position of the freeze point was 7.l in., which is believed to be satisfactorily far from the ring-gasket radius of 10.4 in. The average decrease of the freeze-point radius was ll/16 in. for cach 100°F decrease in bore temperature. The external thermal distortion of the 5-in. INOR-8 flanges, both at temperature equilibrium and during temperature transients, was found to be much less than the thermal distortion of the 6-in. Inconel flanges tested previously.® (The flanges differed only in material and bore diameter.) The INOR-8 flanges operated cooler than the Inconel flanges, as shown in Fig. 3.1. The distortions under equilibrium conditions are given in Table 3.1. Starting with equilibrium bore temperatures of 1300 and 1400°F, all power to the silicon carbide heating elements in the INOR-8 flanges was cut off to produce a transient simuwlating a salt dump. When the bore temperature reached 850°F, the maximum external distortion wa.s 6 mils, indicating that little distortion is induced during tempera- ture transients. This limits the amount of frozen salt that could be trapped between the flange faces during a rapid cooldown and which could lead to permanent flange distortion. No permanent distortion was noted at the outside of the flanges. Upon opening the flange joint, how- ever, the inner faces were found to be slightly convex; the axial dis- placement of the center of the bore was 1k mils. 1"MSRP Prog. Rep. March 1 to Aug. 31, 1961," ORNL-3215, p 20. 2"MSRP Semiann. Prog. Rep. Feb. 28, 1962," ORNL-3282, p 26. 3Tbid., p 32. 42 UNCLASSIFIED ORNL-LR-DWG 73982 1400 S \\ 1200 §} \ \ \\ 1000 °-\ S ™ \\ N & [ ___850°%F NN N N ] W oo [LFREEZING POINT | N\ e [N D \ 4 g :F ‘\:\\w @ . ) 600 Se 0N i & \\‘D'\o 400 | FLANGES: o INCONEL e [NOR-8 200 0 0 2 4 6 8 10 DISTANCE FROM PIPE CENTER LINE {in.) Fig. 3.1. Temperatures of Inconel and INOR-8 Freeze Flanges.at Com- parable Distances from:Center Line of Pipe. Flanges tested in air. Table 3.1. Maximum External Thermal Distortion of Freeze Flanges at ‘Various Bore Tcmperatures Equilibrium Bore Distortion (mils) Tempgrature 5-in. INOR-8 6-in. Inconel (°F) F]-angea Flangeb 1400 22 63 1300 18 6. 1200 12 pal 1000 8 2k ®Distortion measured 4.50 in. from center line. bDis‘cortion measured 5.33 in. from center line. 43 Two nickel ring-joint gaskets were tested. For the first five cycles, an oval ring gasket was used; for the second five cycles, an octagonal ring gasket was installed. Indicated leakage, with the buffer space evacuated, was very low at all times. There was a tendency for the leak rate to increase as the number of cycles increased and to decrease at higher temperatures. Leak rates at the outside of the buffer groove were consistently higher than the leak rates to the bore through the inner seal of this buffer zone. Specifically, for the oval ring the cold leak rate steadily increased from 3.9 x 10-8 to 1.4 x 10-4 cm3/sec after five cycles. There was initially no detectable leakage at high temperatures, but after five cycles the total leak rate was 1.5 x 10-2 cm3/sec at 1200°F. For the octagonal ring the initial cold leak rate was 8.5 x 10-7 cm3/sec, and after five cycles, it was 4.6 x 10-6 cm3/sec. At 1200°F there was no initial detectable leakage, but after five cycles the leak rate had increased to 3.0 x 10-7 cm3/sec at 1300°F. For this brief period of testing, the per- formance of the octagonal ring was superior to that of the oval ring, but both performed far better than required. Thermal-Cycle Test Loop Immediately following the preliminary testing in the secal test facility, one of the tested flanges was paired with a new flange, and the pair was installed in the thermal-cycling facility.? The freeze-flange thermal-cycling facility permits testing of the flange joint under drain- and-fill conditions, steady-state conditions at various operating tem- peratures, and temperature cycling between room temperature and a selected operating temperature. The flange testing consisted of thermal cycling and of attempts to form a frozen salt plug at the flange bore. Forty-three cycles have been completed to date. A cycle starts with the flange bore at ambient temperature and the sump and pipe temperatures held constant. The freeze flange is then preheated until temperature equilibrium is reached (~ 8 hr) at a bore temperature of 650°F; molten salt at the desired high temperature of the cycle is introduced and oscillated between the two sumps until the flange again reaches tempera- ture equilibrium (~ 5 hr); and, finally, the flange is allowed to cool until the bore is at ambient temperature (~ 10 hr). The total cycle requires approximately 22 hr to complete. The first 15 cycles were made with an average flange bore temperature of 1020°F and an ambient tem- perature of 85°F. For the following 28 cycles, the flange bore tem- perature was increased to 1250°F and the ambient temperature to 150°F. The size of the salt ring was predicted on the basis of the flange temperature distribution, shown in Fig. 3.2, as obtained during cycle No. 35, and the expected 20°F temperature difference between the external and internal flange faces, which was noted in earlier tests. The salt 4"MSRP Quar. Prog. Rep. July 31, 1960," ORNL-301Lk, pp 24-25. UNCLASSIFIED . ONNL-LR-DWG 73983 1400 1200 \ ' A 1000 O SALT FREEZE 800 600 NS TEMPERATURE {°F) VY 400 I o TEATEG WITH SALT IN | HERMAL-UYULE TEST LODP ~ & TESTED WITHOUT SALT IN SEAL TESTING FACILITY 200 — NOTE: TEMPERATURES RECORDED AT THE OUTSIDE FLANGE FACE L 0 2 4 6 8 10 2 DISTANCE FRUM PIPE CENTER LINE (in} Fig. 3.2. Temperature Distribution of a 5-in. Freeze Flange in the Thermal-Cycle Test Loop Compared with the Distribution in Air. freezing point of 8LOCF was reached at an external face temperature of 0z0”F, whilch occurred at a diameter of 10.8 in. in the current series of tests. Examination of the salt screen following cycle No. 36 (Fig. 3.3) showed the salt-ring diameter to vary between 10.5 and 11 in. At the greatest diameter the salt was still 4 in. from the ring. Two mechanical dial indicators mounted on & small lathe bed were employed to measure thermal distortion of the flanges. With the flange bore at equilibrium at 1250°F, the maximum combined thermal expansion and distortion was 25 mils. The calculated thermal expansion was 21 mils, and thus the maximum thermal distortion was 4 mils. No permanent distor- tion of the outer flange faces was detected in the cold flanges after 30 cycles. There was a small increase in the interior face warpage of the female flange from the initial out-of-flatness of 14 mils to 18 mils after 36 cycles. DNo warpage was detected on the male flange face. After 36 cycles all bore dimensions were checked and found to be undersize. The female bore decreased by 10 mils, the male bore by 30 mils, and the male lip by 8 mils. For the first 36 cycles, an octagonal seal ring was used. The initial cold seal leak rate for the ring was between 10-% and 10-3 cm3/sec. After the flange joint had reached its initial hot equilibrium temperature (1020°F), the hot leak rate was less than 5 x 10-9 cm3/sec. When the 45 UNCLASSIFIED PHOTO 38664 S O0CTAGONAL RING INCHES F 1R R T B R T e e 11% }111 | i!M l]in%;Ll:iyhlzhIzislsh!!l Fig. 3.3. Freeze Flange Salt Ring After 36 Cycles. flange again reached ambient temperature, the cold leak rate had decreased to 2.6 x 106 cm3/sec. After 36 cycles the hot lesk rate had increased to 3.9 x 10-7 cm3/sec and the cold leak rate was 1.8 x 10-4 cm3/sec; both these leak rates are considered to be acceptably low. An oval ring was tested in the next 7 cycles. The precycling cold leak rate of 3.7 x 10~7 cm3/sec had increased to 1.7 x 10-2 cm3/sec at the begimning of cycle No. 37. The final hot leak rate at the end of cycle No. 43 was 9.3 x 10-3 cm3/sec. oince this rate was considered to be only marginally acceptable, the flanges were opened for inspection. It was found that foreign material had interfered with proper seating of the ring. A secoud oval ring was inserted and is presently being tested. 46 Freeze Test in Thermal-Cycle Test Loop In order to simulate a pump failure, the salt was held in a stagnant condition between the two sumps to determine whether a freeze plug would form at the flange bore. In the most severe test, the piping was kept heated at 1000°F. Although the flange bore temperature was reduced to 350°F under these conditions, only a partial plug formed. It was con- cluded that the MSRE 5-in.-pipe flanges would not plug so long as the pipe heaters were operative. Test of a Freeze Flange in the Prototype-Fuel-Pump Testing Facility The second INOR-8 flanged joint was installed in the pump-testing facility® (described later in this chapter), with an oval ring-joint gasket. ''o date the facility has been operated approximately 300 hr with three complete cycles from startlup to shutdown. The tempcrature dis- tribution of the flanged joint has closely approximated the distributions measured in both the seal-testing facility and the thermal-cycle Llesl loop. No distortion has been noted, and no leak rates greater than h x 1o=4 cm3/sec have been detected with the buffer seal system of this joint. The buffer type of seal system used is similar to the reactor leak-detection system. Although the test results are encouraging, addi- tional experience is needed before the current freeze-flange design is considered acceptable. MSRE Control Rod A full-scale model® of one MSRE control rod and drive was assembled and operated over a range of temperatures from 500 to 1300%F. The drive is shown in Fig. 3.4. The unit has operated through 24,473 cycles (56 3/4 in. full travel in each direction). The rod has been dropped from various heights 862 limes. While over-all operation has been satisfactory, the drive unit failed when one of the lower idler-sprocket bushings galled as the result of a foreign particle. This bushing was replaced with a shrouded ball bearing, and operation was resumed. Tests indicate that the rod position is known within #1/8 in. 94% of the time and t1/4 in. 100% of the timc. Heater Tests Pipe Heaters Testing of removable hardboard-insulated heaters for 5-in. pipe was completed.” The units operated a total of 6136 hr with a pipe temperature S"MSRP Semiann. Prog. Rep. Feb. 28, 1961," ORNL-3122, p 51. 6"MSRP Semiann. Prog. Rep. Feb. 28, 1962," ORNL-3282, pp 24-26. “Ibid., pp 26-29. 47 e [-CONTAINMENT POSITION i~SPROCKET MAGNETIC \GEAR BRAKE \-COOLING- |-MOTOF\’ VESSEL TRANSMITTER AND CLUTCH REDUCER AIR LINE CHAIN Fig. 3.4. Prototype Control Rod Drive Viewed from Above. under the heaters of 14LO0°F. The average power input was 520 watts per linear foot of heated pipe. The over-all functional operation of the heaters was excellent, but the hardboard exterior tended to crack and dust, the soft mineral wool lining dusted after heating, and the metal liner warped. Three alternate designs being considered to overcome these objections consist of units made with blown fused silica (Glasrock Mfg. Co.), reflective metal insulation (Mirror Mfg. Co.), and combinations of these materials. The materials for use in the MSRE will be selected after tests of the various combinations. : Testing of two prototype heater boxes using the Glasrock material indicated an excessively high heat loss of 700 watts/ft. The surface tem- perature of the heater boxes was 2LO°F when the heaters were operated at lMOOOF; this would lead to excessive heat losses in the MSRE cell. Further the material is unable to withstand physical abuse without cracking or chipping. On the other hand it has the lowest coefficient of expansion of any of the materials tested to date and thus has very high resistance to thermal shock. There was no evidence of warping at the joint closures at the completion of the test. All-metal prototype heater boxes made of reflective insulation have been purchased from Mirror Mfg. Co. for testing, and a test unit made of Glasrock, board insulation, and metal reflective sheets is being fab- ricated. Activation Analysis of Insulating Materials Most high-temperature insulation materials will dust to some extent, and the dust will present a problem in reactor maintenance if it is highly activated. The insulating materials listed in Table 3.2 were therefore tested for activation following irradiation for 16 hr at a neutron flux 48 Table 3.2. Results of Activation Analysis of Insulating Materials Observed Radioactivity Stable Element Concentration Sample Material : : Activity Amount Radionuclide (dis/sec-g)a Element (wt %) Super Temp 260-d Zn®S 2.13 x 102 Znb4 7.8 (t0.2) (Eagle-Picher Co.) 60-d Sbla4 1.91 x 10 Sbl23 0.0058 (£0.0005) 40-h Lal40 8.64 x 1073 1al3® 0.0019 5.27-y Co®© 1.99 x 102 Co5® 0.028 (*0.001) 85-d Sc46 1.85 x 103 Sc45 0.00014 45-d4 FeS® 5.60 x 104 Fe58 20.3 (%0.1) Cerafelt LO-h Lal40 3.63 x 103 Lal3® 0.00008 (Johns-Manville) 85-d Sc4° 9.25 x 103 Sc45 0.0007 45.9 Fe5° 1.27 x 103 FeS8 0.46 (£0.02) 15-l Na? 1.10 x 108 123 U.060 (LU.003) Unarco Board 60-d Sbl24 6.57 x 102 Sbl23 0.002 (Union Asbestos and 5.27=y Co8° 2.11 x 104 Co>®° 0.0003 Rubber Co.) 4o-n Lal40 2.26 x 103 Lal32 0.00005 85-d Gc46 1.98 x 103 Sc#5 0.00015 45-d4 FeS52 8.84 x 10° re>8 3.2 (£0.1) 15-h Na24 6.82 x 10”2 Na23 0.035 (+0.002) Carey Temp LO-h Lal4® 3.06 x 103 Lal3® 0.00085 (Philip Carey Manufacturing 85-4 Sc46 1.12 x 10t Sc45 0.0008 Boe ) 45-4 FcS° 5.02 x 103 Fe58 1.82 (%0.02) 15-h Na24 8.57 x 107 Na23 4.40 (*0.05) Fiberfrax LO-h Lal40 2.72 x 102 Lal3® 0.000006 (Carborundum Co. ) 15-h Na24 1.82 x 107 Na23 0.96 (%0.01) Glasrock” 4O-h La4© 9.10 x 103 Lal3® 0,0002 (Glasrock Manufacturing Co.) 85-d Sc4® 3.96 x 102 Sc45 0.00003 15-h Na2% 7.50 x 10% Na23 0.0038 Glasrock Cement LO-h Lal40 1.36 x 10% 1al3® 0.0003 (Glasrock Manufacturing Co.) 15-h Na24 6.24 x 102 Na23 0.033 #Measured approximately 24 hr after reactor discharge. bDetermined from the count rate observed for each radionuclide. ®Density: 25 1b/ft3. of about 7 x 101t neutrons/cmz-sec. The data obtained (see Table 3.2) indicate that Glasrock, which is nearly pure silica, may be the only ceramic insulation that can be utilized uncanned in the MSRE. Drain-Tank Coolers The prototype cooling bayonets for removing afterheat from fuel in the MSRE drain tanks were thermally shock tested 384 times without failure. The test system was then shut down for installation of proto- type thermocouples (see later section of this chapter) on the l1-in. cooling bayonets that slide into the 1 l/Z-in. thimbles in the salt pool. 1In previous operations, thermocouples installed in this area came loose in a few cycles. The reactor-grade thermocouples now being tested are intact after 6 cycles and 48 hr of exposure at 1L4OQCF. 49 A carbonate salt is being used in place of the MSRE fuel salt in these tests to avoid beryllium-handling problems. The carbonate consists of LipCO3, NapCO3, and KpCO3 (38-40-32 wt %). Its physical properties at 1200°F are given below, along with those of the fuel salt, for comparison: Fuel Salt Carbonate Salt Density, 1b/ft> 154.5 138.5 Average specific heat, Btu/1b*°F 0.455 0.391 Viscosity, cp 7.64 7.8 Sampling and Enriching System Detail Design Drawings of nearly all the components of the sampling and enriching system® were issued for comment. Detail drawings of the transfer tube and the spool piece between the disconnect flanges, with its positioning tool, were approved for construction. Instrumentation flowsheets and panel layouts were also approved for construction. Mockup of Sampling and Enriching System A full-scale mockup of the sample transfer box (area 3), the drive box (area lc), and the operational valve is being fabricated for installa- tion on the Engineering Test Loop. The transfer tube, which had to be shortened to fit into the space available, has bends and slopes that are the same as those to be used in the reactor system. The mockup will be used to demonstrate that each component operates satisfactorily and to develop adequate sampling and enriching procedures. The sampling-capsule access-chamber mockup was successfully hydrostatically tested at 90 psig to check the tightness and strength of the access port. The copper sampling capsule was modified to simplify handling, and remote removal of a sample from the revised capsule was demonstrated. Core Development Model Operations The full-scale MSRE core model was operated at 85°F with water to make preliminary measurements of the velocity distribution, wall heat transfer coefficients, and solids-handling characteristics. Since the viscosity of water is lower than that of the fuel salt, the model is not exactly similar to the MSRE with respect to fluid dynamics. The pre- liminary tests will be usetul, however, in planning later tests with a more viscous liquid. ®Ibid., p 30. 50 Velocity Measurements The peak resultant velocities at the bottom of the core-wall cooling annulus were measured at various flow rates as a function of the angle around the core. The data obtained are presented in Fig. 3.5. As may be seen, the velocity profile is flat, indicating that the flow entering the lower head from the annulus is distributed uniformly. Based on the observed resultant velocity of k.55 ft/sec and the known axial velocity component of 2.15 ft/sec at 1200 gpm, the average tangential velocity component is k.0 ft/sec at the bottom of the core annulus. A discrepancy was observed between the flow in the one-fifth .scale model and the full-scale model. In the one-fifth scale model, the flow in the lower head was observed to have a large rotational component, whereas, in the full-scale model, the flow was observed to be almost purely radial. On examination of the one-fifth scale model, it was noted that the swirl killers were not resting on the lower head but, rather, wcre celevated 1/8 to l/h in. off the head. This allowed water to flow underneath the swirl killers and maintain its rotational component. New swirl killers were made for the one-fifth scale model, and the flow then behaved the same as that in the full-scale model. With this change the measured heat transfer coefficients at the center line of the one-fifth UNCLASSIFIED ORNL-LR-DWG 73984 L, . gl . l o 's 0 o I o 1200 gpm 4 ° Qv v pd .>__ . G 3 O s w , > a’ 2 L Q > it % o) 2 -0 1o o} Q 4 2 600 gpm > 7 w o = ) E ¢ — 3 < A Fa 4 hd Y 300 gpm - o 6] 90 {80 270 360 ANGLE AROUND CORE (deg) Fig. 3.5. Resultant Velocity Distribution Around Bottofi of Core-Wall Cooling Annulus at Various Flow Rates. - Lt 51 scale model were higher than those measured previously, so the change was inferred to be beneficial. Measurement of Heat Transfer Coefficients Heat transfer coefficients are indices both of wall cooling and of wall shearing, which tends to sweep out sediment. Coefficients measured in the lower head at a radius of 17 in. were found to be proportional to the 0.8 power of the flow rate, with a value of 522 Btu/hr.ft2-CF at 1200 gpm. Extrapolating the measurements to reactor conditions, the coef- ficient for the reactor at 1200 gpm was estimated to be 1500 Btu/hr-fta-OF. As a check on the water heat transfer coefficient, the fluid velocity profile next to the wall was measured at the same point the coefficient was measured. The velocity profile shown in Fig. 3.6 was found. A heat trans- fer coefficient can be calculated from the data of Fig. 3.6 by assuming flat-plate geometry and that the distance between plates is twice the distance from the wall to the point of peak velocity. The calculated heat transfer coefficient for this case is 522 Btu/hr-ftz-OF, which is in fairly good agreement with the measured value of 667 Btu/hr-ft2.CF. Attempts wcre made to infer the heat transfer coefficient at a position about 4 in. from the core center line from data obtained with a UNCLASSIFIED ORNL-LR-DWG 73985 3.0 n O h7¢ » / FLUID VELQCITY (ft/sec) o / o NORTH POSITION N N e EAST POSITION 0.5 | & SOUTH POSITION A NORTH WEST POSITION 0 0 0.5 1.0 1.5 2.0 2.5 DISTANCE FROM WALL (in.) Fig. 3.6. Velocity Profile of Fluid Near Wall of Lower Vessel Head 2L Rudius of 17 in. 52 heat meter. These attempts were unsuccessful because there was no pre- dominant direction of flow at that position; rather, the flow was "gusty." The flow was first in one direction and then another. It was determined that the heat transfer coefficient would be lower than at a radius of 17 in., but the exact value was not obtained. Further attempts to measure the heat transfer coefficient in this region will be made. Experiments on the Behavior of Solids Experiments were run to obtain data for use in predicting the behavior of solid particulate matter in the fuel salt, especially its tendency to settle out in the reactor vessel lower head. The experimental procedure consisted of injecting 3 1b (1362 g) of sized iron filings into the model inlet during operation. After the solids were added, circula- tion of water was continued for various periods nf time (0 %o 8 hr). The core was then drained, and the quantity of solids that remained in the bottom of the reactor vessel was determined. As far as the solids werc concerned, the loop was considered as a once-through system. The sedimen- tation data obtained are summarized in Table 3.3. Table 3.3. Collection of Sediment in Lower Head of Core Model Following Addition of 1362 g of Iron Filings Nominal Time Pump Running Weight of Iron Run Mesh Screen Required Time After Filings Re- Nd’ Size Aperture to Add Solids covered from ) of () Solids Added Lower Hcad Solids (min) (hr) (g) € 150 1oL 35 0 127 2 150 104 35 8 0# 3 60 248 20 0 823 L 60 2L8 20 8 750 5 150 10k 30 71/2 02 %A11 solids swept out with water. In the case of the fine solids, there appeared to be some tendency for sediment to form during their first pass through the core. Additional running time caused them to be picked up, however, and swept out.. 1In the case of the coarse material, a larger fraction settled out on the first pass through the core and very little was removed with additional pumping time. The 150-mesh solids that remained in the lower head appeared in a concentrated pile near the drain line. On the other hand, the 60-mesh solids, as shown in Table 3.3, left larger amounts of sediment that tended to form a torus around the drain. These larger piles were 2 to 3 in. wide and from 1/2 to 1 in. thick, and they were against the drain line on one side. The results of these tests indicate that the MSRE drain line will 53 not plug as the result of sedimentation of solids. Although no sedimen- tary materials are expected in the reactor, these experiments provided confidence in the design. MSRE Engineering Test Loop (ETL) The engineering test loop® was modified to include a graphite con- tainer in the salt-circulating loop, and most of the Inconel pipe was replaced with INOR-8 pipe. The loop was then placed into operation with- out graphite in the graphite container. Most of the previous zirconium- containing flush salt was removed, and the drain tank was flushed with newly made coolant salt. Information was obtained during 1500 hr of operation on the operational characteristics of the graphite-container access joint, on the chemical analysis of the salt being circulated, and on the formation of salt deposits above the liquid level. After receipt, inspection and cleaning of the graphite to be tested, the loop was shut down for installation of the graphite. Also, as part of the proposed graphite-treatment procedure,lo the flush salt was treated with HF and Hp2 in the drain tank to lower its oxide content before it contacted the graphite. Details of these tests are discussed in the following sections. Frozen-Salt-Sealed Graphite-Container Access Joint The flush salt was circulated at 50 to 70 gpm and 1200°F through the loop and the graphiie container while awaiting delivery of the ETL graphite from the vendor. The graphite container, which is an 8-in.-diam INOR-8 pipe positioned vertically in the loop,9 has a flanged closure that incorporates a frozen-salt seal to prevent contact of molten salt with the conventional metal oval-ring seal. This closure is similar to a slightly larger one that will be used in the MSRE. A seal of this type between graphite and meta) was tested previously.tt The ETL graphite-container closure is an extension of the 8-in.-diam graphite container with a 20-in.-long tapered plug inside to form an annutus varying from l/h—in. radial clearance at the lower end to l/lé-in. radial clearance at the upper end. The MSRE closure is 10 in. in diameter and 28 in. long, with gaps of l/h and 1/8 in., respectively. The tapered gap prevents axial movement of frozen salt as the result of surges in salt pressure. The seal was made and broken three times, and it was leak-tight during 1500 hr of operation. Since gas is trapped in the annulus during B S ®Ibid., pp 34-4O. 107, L. Crowley, "Preliminary Operating Seguence for Graphite Testing in the ETL, " March 19, 1962, unpublished internal report. 117. L. Crowley and W. B. McDonald, "A Metal to Graphite Joint for a Molten Salt System," ORNL CTF-60-8-82, August 17, 1960. 54 filling, either initial evacuation or latcr ovcr-pressurc or venting must be used to force salt into the cooled region. On the ETL closure the gas was partially vented after the loop was filled. An interesting observation made while determining the minimum cooling air required for the seal was that when the cooling air flow was reduced below the critical value, the salt plug moved upward; however, instead of moving uniformly around its circumference, in which case gas compression would have limited its travel, the salt extended upward in one semi-molten sector that was contained within frozen salt at the edges. This character- istic precludes the use of trapped-gas compression for preventing salt from reaching the metal oval ring unless the entire-joint is held at tem- peratures above the salt melting temperature. The temperature distribution and a diagram of the closure are shown in Fig. 3.7. Salt Composition In order to remove as much of the ZrF) from the system as possible, the original flush salt was removed, the freeze valves were purged, and new salt was introduced into the unused drain tank (previously designated the fuel drain tank). Salt samples were removed from the pump while in UNCLASSIFIED ORNL-LR—-0OWG 73986 TN, PI N N gl | / THER%&?OUPLE | . | | : : ' : | | I Ak ‘ 18 [-—— 0 COOLING AIR FLOW RATE: 23.8 s¢fm - ] EXTERNAL COOLING —— {14 1 INTFRNAI + 97 FXTFRNAI }; 344kw RFMOVFN — Yl ON 8-in. for use in measurement of molten-salt levels in the MSRE fuel and coolant pump bowls was continued. A four-month initial test at teme peratures between 900 and 13OOOF was terminated June 4. During the last three months of this period the temperature was maintained at 1200°F, and the molten salt was held at a constant level. The maximum variation recorded during this period was O.l in.; the span of the instrument is 5 in. Tests made at the end of the four-month test period revealed no evidence of calibration shift, hysteresis, or other degradation of the performance characteristics of the instrument in excess of 2% of full scale (0.1 in.). Calibration curves made three months apart are shown in Figs. 3.14 and 3.15. There has been no indication of buildup of solids as a result of vapor deposition in the core tube of the instrument; how- ever, solid deposits were found on a spark-plug level probe removed from the test stand at the conclusion of the four-month test. After completion of the four-month test, the instrument was drained and the temperature was cycled three times between 500 and 1000°F. Subsequent checks revealed no evidence that the instrument was damaged by this repeated heating and cooling. The instrument was returned to operation at 1200°F, and long- term testing is continuing. A third instrumcent is being constructed and will be inslalled on the MSRE pump testing facility to obtain experience under operational conditions approximating those which will be encountered in the MSRE. This instrument will use an INOR-8 metal-ball float, since graphite will not float in the barren salt used in this facility. The transformer will be mounted above the float because this type of assembly is preferred for the MSRE installation and because tests of a similar unit, described above, revealed no objectionable operational characteristics. Single-Point Temperature Alarm System Investigations of single-channel thermocouplc alarm switches for use in freeze-flange and freeze-valve monitoring and control systems con- tinued. Several components of the Electra Systems Corporation=% moni- toring system were received in late March. The system included a dual- limit alarm module (ET-4210), two single-limit alarm modules (ET-4200), a control module (ET-MBOO), and the power supply and module enclosure. The units were tested in the laboratory. As previously reported, &4 the single-limit alarm modules appear to be suitable for freeze-flange 23"MSRP Semiann. Prog. Rep. Feb. 28, 1962," ORNL-3282, pp 61-66. 247pid.; p 60. 69 UNCLASSIFIED ORNL-LR-DWG 73988 O CALIBRATION 6-6-62 90 |— ® CALIBRATION 3-16-62 { TRANSFORMER [N 1225°F FURNACE 80 // ® 70 / S 60 3 << Ll o = n T B0 ° o f O (®] =i < = & 40 30 o 20 10 o 0 » 4 5 6 7 8 9 10 TANK NO. 2 SCALE READING SHOWING MOLTEN-SALT LEVEL CHANGE (in)) Fig. 3.14. Calibration Curve of Recorder of Pump-Bowl Liquid-Level Indicator No. 1 Before and After Three Months Operation at 1200°F. 70 UNCLASSIFIED ORNL-LR-DWG 73989 100 ’ ’ ® O CALIBRATION 6-6-62 90 |— ® CALIBRATION 3-16-62 TRANSFORMER ABOVE FURNACE AND NOT HEATED 80 /# ° {/ g 60 (&) < (0] (0 [—. 2 < 50 /T (&) O} & (&) | —<’ / > 40 5 30 20 10 U 4 5 6 7 8 9 10 TANK NO. 2 SCALE READING SHOWING MOLTEN-SALT LEVEL CHANGE (in.) Fig. 3.15. Calibration Cirve of Recorder of Dump-Bowl Liguid-Level Indicator No. 2 Before and After Three Months Operation at 1200°F. b monitoring. Recent tests indicate that control module ET-4300 is suitable for the valve-monitoring system. The evaluation thus far is based only on laboratory tests. The Electra System was placed in field operation on a freeze-valve testing facility in mid-August. It is now undergoing full field tests utilizing the control system to be used for the reactor. The Electra System is composed of modular units, as previously described.2® The internal wiring of one module card containing two single-alarm switches is shown in Fig. 3.16, and the entire system is shown in Fig. 3.17. The top unit is the power supply and the bottom unit the module enclosure. The module enclosure will handle 20 single-limit alarm switches or 10 control modules. Tests of the Daystrom Magsense Control Relay, Model A-82, were completed. As previously reported, &2 this unit could be utilized for the flange- and valve-monitoring systems; however, the individual units would have to be integrated into a system. This would require adding alarm-1limit lights, relays, and enclosure, as well as a common power supply. Further, the input resistance of the units is so low (350 ohms) that it could cause a problem with long thermocouple leads. Temperature Scanner Development work on the mercury-jet-commutator thermocouple-scanning=e system is continuing. The switch was modified to reduce spurious noise generation. An alarm discriminator was designed and constructed, and the reference-thermocouple isolation system was improved and packaged with the alarm discriminator. A differential dc amplifier with variable band- width was received, and an existing 17-in. oscilloscope was repaired to display the thermocouple signals. The complete system was assembled and tested in the laboratory in early May. A diagram of the system as it now exists is shown in Fig. 3.18. The laboratory tests with simulated thermocouple signals were successful, and the system was installed on a liquid-level test facility in early August. One hundred thermocouples from the test facility were used as input signals to the scanner. At present, the system continues to operate satisfactorily after more than 700 hr of operation. Operation will be continued to determine the operating life of the mercury switch. The diagram of the system (Fig. 3.18) shows the principle of opera- tion. One hundred thermocouples are switched in sequence by the mercury- jet switch at a rate of 2000 points per second. The output of the switch is fed to an integrating system consisting of a capacitor and the re- sistance of the dc amplifier. The signal level is held during the point- 25"MSRP Prog. Rep. March 1 to Aug. 31, 1961," ORNL-3215, p 25-27. 26"MSRP Semiann. Prog. Rep. Feb. 28, 1962," ORNL-3282, p 59. Fig. 3.16. Dual-Limit Alarm Module. UNCLASSIFIED PHOTO 38688 cL SERIES 4000 ypERAFINS MNITOR SYETS vSTEMS COR ELICT::L&YONV CALIFORNIA AL roAtpn powWER SUPFLY 5 £ pese -, G o B v 8 K - < % - R ¥ Pig. 317 1200-rpm MOTOR 73 o, gr4008 fr-a00 WODEL "/ UNCLASSIFIED PHOTO 38689 UNCLASSIFIED ORNL-LR—DWG 58910 R3A THERMOCOUPLES ON REACTOR PIPING . TWO DOURIL F-POI F DOUBLE- THROW CHOPPERS OPERATING SYNCHRONOUSLY n ‘ ’ [ ONE COUPLE ABOVE xREFERENCE REFERENCE je——1 { | | J”L' 5| THERMOCOUPLE — | L TN T K< > R = O o e < - :5:; | 2 , 00000 | ~N— P i ~ | ... - ~17-in. OSCILLOSCOPE l«—REACTOR *REFERENCE APPROXIMATELY PIPE EQUAL TO SCANNED THERMOCOUPLE SIGNAL | = o B DISCRIMINATOR ALARM ANNUNCIATOR MERCURY-JET SWITCH SETPOINT (100 CONTACTS) + 50°T Q K=" = N o REFERENCE THERMOCOUPLE ON REACTOR PIPE fig. 3.18. Block Diagram of Temperature Scan §G 1 i T4 to-point switching time. The reference thermocouple signal is fed to two double-pole double-~throw choppers with a capacitor between the poles of each chopper. The reference thermocouple is connected in opposition to the output of the mercury switch. The choppers store the charge from each side of the reference thermocouple during one half of the operating cycle and then switch it across the large integrating capacitors in each out- put line of the mercury switch. The difference signal is fed to the dc amplifier and amplified to 1 volt. The output of the amplifier is fed to the oscilloscope and to the alarm discriminator. The alarm discriminator can be adjusted to produce an alarm when the difference signal exceeds an adjustable range of 50 to t300°F. If the reference signal and the commutated signal are equal, a straight line is seen on the oscilloscope. If a signal from one thermocouple is higher than the reference, it is fed through the system as an ac signal and appears as a positive pulse on the oscilloscope. By using a synchronizing pulse from the switch, the 100 thermocouple signals are displayed on the oscilloscope in a fixed position so that each signal is identified with a particular thermocouple. The switch modification consistcd of installing shorting straps between the coalescer coll and the common mercury pool, as shown in Fig. 3.19. This reduced the static charges which were generated between the coalescer coil and the mercury pool and resulted in a significant re- duction in the spurious noise generated in the switch. A block diagram of the alarm discriminator is shown in Fig. 3.20. This circuit is designed to produce an alarm when pulse signals having an amplitude of 1 volt or greater and a repetition rate of 20 pulses per second are applied to the input. With proper adjustment of the differen- tial dc amplifier, the discriminator may be adjusted to produce an alarm with signals as small as t50°F. A pulse-integrating (count-rate) circuit prevents spurious alarms from random noise pulses. Since the mercury scanning switch samples each point 20 times per second, input pulses resulting from a true alarm condition will have a repetition rate of at least 20 pulses per second and an alarm will occur. The discriminator- alarm circuitry is completely transistorized and is packaged in a 3 l/Z-in.-wide, T l/Z-in.—high, panel-mounted housing. Also packaged in the housing are the reference-thermocouple isolation system, a sync-pulse amplifier for driving the oscilloscope trigger, and the necessary power- supply circuits. Thermocouple Development and Testing Radiwlor Thermocouple Tect. Teoting of mecchanical thermocouple attachments for use on radiator tubes®’ in the MSRE was resumed after the test apparatus was modified to eliminate errors in measuring the inner wall temperature. Provisions for purging with inert gas were added; the reference thermocouple was attached to the inner wall of the INOR-8 tube by spot welding; and thermal insulation was placed between the thermo- couples and the copper slug (Fig. 3.21). =T Thid., p 98 UNCLASSIFIED ORNL—LR—-DWG 72559 SIGNAL PIN ROTOR ASSEMBLY HARNESS GASKET,NOZZLE /ASSEMBLY ( . NOZZLE COUNTERWEIGHT A II é 'Ill’ ‘ S ¢ Taks §v—=—, g% o1 L. \\- CLAMP ls ’\ Q / ASSEMBLY ” .,‘ " // > COALESCER "0 RING \\\\\ SHORTING STRAP \ ~MERCURY POOL | 1@'@ COO 00 1O i —_ e . —— MOTOR ASSEMBLY J\ Qg COOO00O A Fig. 3.19. Model 210 Delta Switch. 76 UNCLASSIFIED ORNL-LR-DWG 72558A " | DISCRIMINATOR .y AT INPUT SIGNAL TRISGER LSE 1 volt >+ 50°F SHAPER CIRCUIT 0 ) o O] I O EMITTER REI AY RIVER gé URIVE FOLLOWER Fig. 3.20. Block Diagram of Alarm Discriminator. UNCLASSIFIED ORNL-LR-DWG 73990 REFERFNCE THERMQCOUPLE SPOT WELDED TO WALL THERMAL HEATER THERMOCOUPLE INSULATION INOR-8 TUBE INCONEL-SHEATHED TEST THERMOCOUPLE Z22 m : 3 SOSNN N AN SN N S SO N SOOSNNN SN SNSRI AT A T A AT 3 WS Lt Lo L ) 4 > 2 L7 7 Z Z = ; P AT 3 ? | NNNNN\N| 1% 7 e % / . I . : _ASSESSXY < aasSsSSsSs S SS OO IS VAR A R I I I T T T T T T T T I T T I T T I T M LTI T L LI LI LLL LALLM LS LL LU L R R R R, PURGE-GAS “ PURGE-GAS WEDGE-FITTED LARIRIDGE=- “INCONEL CLAMPING BAND INLET ) FXHAIIST COPPER SLUG TYPE HEATER ATTACHFN TN TEST THERMOCOUNLL WITH GOLD-NICKEL BRAZING ALLOY HEATER AIR STREAM POWLR LEADS Fig. 3.21. ©Sectional View of Radiator Thermocouple Test Apparatus. Several test thermocouples were preparcd with one section of a Lwo- piece clamp made of 3/16-in.—wide, 0.020-in.-thick Inconel attached to the sheath near the end seal with gold-nickel brazing alloy. The thermo- couples were then attached to the tube by placing the npper section over the tube, engaging the lower section, and crimping both, as shown in Fig. 3.82. Several tests were run with Lhe thermocouples attached as described above and with thermal insulation consisting of different thickncss of Fiberfrax board or paper placed over the junction end of the thermocouple. The thermocouple shown at the far right in Fig. 3.22 is insulated with Fiberfrax paper, and the thermocouple second from right is insulated with Fiberfrax board. The accuracy of the wall-temperature measurements with - e eecem UNCLASSIFIED PHOTO 38840 L Fig. Z.22. Radiator Tube Thermocouple Attachments. 78 the test thermocouple was improved considerably with both forms of , Fiberfrax insulation. The flexible paper form of Fiberfrax is preferred over the rigid board, however, because it requires no machining or pre- forming prior to installation. Typical results obtained in recent tests of thermocouples attached and insulated as described are listed in Table 3.4. The test thermocouple used in run 5 was covered with Thermom, a heat-conductive cement rated to 1250°F, and 3 layers of l/8-in.-thick Fiberfrax paper. It is not known yet whether Thermom will be acceptable for this application. A test specimen of a typical thermocouple attach- ment covered with Thermom is now undergoing a corrosion test. Engineering Test Loop Installation. Eight MSRE-prototype surface- mounted thermocouples were installed in the ETL facility to determine, under simulated operating conditions, the reliability of attachments and the accuracy of wall temperature measurements taken with thermocouples located on the walls of pipes and components adjacent to heaters. Sheathed, l/l6-in.-OD, single-conductor two-wire, and l/8-in.-OD, sheathed, duplex thermocouples were mounted in pairs adjacent to 30-ga bare-wire reference thermocouples at three locations. A similar pair of thermo- couples was located adjacent to a reference thermocouple installed in a well at a fourth location. A typical installation is shown in Fig. 3.23. A1l the thermocouples have been in service since April 16, 1962, and are still performing satisfactorily; however, they have accumulated only 1500 hr of operation time at 1040 to 1200°F during this period because of loop downtime. With salt circulating at the above temperatures, differ- ences in readings between the test thermocouples and the respective Table 3.4. Results of Tests of Radiator Tube Thermocouples Approximate Inner Wall Tect Thermocouplc Tcmpcraturc fig?& Air Flnw Tpmpgrflturc Temperature Diffgrcncc (£ps) (°r) (°F) (°F) 1 0 996 985 -1l 90 990 95k -36 % 0 e uslk -1k 70 95 905 -30 g 0 1009 996 -13 50 1009 983 -26 L 0 1009 996 -13 50 1000 972 -28 5 0 1007 1002 -D 50 1010 1001 -9 “The test thermocouples used in runs 1, 2, 3 and 4 were insulated with Fiberfrax board; the run 5 test thermocouple was covered with Thermom and insulated with Fiberfrax paper; all readings were taken with the test thermocouple downstream of the air flow. Fig. 3.23. Thermocouple Installation on Engineering Test Loop Piping. UNCLASSIFIED PHOTO 38195 6L 80 reference thermocouples, as noted periodically, were inconsistent. The limits of variation in temperature difference between an individual test thermocouple and its respective reference thermocouple are listed in Table 3.5. Drift Test. Six Inconel-sheathed MgO-insulated Chromel-Alumel thermocouples are being tested to determine the stability of calibration at MSRE temperatures. Over 5000 hr of soaking time in 1200 to 1250°F air has been accumulated to date. None of the thermocouples have drifted more than f2°F equivalent in emf output. Bayonet Thermocouples. Ten MSRE-prototype wall-mounted thermocouples were installed in the drain-tank bayonet-cooler test facility to determine the effects of fast temperature transients on the life of these thermo- couples and to determine how fast they would respond to transients. The thermocouples tested consisted of both l/l6-in.-OD, sheathed, single- conductor, two-wire and l/8-in.-OD, sheathed, duplex thermocouplecs with Junctions grounded to walls and end seals. One l/8—in.—OD unit failed immediately after startup of the test. Its early failure is thought to have been due to faulty construction, since the other nine units have been subjected to rapid temperature changes between 1200 and 600°F for over a week without failing. The thermocouples that endure the thermal- cycling test will be checked for speed of response. Freeze-Valve Thermocouple Tests. Six MSRE~prototype wall-mounted thermocouples were installed on a freeze valve in a test conducted to determine the durability of these units in this type of service and to determine the accuracy of the measurement of wall temperature at the cooled and heated area of the valve. Sheathed, l/l6-in.-OD, single=- conductor and l/8-in.—OD duplex thermocouples were mounted in pairs adja- cent to a 30-ga bare-wire reference thermocouple at three locations on the valve. The locations and methods of attachment are shown in I'ig. 3.24. The test thermocouples agreed with the reference thermocouples, Table 3.5. Variations in Readings of Thermocouples Installed on the Engineering Test Loop Outside Diameter Limites of Varia- Type of Re%grezce Station of Test tion ig Tempersa.- Thermocouple Ther@ocouple ture Differcnce (1in.) (°F) No. 30 AWG bare wire 1 1/8 +1 to +6 1/16 £l to + E 1/8 -2 t6 +4 1/16 -2 tao 0 - 1/8 -3 to +5 1/16 =2 to +5 1/8-in.-0D, sheathed, 4 1/8 =5 to =6 in well 1/16 =3 to +6 Fig. 3.24. Thermocouple Installation on the MSRE Freeze Valve. UNCLASSIFIED PHOTO 38800 8 82 except that during rapid heating and cooling, differences of 15 to 20°F in temperature readings were noted. All the thermocouples were still func- tioning after two weeks of intermittent operation of the valve. No sig- nificant difference in the performance or durability of the l/l6-in. single-conductor and l/8-in. duplex thermocouples has been noted. Thermocouple Disconnects. A six-circuit radiation-resistant thermo- couple-disconnect assembly that was fabricated from commercially available components is shown in Fig. 3.25. This disconnect is simpler in con=- struction than the assembly previously described®® and is more compatible with MSRE requirements. The plug and jack panels were supplied by the Thermo Electric Company and are similar to their standard plug and jack panel assemblies, except for the insulating material, which is Electro- bestos. The housings are modified FS-type conduit boxes. The jack housing (top half) has a removable back plate to facilitate wiring connec- tions. Swaged-type tube fittings are utilized to support and restrain the individual metal-sheathed thermocouples. A disconnect of this type was tested 1n the remote-maintenance test facility and operated satisfactorily. Alignment of the pins during remote maintenance operations will be accomplished by means of a guide incorporated in the handling tool. 28'"MSRP Semiann. Prog. Rep. Feb. 28, 1961, " ORNL-3122, pp 61-62. neect. Fig. 3.25. 83 UNCLASSIFIED PHOTO 38838 g o ol (LT \ A Six~Circuit Radiation-Resistant Thermocouple Discon- PAGES 85 to 86 WERE INTENTIONALLY LEFT BLANK 87 4, METALLURGY Corrosion Effects of CF4 INOR-8 specimens, as reported previously,l sustained only superfi- cial attack by CF4, vapor in 500-hr tests at 1112 to 14T72°F. The presence of both oxide and fluoride reaction products were detected, however, on specimen surfaces examined under the electron microscope and diffraction camera. In order to determine whether these films had effected passiva- tion of the INOR-8 surfaces, two additional experiments were run with specimens immersed in a fuel salt of the composition LiF-BeFs-ZrF,-ThF,- UFs (70-23-5-1-1 mole %) and with CF4 vapor passed continuously through or over the fluoride salt. The two experiments were performed in 2-in.-diam isothermal INOR-8 pots at temperatures of 1112 and 1292°F, respectively. Although both were scheduled for 1000 hr of operation, depletion of the CF4 supply at the higher temperature required termination of this experiment after TO0O hr. Specimens being tested at 1112°F were removed from the vessel at 100-hr intervals. The specimens exposed to the salt underwent negligible weight changes and showed no consistent pattern of weight loss or gain; however, specimens tested in the vapor phase exhibited a slight weight increase, in accordance with previous test results. Specimens immersed in salt at 1292°F gained substantial weight (1 to 3 mg/cm?), and in some cases a metallic-appearing coating was noted. Analyses of the coating showed it to consist principally of nickel, with relatively large amounts (>5%) of molybdenum. Specimens in the vapor phase showed small, although consistent, weight losses and were covered with a greenish reaction product. Metallographic examinations of the specimens tested at 1112°F revealed no evidence of surface changes. Specimens tested at 1292°T were lightly pitted in areas not covered by the metallic coating, as shown in Fig. 4.l. Areas covered by the coating were unattacked. Some evidence of carburization could be detected metallographically; however, chemical analyses of the specimens after the test showed no changes in carbon content compared with that of as-received samples. The results of these and earlier experiments indicate that CF4 vapor is effectively nonreactive toward INOR-8 at 1112°F but that minor attack may be promoted by CF4 in the presence of fluoride salt at 1292°F. At least part of the attack at 1292°F appears to be attributable, however, to hydrolysis of the CF4 by traces of moisture in the system. This con- clusion is based on the detection of small amounts of CO and COs in the gas leaving the test vessel. 1"MSRP Semiann. Prog. Rep. Feb. 28, 1962," ORNL-3282, pp TT7-T9. 88 UNCLASSIFIED Y-46810 Fig. 4.1. Corrosion Effects of CF,; on INOR-8 at 1292°F. Welding and Brazing Studies Heat Exchanger Fabrication Tube-Joint Design and Fabrication. A final design and a welding procedure were established for the tube-to-tube sheet joints of the MSRE primary heat exchanger that incorporated modifications of the previous design® to provide tf'or inspection of the back-brazed region. The joint, shown in Fig. 4.2, was changed by flaring the weld end sufficiently so that weld rollover would not interfere with the insertion of an ultrasonic inspection probe into the tube. The welding conditions established to assure a minimum weld penetration of 42 mils are listed below: 2Tbid., pp 81-8k4. 89 UNCLASSIFIED ORNL-LR-DWG 65682R2 TUBE TREPAN GROOVE WITH /BRAZING ALLOY RING WELD SIDE/ \TREPAN (a) BEFORE BRAZING 77 //////////// 7 (L) AFTER BRAZING Fig. 4.2. Tube-to-Tube Sheet Joint Design for MSRE Heat Exchanger. Welding current 39 to 40 amp Welding speed 4o TOM Electrode diameter 3/32 in. Electrode-to-work distance 0.035 in. Electrode position 0.005 in. outside Jjoint interface Inert gas Argon A test assembly, shown in Figs. 4.3 and 4.4, was fabricated to demonstrate suitable fabrication procedures and welding techniques and to study the effect ot size on the back-brazing operation. The assembly contained 96 welded and back-brazed joints. Welding was completed without difficulty, and good flow of braze metal, indicated by filleting, occurred in all but one of the joinls. Metallographic investigation of this brazed joint revealed that contamination in the vieinity of the feeder holes had interfered with flow of the alloy, as may be seen in Fig. 4.5. To avoid a similar situation in the MSRE heat exchanger, precautionary inspection measures have been included in the fabrication procedure. Tube-Joint Inspection. Methods of inspecting tube joints were developed that are applicable to the MSRE heat exchanger. These include radiography of the weldments and an ultrasonic Lamb-wave technique for inspecting the brazed joints. Test welds were examined with both x-ray and iridium sources under a variety of conditions and with curved strips of film placed in the 20 UNCLASSIFIED PHOTO 58284 Fig. 4.3. Weld Side of Test Assembly with MSRE Heat Exchanger Type of Joint. tubes at the welds. Of those tested, the optimum radiographic conditions found for the heat exchanger weld were: X-ray energy 160 kv Film to focal distance 48 in, Tncident angle 3035 deg Time 2.5 min Film type M Four to six exposures were required to completely inspect each weld; however, the tube arrangement on the MSRE heat exchanger allows for several tube welds to be radiographed at one time, and the operation will be more economical than for the test welds. The defects observed were somewhat distorted because of the method of film placement, but, despite this, pores several mils in diameter were detected. The ultrasonic Lamb-wave technique developed for the evaluation of bonding in brazed tube Jjoints utilizes a small probe that fits inside the l/2—in.-diam tube. The probe contains two piezoelectric crystals a9l UNCLASSIFIED PHOTO 58287 2 N 3 RATORY QOAK RIDGE NATIONAL LABO U Fig. 4.4. Braze Side of Test Assermbly with MSRE Heat Exchanger Type of Joint. that serve as ultrasonic generator and receiver, respectively. The proper ultrasonic frequency and beam incident angles were determined, and a custom probe was designed and fabricated. Approximately 20 brazed joints with various degrees of bonding were evaluated using this system, and metallographic examinations are in progress to provide correlation of the test results. A mechanical device was designed that provides for uniform scanning of the probe in the tube and thus assures complete inspection coverage. This device should permit more rapid inspection and may also allow antomatic recording ol uubonded arcas. 92 UNCLASSIFIED | Y-46952 TUBE WALL Fig. 4.5. Defective Braze Caused by Metal Chips in Feeder Holes. Remote Brazing The brazing procedure and equipment described previously® for remotely fabricating INOR-8 pipe were used in the preparation of sgix prototype joints. Evaluation of these by ultrasonic techniques developed® previously indicated only minor nonbonding in five of the brazed areas. The sixth joint contained a number of unbonded areas in the upper portion of the braze. The bonding in a typical remotely brazed joint is described in Fig., 4.6. Welding of INOR-8 A mechanical properties investigation is being made of INOR-8 weld material from heats to be used in the MSRE. As part of thc preliminary studies, data were obtained on the effect of stress-relieving on stress- rupture properties. These initial tests were made using transverse weld SIbid., pp 84-86. “K. V. Coock and R. W. McClung, "Development of Ultrasonic Techniques for the Evaluation of Brazed Joints," ORNL-TM-356 (in press) . 93 UNCLASSIFIED /OUTER SURFACE ORNL-LR-DWG 69026R NOMINAL LENGTH END EFFECT OF JOINT (1in) . END EFFECT FROMFILLET .77 ), 7] I FRoM FILLET 1 o = [ . - : | - ST ! | | } L I I I I | | 1 | - - | . | : i I fin— | | | | Tt || BONDED AREA — e = = . || UNBONDED AREA = - o T e - | e =i - | | y | 2in. _’ } | | I | | | | | | I | 3in — i 1 r | | | | = 2 I =) -e | = | 5 ¢ 3} | = { = I < 4in. | = . i 4 w | i | ] e | = I 3 I [} © I S I < l S | . I / | Sin. | * | I | | i | | | S I I [ i N ! | ’ l 6in.—D| : ' - 1 i . j = ! I ' | | S g I | I i s 4 | . | { | 1 7in.—| I | I ! s i 2 i ! I ! | | - | Fig. 4.6. Results of Ultrasonic Inspection of a Remotely Brazed Pipe Joint. 9% specimens cut from a weld-containing base plate (heat No. N1-5090) and weld wire. Two samples were tested at 1300°F in the as-welded condition and two in the stress-relieved condition (2 hr at 1600°F). The data obtained are presented in Table 4,1, It is evident that the stress- relieving treatment improved the over-all properties of the weldment, probably because of a redistribution of microconstituents, such as grain-boundary carbides. Table 4.1. Elevated-Temperature Stress-Rupture Test Data on Transverse INOR-8 Weld Specimens in the As-Welded and Stress-Relieved Conditions Time to Rupture Total bStrain Sample Condition (hr) (%) As welded 113,9 %445 79.1 6.6 Stress relieved 234.,2 125 233,9 1269 Tests were also performed on welds made with the short-arc semi- automatic welding process. Tensile tests at 1300°F provided the data presented in Table 4.2. The low ductilities were probably due to defects in the weld. It is expected that improvements in the process can produce welds having properties comparable with those of welds produced by the manual, tungsten-arc process. Table 4.2. Tensile Tests at 1300°T of Transverse INOR-8 Short-Arc Weld Specimens in the As-Welded Condition Sample Tenesile Strength Yield Strength Elongation No. (psi) (psi) (%) 1 58,300 47,500 Ted 2 60, 800 4'7, 500 965 (2) 67,300 47,200 15,5 aTypica] 1300°F tensile data for manual, tungsten-arc, INOR-8 welds. 95 Mechanical Properties of INOR-8 Properties of Reactor-Quality INOR-8 The mechanical properties of typical INOR-8 material procured for MSRE construction are being determined in order to evaluate the effects of large-gquantity production and improved-quality requirements. Speci- mens for tensile and stress-rupture testing were prepared from representative INOR-8 plate, including heats N1-5055, N1-5075, and N1-5081. Tensile properties will be tested from room temperature to 1800°F, and stress-rupture properties will be determined in air at 1100, 1300, and 1500°F. Thermal Fatigue of INOR-8 The thermal-fatigue properties of TNOR-8 are being determined with methods described by Carden.> Fifteen tests were completed at maximum test temperatures in the range 1250 to 1600°F and cycle periods of 1 and 10 min. Results of these tests are presented in Fig. 4.7, which UNCLASSIFIED ORNL-LR-DWG 74157 E £ > . = . INCONEL (13C0°F MEAN) w g . l% 2 N \\ >~ ™ a 10-2 — \\ N - ——TYPE 304 STAINLESS - o —— (1300°F MEAN) S ~ =z N S E:I 5 """-%,_ \\ = .-'0"'--\. = \\\ g q’ T TR E 2 | i ‘--._‘-_‘ ? OPEN SYMBOLS # | min /cycle ~Pr- ~ [ J Q 453 |- SOLID SYMBOLS 10 min/cycle AN | 2 — o 1250-1350°F T Z 5[ a 1400-1450°F - —— n 1500 =1550°F | MAX 4 —— { 1600°F 2 104 ' - - 0 2 5 10° 2 5 10° 2 5 1% 2 5 10 M, CYCLES TO FAILURE Fig. 4.7. Thermal-Fatigue Data for INOR-8 (Heat NL-5050) in Com- parison with Inconel 'and Type 304 Stainless Steel. gives the plastic strain range (maximum internal width of stress-strain loop) versus the number of cycles to failure for LNOR-8, Inconel, and 5A. E. Carden, "Thermal Fatigue — An Analysis of the Experimental Method," ORNT.-TM-LOS (4in preparation). 96 type 304 stainless steel. The data on Inconel and stainless steel are included for comparison. The thermal-fatigue data of Fig. 4.7 indicate that Inconel and type 304 stainless steel have greater ability to sustain plastic flow than INOR-8. An analysis of the temperature difference versus cycles to failure for comparable maximum temperatures and frequencies shows, however, that INOR-8 will sustain more restrained thermal cycles and that INOR-8 requires a greater temperature difference than type 30k stainless steel or Inconel to produce failure in a comparable number- of cycles. The plastic strain per cycle is less in INOR-8 because of its lower thermal coefficient of expansion and higher elastic strength. Considerable plastic flow was observed in INOR-8 during the hold time in the 10-min per cyc¢le test. This phenvwmenovuy is being studied for hold times of up to 24 hr. Graphite Studies Evaluation of Grade TS-28) Graphite Specimens of grade TS-281 graphite were evaluated for molten salt and mercury permeation in standard screening tests, and the specimens met MSRE design requirements. This graphite was produced for the Engineering Test Loop in the size required for the MSRE and with essentially the same characteristics as those of the graphite for the MSRE. ' Exposure of TS-281 test pileces to molten salt at the conditions described in Table 4.3 resulted in an average salt permeation of 0.2% of the bulk volume. The MSRE specification allows a maximum of 0.5% permeation. Impregnation of similar specimens with mercury at the conditions described in Table 4.3 resulted in an average weight gain of 1.2%, as compared with a maximum allowable MSRE specification of 3.5%. A comparison of the permeation of TS-281 graphite with that of experimental grades of similar graphite (reported in Table L.lL) Table 4.3, Test Conditions for Standard Permeation Screening Tests Test Molten-Salt® Mercury Conditions Test Test Temperature, °T 1300 70 Test period, hr 100 20 Pressure, psig 150 470 41iF-BeF, -ThF,-UF, (67-18.5-14-0.5 mole %), 97 Table 4,4, Salt Permeation for Grade TS-281 and Similar Experimental Grades of Graphite . Size of Cross Bulk Volume Géaphlte section Permeated rade . (in.) (%) TS-281 2 1/4 x 2 1/4 0.2 CGR-X 4 (diameter) 0.02 CGB-Y 2 1/4 x 2 1/4 0.04% indicates that full-scale or commercial fabrication imposes the penalty of greater porosity. In the testing of the TS-281 specimens, the graphite bars were found to be slightly more porous and less dense toward their central axis; however, these regions were well within the requirements specified for MSRE graphite. Radiographic examinations are being made to determine: Lhe distribution of salt impregnation in CGB-Y and TS5-281 graphite. Effects of Thermal Decomposition Products of NH4F-HF on Graphite- INOR-G Systems ' A determination was made of the effects of temperature on the purging procedure® developed to remove oxygen from grephite by utilizing the decomposition products of NH4F+HF. The effects on INOR-8 were examined concurrently. Grade R-0025 graphite and INOR-8 specimens were exposed isothermally for 20 hr in Inconel containers at temperatures varying from 392 to "1300°F., Both 0.2 and 1.0 g quantities of NH4F-HF were included in the test system. The following qualitative results were obtained from this study. 1. Both 0.2 and 1.0 g of NH4F.HF effectively removed oxygen con- tamination from the graphite, except in one instance. The test made with 1.0 g at 392°F was inconclusive and is being checked. 2. For purges with 0.2 or 1.0 g of NE,F+HF there was a minimum reaction of the INOR-8 specimens at 752°F, as determined by weight gain. %3, In general there was less reaction (weight gain) of INOR-8 at all purging temperatures for the tests with 0.2 g of NH4F.HF than for tests with 1.0 g. S"MSRP Prog. Rep. April 30, 1960," ORNL-2973, p 59. 98 4, The reaction (weight gain) of INOR-8 varied in proportion to the amount of graphite present and was least in the test with no graphite. This increased attack in the presence of graphite was probably caused by oxygen products available in the graphite. Fabrication of Gds0s; and Gdo03—A1-05 Pellets A study was begun to develop the fabrication procedure necessary to make cylindrical shapes of Gd>03 and Gdz03—Alz03 mixtures. These parts are of interest as control rod elements for the MSRE and are being designed to withstand a nitrogen atmosphere contaminated with air and moisture at temperatures up to 1400°F. Al-03 is added to minimize hydrolysis and subsequent deterioration of the GdzOs. A preliminary test was made of the sintering characteristics of the materials under a single set of conditions. Pellets containing 0, 20, and 30 wt % Alo05 were prepared by dry blending, pressing, and 51nter1ng in hydrogen at 1750°F for 3/4 hr. All sintered pieces were set on molybdenum sheet. Of the compositions tested, only Gdz0s gave a sound pellet with good density. 'The density and shrinkage data are listed in Table 4.5. The mixtures of (Gds0s and Al,05 showed evidence of eutectic . melting or the formation of a low-melting-temperature compound that indicated too high a sintering temperature. Additional specimens of Gdp0s—Al505 mixtures with various green densities have been prepared and will be fired at 1600, 1650, and 1700°C to determine optimum sintering conditions. Table 4.5, Characteristics of Gd,03-A1,03 Mixtures Sintered in Hydrogen at 1750°C for 3/4 hr . . . Density Shrinkage Composition (% of theoretical) (%) Gd203 93 23 80 wb % Gda02—20 wt % Al20a 60 13 70 wt % Gd03—30 Wt % Al,0 Melted Pellets of Gds05 with green densities ranging from 2.9 to 3.8 g/cm3 were tested for determining the effects of green density on sintering characteristics. Shrinkage and bulk density data for these pellets are presented in Figs. 4.8 and 4.9. Using this information, thin wall cylinders having the dimensions required for MSRE control rods will be tabricated for proof testing. UNCLASSIFIED CRNL-LR-DWG 74158 N 25 >, DIAMETER ~ 24 B | \\ uj T LENGTH - UNCLASSIFIED g 1N N “ ORNL-LR-DWG 74159 z 22 €76 T \Q e -}) \ > N\ = g 20 [z e z 8 - i / \ o // w 44 et 16 ' Z 7.0 2.8 3.0 3.2 3.4 3.6 3.8 4.0 ® 2.8 3.0 32 3.4 3.6 3.8 4.0 GREEN BULK DENSITY (g/cm3) GREEN BULK DENSITY (g/cm3) FPig. 4.8. Sintering Shrinkage of Fig. 4.9. Bulk Density Change of Gd,04 Soaked at 1750°C in Hydrogen for ‘ Gd,03 Soaked as 1750°C in Hydrogen for 3/4 hr. 3/4 hr. 66 100 5. IN-PILE TESTS ORNL-MTR-47 Molten-Salt Irradiation Experiments Experiment ORNL-MTR-47-L The carbon tetrafluoride found in the cover gas over molten-salt fuel irradiated in graphite crucibles in assembly ORNL-MTR-47~3 is considered to have formed at the salt-graphite interfaces and to have passed through the graphite to the cover gas.l It is postulated that if the CF) had been forced to pass through molten fuel, it would have been consumed. Two ex- perimental assemblies, ORNL-MIR-47-4 and 47-5, were designed to test this hypothesis and to study the formation of CF). Six cylindrical INOR-8 capsules,g four containing a central graphitc core submerged in fuel salt and two containing graphite crucibles, were installed in assembly ORNL-MTR-47-L. The capsules were immersed in molten sodium, which served as a heat transfer medium. The primary purpose of the experiment was to determine whether CF), would exist in the cover gas over MSRE fuel irradiated in contact with completely submerged graphite. Because the fuel could not be maintained molten during reactor scrams and shutdowns, the graphite surface was intermittently in direct contact with the cover gas through cracks in the solid fuel. The crucible capsules, which contained exposed graphite in contact with fuel, provided a basis for comparison with the capsules of the 47-3 experiment and an evalualion of the effectiveness of graphite submersion in preventing the net gener- ation of CF). Also, experiment -4 provided for @ [further demonstration of the compatibility of the fuel—graphite—INOR-8 system under thermal conditions at least as severe as those expected to exist during MSRE -pperation. The temperature history of the 47-4 capsule during the irradiation period of March 15 to June b is summarized in Table 5.1. The four sub- merged-graphite capsules contained central Chromel-Alumel thermocouples. The internal temperature distribution (Fig. 5.1), obtained by use of the computer program TOSS,- indicates that the thermocouple should give, essen- tially, the graphite-to-salt interface temperature. The temperature history of the uninstrumented crucible capsules was calculated based on temperature measurements in the surrounding liquid sodium. The muxlmum.mcasured tem- perature among the four instrumented capsules was 1400 T 50 °F. The mean measured temperatures of the other three instrumented capsules were approx- imately 1380, 1370, and 1310”F, respectively:. The corrccponding caloniated INOR-8 capsule wall-to-salt interface temperatures were 1130, 1125, 1115, 1"MSRP Prog. Rep. Feb. 28, 1962," ORNL-3282, pp. 97-110. 2Tpid., pp. 110-112. | SD. Bagwell, "TOSS - An IBM 7090 Code for Computing Transient or Steady State Temperature Distributions," K-1494, December 1, 1961. 101 Table 5.1. ORNL-MIR-47-4 Capsules Temperature History of Fuel Salt in Time at Temperature (hr) Temperat?ES)Interval Submerged Graphite Core (24, 4A, 54, 6, 6A)P Graphite@d Crucible (4) Steady-state operation 0-100 390.6 100~700 44,0 700750 1203.7 750800 254,0 800850 850—200 Total 1892.3 Nonsteady-state operation 51.7 Total irradiation 1553.4 390.6 12.1 3.8 17.6 11.4 1456.9 1892.3 51.7 1553. 4 SCalculated temperature history based on temperatures meas- ured in submerged graphite capsules. bCapsule identification number. UNCLASSIFIED ORNL-LR-DWG 67930 1600 v 1c | P' . ' : /IGRA TI E CORE 1 1400 L,, FUEL SALT =N | ; "N\ | L~ CAPSLLE waLL \\ W A 1200 STV N AT N S S A -« # _ 'R 2~ 'SoDiUM w =4 —_ ¢ tooo : il“‘.__—.——--—-j“*‘SUUIUM TANK WAI I, w g B0O [— 2 wotts/g GAMMA ENERGY S I & 74 waits /cc FISSION (FUEL} o = 38,300 Btu/hr FROM -——VARIABLE GAS GAP =z SODIUM TANK = 800 ———— VERTICAL TEMPERATURE DISTRIBUTION ENVELOPE TEMPERATURE GRAPHITE 400 - MIDPLANE & EXPERIMENTAL DATA 200 “WATER JACKET L<\‘ COOLANT H»0 105°F O ‘A'. o 0.2 0.9 0.6 Q 0.2 0.4 0.6 0.8 DISTANCE FROM CAPSULE DISTANCE FROM INNER TANK CENTER LINE (in.} SURFACE (in.) Fig. 5.1. Radial Temperature Distribution of MSRE Capsules. 102 and 1110°F. The tcmperature of the surface of the INOR-8 end cap in con- tact with the fuel salt was calculated to be the same as that of the graphite-to-salt interface temperature. The area of the end cap wetted by the salt was about 12% of the total 5.6 in.2 of metal surface in contact with the salt. The accumulated time during transient operation includes only times for temperature changes of approximately 30°C or more. Of the 121 such temperature changes recorded, 60 included decreases to the solidus tem- perature of the fuel salt. The assembly was removed from the MIR, partially disassembled, and returned to ORNL, where it was completely dismantled. Postirradiation examination of the capsules is in progress. As an aid to interpretation of irradiation effects, an out-of-pile experiment was ¥iih that duplicated the temperature history, including the transients. Experiment ORNL-MTR~-47-5 The 47-5 experimental assembly was designed and constructed to permit gas sampling from two capsules during irradiation to study the CF) content, if any, of the offgas from fissioning MSRE fuel in contact with submerged graphite. These capsules (Fig. 5.2) are 1 in. in diameter and 2.25 in. long, and they contain a central CGB graphite core 1/2 in. in diameter and 1 in. long submerged approximately 0.3 in. in about 25 g of fuel. The two capsules differ only in the uranium concentration in the fuel. Additional capsules included in the assembly are designed to provide a range of oxidation-reduction levels that includes what might be both early and late stages in the life history of a reactor. This is accom- plished by altering the accessibility of chromium from the container alloy (INOR~8) to the fuel. These capsules are externally similar to the purged capsules, but the graphite corés are markedly different. A high metal-to- graphite wetted-surface-area ratio (46:1) is obtained (Fig. 5.3) by using a central core of INOR-8, except for a 0.l-in.-thick CGB graphite wafer. A low metal.-to-graphite surface area ratio (1:4) is obtained (Fig. 5.4) by using a close-fitting graphite crucible 0.7 in. in internal diameter and 1.5 in. long, with a central graphite core that is 0.34 in. in diameter and 1.25 in. long. An extreme case in the oxidization study is a capsule with no metal, that is, with graphite prepermeated with fuel so that the only fuel in the capsule is contained within the pores of an isclated graphite gere. Two small capsules of this type (Fig. 5.5) are included and will be operated at different power densities. These are 5/8 in. in diameter by 2.25 in. long and contain 0.32-in.-diameter by 1 3/h-in.-long centrally suspended AGOT graphite cores. Irradiation is scheduled to begin the week of September 17, 1962. UNCLASSIFIED ORNL-LR—DWG 71570R INOR-8 :NOR-8 COVER GAS CCVER GAS PURGE LH\E\ FURGE LIME~ SN '| N INOR-8 CAP —., Yy N v Y RTINS R A A R \ b 1 A ! PUNCTURE AREA N\ ' ) ’*—fl—HE‘-'U(“g Eg;’)ER GAS GAS SAMPLING — \ | b= | A=A INOR-8 2= s INTERFACE AREA™ k- IS80: -1 M OLTEN-SALT FUEL 3sem?y (LA 7] (25 ql A~ I T 1A INOR-8 CaN—"" [} -~ Rl - o1 . e % R CHROMEL-A_UMEL [/}~ = S CGB GRAPHITE THERMCCCUPLE —{£==7 S f//_—INTERFACE AREA /1 LR (125 cm?) & P NOR—8 CENTERING PIN N CKEL POSITIONING LUG Fig. 5.2. Svbmerged and Purged Graphite— Molten-Salt Capsule. & UNCLASSIFIED ORNL-LR-DW{ 71568R2 NICKEL POSITIONING LUGS (2)\ f —— INOR~8 THERMOCOQUPLE WELL /— INOR-8 CAP N b\, / » HELIUM COVER GAS — (3 cm”) 77L& 7 S~ INOR—8 CAN PUNCTURE AREA Np SN bl FOR GAS SAMPLING” /N /N \]|—— MOLTEN-SALT FUEL INOR-8 N R (19 ) ANTERFACE AREA N N (7 cm") Ny &Y e e Y “\ ... fl I § HELIUM GAS Gap—" NEST ™ EXH N RIS N 25 g} 7/ .SE INOR-8 5] [RXRIG Y - — P INTERFACE APEA . R N - —- 5eme N k) N \E=7/ 48 em®) < R N - n [ KK KK K] ' : \_—/é: AGOT GRAPHITE Y REEED N CGB GRAPHITE N TR IMPREGNATED WITH -~} INOR—8 CAN INTERFACE AREA—— K= ; MOLTEN-SALT FUEL N \ /_ 2 N-- -2 7 -5 (z g ! {1 em?) N[/ WA —p—"NOR-8 cCRE 2 6 i N pe— 7 Yy, \ N - - : \ CHROMEL-ALUMEL N THERMOCOUPLE ———?F FUEL RESERVOIR 0 Vs 1 c % INCH "INCH NICKEL POSITIONING LUG Fig. 5.4. Submergei Graphite-Wafer-dMolien- Fig. 5.5. Impregnated-Graphite-Molten-Salt Salt Capsule. Capsule. 701 105 Postirradiation Examination of Experimental Assembly. ORNL-MTR-47-3 Examinations of irradiated specimens from assembly ORNL-MTR-47-3, the assembly in which CF) was first detected,? provided some additional informa- tion on the corrosion products and on the distribution of radiocactivity and salt in the graphite. The possibility that much of the CF) could have been formed from the fluorine released from the frozen salt after shutdown, as found in test 47-4, placed new light on the interpretation of the results of test 47-3. The fuel mixture irradiated in this experiment was initially LiF-BeFy~ZrF) -ThF) -UF), (69.5-23-5-1-1.5 mole %). Chemical Analyses of Irradiated Fuel for Corrosion Products Corrosion~product analyses were carried out on an aliquot of the solu- tion produced during dissolution of the irradiated fuel to obtain gas samples for reducing-power analyses. The aliquot was evaporated to dryness three times after additions of methanol to remove the excess boric acid used to improve the initial dissolution. The dry residue was dissolved in 10 to 15 ml of 9 M HNO, with 2 ml of 3 M hydroxylamine to reduce any hexa- valent chromium to the trivalent state, and the thorium was removed from the solution on an anion-exchange column. The solution was then diluted to 5 N HNO,, and the uranium was removed by extraction into 20% TBP in hexane or Amscd. The iron, chromium, nickel, and molybdenum corrosion products were determined by the intensity of their spectral lines compared with an added internal standard. The readings were made with a densitometer on the photographic plate. The analyses for corrosion products, as shown in Table 5.2, gave values somewhat different from those previously reported.® The corrosion- product concentrations that indicate attack on INOR-8 coupons have been revised to the order of 100 ppm, rather than several hundred or thousand ppm, and therefore the data of Table 5.1 for iron, chromium, and nickel indicate mildly reducing conditions. They present some puzzling aspects, however, in that differences are not discernible between capsules that pro- duced widely different amounts of CFj. This can be explained if the larger amourls of CF), arose from attack of the graphite by fluorine released after freezing of the fuel. Another point of interest is the absence of molybdenum (below the 50 ppm limit of detection) in the fuel in spite of obvious attack on some of the molybdenum coupons. The molybdenum may have volatilized as MoFg, but its presence in a reducing melt would be difficult to reconcile. It would be expected, however, in an atmosphere that would produce XeF) or CFy. DNot finding the molybdenum in the fuel may indicate that it deposited on the wall of the capsule. 4"MSRP Quar. Prog. Rep. Feb. 28, 1962," ORNL-3282, p. 97. STbid., p. 110. 106 Table 5.2. Results of Corrosion-Product Analyses of Fuel Irradiated in Assembly ORNL-MTR-47-3 Corrosion Products (ppm) Capsule Sample Cr Fe Mn Mo Ni 3 1 60 310 <02 <40 70 ob 790 230 36 <50 1100 8 1 20 <130 <10 <60 50 2 120 <210 <20 <120 100 15 1 40 220 <20 <80 <230 2 £0 230 20 <100 <310 Values preceded by the less than symbol indicate the lowest detcctable limit. bThe unusually high values Tor both chromium and nickel, as well as the presence of manganese, indicate the possibility that the sample was contaminated. The compound XeFu, which was recently discovered at the Argonne National Laboratory,® is being studied. As mentioned later in connection with experiment L7-k, there is a close correlation between fluorine or evidence of fluorination in the form of CF), and a pronounced deficiency of xenon in the irradiated cover gas. This correlation and an cbserved increase in the xenon-to-krypton ratio in samples collected over a long period of time are circumstantial evidence that XeF), was formed in some capsules. Conversely, the presence of Xenon in other capsules is a possible index of the absence of fluorine, at least during the time favorable for reaction with xenon, and this must overlap part of the range in which CF) is formed. The reaction between xenon and fluorine proceeds at 400°¢. Distribution of Radioactivity and Salt in Irradiated Graphite Data on the distribution of radiocactivity and sait were obtained from cored samples of graphite parts. Patterns of radiocactivity were determined by gamma spectrometry, and the sall distribution was obtained from petrographic studies with an optical microscope. The scatter of the data preclude immediate interpretation, and the data will therefore be held for examination along with data from future similar experiments. Postirradiation Examination of Experimental Assembly ORNL-MTR-47-4 The hypothesis that CF), should react with the melt rather than accu- mulating in the cover gas in irradiated capsules containing molten fluoride ®H. H. Claasen, H. Selig, and John G. Malm, letter to editor of J. Am. Chem. Soc. dated Aug. 17, 1962, in press. 107 fuel and submerged graphite was tested further with the irradiation of four capsules, described previously,? at a flux of 1013 neutrons/cmé-sec for more than 1500 hr in the MTR. The data sought in the test, namely the partial pressure of CF) at operating conditions, were obscured in two of the capsules opened thus far by large amounts of radiolytic fluorine from the frozen fuel. Of the fluorine content in the gas phase, which was 1% of the fluoride in the fuel, some 80% was free fluorine and the remaining 20% was CFy, most of which probably was a secondary reaction product arising from attack on the graphite by the fluorine released by radiolysis after shutdown. Two smaller capsules containing molten salt in graphite crucibles were irradiated in the same assembly (ORNL-MTR-47-4) and received similar neutron dosages and heating cycles. These capsules were also described previously.? These two capsules were included in the assembly to provide conditions favorable for the accumulation of a measurable steady-state con- centration of CF)y at two power densities; one capsule contained twice as much UF), as the other capsule. The one capsule opened thus far operated at a slightly lower power density than the other because of its location. No fluorine and only the modest amount of CF) for this type of configura- tion were found. Iluorine atoms undoubledly are formed in fissioning fuel at an apprecilable rate as a result of radiolysis. The rate of formation is probably proportional to the power density. The lack of corrosion of metal or graphite indicates that rapid reverse reactions in the molten fuel consume the fluorine atoms and lower the steady-state concentration of elemental fluorine to negligible levels. All other chemical components in the operating fuel system, including moderator and container, remain at virtual thermodynamic equilibrium as long as the fissioning fuel is molten. As shown by some of the results described below, restricted mobility in the frozen crystalsuof fuel, along with the pronounced drop in tempera- tures that prevails in irradiation test capsules after reactor shutdown, may, on occasion, lead to off-equilibrium conditions in the form of ex- cessive amounts of unreacted elemental fluorine. L1n such cases, the reverse reaction mechanisms that consume fluorine are even more drastically curtailed on shutdown than is the rate of energy release that gives rise to fluorine atoms. The resulting buildup of fluorine molecules in the frozen capsule contrasts markediy with the situation at operating conditions. Also the marked contrast between various frozen samples implies that small differences in geometry or freezing history may be of controlling importance. Not all freezing liberates fluorine. Further there does not seem to have been a significant amount of fluorine formed in the 30 freezings that occurred during the in-pile exposure, probably because of an insufficient time lapse in some cases. Also there were capsules in which no fluorine was found even after the final lengthy cooling period. The gquestion of why some capsules did and some did not contain fluorine can be answered only conjecturally on the basis of current information. 108 Degcription of Capsules The capsules were constructed of l/l6-in.-wall INOR-8 tubing. The larger type contained 25 g of fuel of the nominal composition LiF-BeFy-ZrF) ~ThF), -UF), (70-23.3-5-1-0.7 mole %), which gave the analysis shown in Table 5.3; that is, 71.0-22.6-4.7-1.0-0.7 mole %. The capsules were filled by liquid transfer to a level controlled by blowing excess fuel back through the dip leg associated with the vent line.* Somewhat less than the design volume of 3.5 em3 of vapor space at temperature probably remained in the capsule, perhaps because of capillary effects. Because of contraction of the fuel, principally on freezing, there was 2.5 cm’ more vapor space at room temperature than at the operating temperature. The room~-temperature volumes, as calibrated with known volumes after puncture and gas collection, were about 4.5 cm3. Initially two concentrations of UF), were to be used in pairs of the large capsules, but breakage of a transfer line during filling, attributed to embrittlement by sulfur Lrom an unknown source, eliminated the possibility of meeting schedules with the more concentrated fuel. : Table 5.3. Chemical Analysis of Fuel in Large Capsules in ‘ Assembly ORNL-MIR-47-4 Nominal . . - Corrosion Uranium Tonic Constituents (wt %) Products (ppm) Content _ - " - — (mole %) U4t Li” 7zr4t Be Th* o Ni Cr Fe 0.7 3.89 11.6 10.1 4.8 5.8 63.6 25 24 o7 L.47 7.5 10.5 10.4 s 5 5.2 61.5 15 40 123 B A NI KTV The smaller capsules were loaded with molded ingote of about 10 g of fuel contained in graphite crucibles or liners, and in one of them the UF) concentration was increased to 1.4 mole % to provide a power density of about 160 w/cm3. The cover-gas space had a volume of about 2 cmS. There were no thermocouples in these capsules, and the fuel should not have been in contact with any metal, although traces of solvent probably distilled to the interface between the graphite liner and the INOR-8 capsule wall. Fabrication and Loading of Capsules The metal capsulc partc were cleaned by firing in hydrogen gas at lOOOOC, but a vacuum of questionable quality was inadvertently imposed while the parts were still at temperature. The resulting contamination by air caused a greenish oxide film on the INOR-8 parts, but time for another treatment had elapsed before the error was noted. The film was partially removed with steel wool prior to assembling and welding the parts in a helium-filled glove box. The graphite core in capsule 24 was cleaned by 109 heating to 2000°C in an evacuated system. The graphite parts in the other capsules were cleaned by flushing with a molten fluoride salt. The larger capsules were then connected with a filling system and heated. The molten fuel salt was run into each capsule and the excess blown back with helium pressure. The fuel salt level was monitored with a television x-ray system. Final closure was made in a glove box where the external ends of the fill tubes were crimped and welded shut. Fuel The results of ionic analyses of the bulk constituents and analyses of the corrosion products of the two fuels prepared for use in experiment ORNL-MTR-47-4 are presented above in Table 5.3. The fuel salt was pre- trecated at 750 to 800°C by sparging for 2 hr with hydrogen gas and then for 8 hr with a 5:1 hydrogen—hydrogen fluoride gas mixture, followed with a 48-hr sparge with hydrogen gas. Graghite The CGB graphite used in experiment 47-4 had a surface area of 0.71 mz/g as determined by the BET method. The other properties of this graphite are listcd below: Permeability of a 1.5-in.-0D, 6.56 X 10™% cm?® of He (STP) 0.5-in.-ID, 1.5-in.-long specimen per second Density (Beckman air pycuometer) 2.00 g/cm’ Bulk density 1.838 g/cm’ Bulk volume accessible to air 8.7% Total void volume as percentage of 19.5% bulk volume A semiquantitative spectrographic analysis of the graphite, as shown in Table 5.4, revealed only the usual low leveles of trace elements, some of which may have been introduced to the specimens during the various Table 5.4. Spectrographic Detection of Trace Elements in Graphite Used in Experimental Assembly ORNL-MI'R-47-4 Element Level Detected Element Level Detected Detected (ppm) Detected (ppm) Al 10—100 Fe 10-100 Be 1—-10 Mg 1-10 Ca 1-10 Ni 10-100 Cr 1-10 Pt 10-100 Cu 10100 Si 1—10 110 machining and handling operations. The elements analyzed for but not de- tected are listed below: Ag Hge Sr Mn As Ta Am Mo Te B Na Ti Ba - P T1 Bi Po \ Ca Pad Zn Co Ru - Sr Ga Sb Cb Ge Sn Decay Energy The [fuel salt froze within 5 min after the reactor shutdown that terminated the irradiation of assembly Y-k, caleuwlations ol Ue decoy energy in the large capsules indicated that during the first 1l-hr period following the initial 5-min freezing period about 10 w-hr of energy was dissipated, whereas during normal irradiation about 840 w was dissipated. Irradiation and Temperature Data Fxposure data for assembly 47-4 are summarized in Tables 5.5 and 5.6. Differences in flux for the various capsules were due to the flux gradlent in thc beam hole at the MER. The values for the thcrmal flux were obtained from the activation of CoP¥ in the stainless steel wires thal were used as dosimeters, and were confirmed by counting Crdl. They apply to 1553 hr during the period from March 12 to June 4, 1962. The flux was determined on both the expused and the shaded side of the capsule, but only averaged values are tabulated. ''hesé values udre higher than the design figures, and the gradient is twlce the design value. If these data arc correct, the burnup in the capsules varied over a threefold range. No explanation has been found for the discrepancy between these values and the design estimates, which are more consistent with heat transfer calculations. The temperatures were controlled from capsule 24, and the changes in the retractor position required to hold a constant temperature were not unusual. The thermocouple reading for capsule 45 appeared to drift down- ward by about 35°C during the 12-week exposure, but this was the only symptom of deviant temperature recordings. Production of Xenon and Krypton Yiclds of long-lived and staple isotopes of krypton and xenon, respec- tively, were about 4 and 21%, or 0.038 and 0.20 cm3 for 1% burnup of 1 g of U235, Thus the expected yields in capsule 6_were about 0.05 and 0.3 em3, and in capsules 24 and 36, 0.27 and 1.k em3, according to the burnups listed in Tables 5.5 and 5.6. Table 5.5. Exposure Data for Large Capsules in Assembly 47-4 Uraniuwm Contant Th2rmal-Neutron Fast-Neutron Temperature (°C) Weight of Flux® Based (>3 Mev) Flux Fuel Power Calculated Capsule Fuel on Co®° Based cn Co°8 Graphite- INOR-8-to- Density Density Burnug (g) Mole & , Activation Activation to-Sals Salt (g/em®) (w/em?) (% U?3°2) -8 (neatrons/cm?-sec) (neutrons/cm®-sec) Interface Interface x 1013 x 10*2 3¢ 25.522 0.7 0.933 750 8¢ 26.303 0.7 1.023 750 i2 25.174 0.7 0.979 2.10 2.1 6804 610 2.6134 67 5.5 24 25.3% 0.7 0.987 2.7 3.2 76cd 605 2.5412 83 7.0 36 24,.386 0.7 0.968 2.7 3.3 7.4 595 2.5863 85 7.0 45 25.598 0.7 0.996 3.35 3.8 7od 600 2.5863 117 9.7 aAverage external neutron flux. bEstimated temperatures. cUnirradiated-controls. 61Thernocouple readings vprior to termination of final irradiation cycle. Table 5.6. Exposure Data for Small Capsules in Assembly 47-4 U \ Content Thermal -Neutron Fast-Neutron Tempe rature Weight of Tanium o Flux® Based (>3 Mev) Flux o g:ntml“ Fuel Pover Calculated Capsule Fuel on Co%0 Fased on Co’ Pogion Density Density Burnu? (g} Mole % Activation Activation ‘(§C) (g/cm?) (w/cn?) (% U23°) oie g (nevtrons/cm® - sec) (nevtrons/cm?-sec) x 1013 x 1012 1b 9.381 0.7 0.365 750 3P 6.805 0.7 0.265 750 5 9.€29 1.47 D.737 895 4 10.101 1.47 J2.758 4.79 5.2 895¢ 2.5270 260 1.4 6 9.915 0.7 2.386 1.31 1.3 71L5¢ 2.5803 43 1.3 aAverage external newtrcn Tlux. bUnirradiated controls. cThermocouple readings prior to termination of final irradiation cycle. TTT 112 Gas Analyses The capsules were punctured at ORNL in the Building 4501 hot cells on a flat surface previously machined for the purpose. The screw-driven puncturing tool was sealed by a bellows and, at the flat surface, by a Neoprene O-ring. Gas escaped into a calibrated collection system that had been evacuated and leak checked. A Toeppler pump was used to transfer gases collected at low pressures. The analyses were made by mass spectrom- etry; those on samples containing fluorine were made at K-25. Data from Capsules 5 and 6. The volume of helium (from the glove box atmosphere) and argon (from the welding torch blanket) was only one half the free volume of irradiated capsule 6, as shown in Table 5.7. This could have resulted from an effective temperature of 300°C during welding. The amounts of xenon and krypton are in agreement with the calculated yield. This may mean that XeF) was absent from the capsule, or that no F, was present, or, with even less assurance, that the CF) pressure in tfiis cap- sule was representative of that to be expected ian thc graphite at oper- ating conditions. If so, the CF) pressure was around 120 mm Hg at 715°C., Table 5.7. Analyses of Cover Gas in Irradiated Capsule 6 Volume of gas removed: 3.2 cm? Measured gas space in capsule: 2.5 cm? Corrected volume of capsule gas: 1.7 cm? Eetimated gas space at operating temperature : 1.5 cm’ Quantity ' Gas — ' Volure (cm?) vol % Corrected vol % CF, 2.2 4.2 0.07 Air 47 0 1.5 He 12 23 0.38 } 1.9 A 27 51 0.86 ’ Kr 2.2 4.2 0.07 } 0.4 Xe 9.9 19 Q.32 ) Kryplou=075 and =zcnon-133 wore the anly radinactive species noted in the gas. The isotopic distribution of xenon and krypton found by mass spectrometry of gas from capsule 6 was as follows: Distribution ' Distribution Xel36 42 .06 Kr86 51.88 Xel34 31.77 Kr8? 7.51 Xel32 16.82 Kr84 26.62 Xel31 9.3 Kr83 13.99 113 The data of Table 5.8 for control capsule 5 indicate that there was very little inleakage of air during sampling of capsule 5. The volumes of helium and argon are in reasonable agreement with the values obtained for capsule 6. The high temperature of the capsule during welding seems to be a reasonable explanation for the small volume of those gases. Table 5.8. Analyses of Cover Gas in Control Capsule 5 Volume of gas collected: 1.3 cn’ Measured free volume of capsule: 2.7 cm’ Gas Quantity (vol %) Volume (cm?) He 15 0.2 A 79 1.0 } 1.2 0] 0.7 0.01 CO+N 5.3 0.007 Data from Capsule 24. When irradiated capsule 24 was punctured, 84 cmd (STP) of gas was collected. This was the amount remaining after some reaction with the apparatus. A correction has been made for the amount that reacted with the mercury in the Toeppler pump and developed a black scum that was identified by x-ray as HgpoFo. The gas was later identified as predominantly Fo. ©Since the collection system was not designed for a reactive gas, an appreciable amount of Fo must have been lost to the metallic parts of the system. The remainder reacted with glass to give S5iF) and O, and with other oxides to give Op. BStudies of the gas were hampered by the presence of an excessive amount of Tel2d activity. The free volume of capsule 24 at room temperature was 4.6 cm3, based on the pressure change when helium was expanded into the capsule from a known volume. Based on mass spectrometry ot two O.l-c:m?J samples, the 84 cm3 of gas was 5% He, 0.1% Xe, 0.4% Kr, 4% (CO + Np), 6% COp, and 17% CFy. The remainder was Op and SiF) in roughly equal amounts, matching the products of the reaction 5i0, + 2F, — LlFu pt O2 p o 2 2 Portions of gas were collected under a variety of conditions. The krypton yield listed above is in fair agreement with the calculated value, but the xcnon yield is far Loo low. Later samples, obtained after long exposure of the capsule to an evacuated 3-liter container, had 0.3 em3 of xenon and less than one-tenth that amount of krypton. This, like other cases of un- explained holdup of xenon, seems now to be assoclated with the presence of crystalline Xel). 114 The CF) content, approximately 15 cm3, would have corresponded to a partial pressure of nearly 20 atmospheres had it been present at operating temperature; this is higher by a factor of 100 than found in capsules that yielded no evidence of free fluorine. Capsule 24 was sawed open to reveal a longitudinal cross section. No damage or evidence of corrosion was visible under low magnification. Pet-~ rographic samples from random locations were in general too fine grained (crystallites less than 1 u in size) to be identified. One sample from near the graphite near the center of the capsule was clearly recognized to contain the normal phases and to have green U4’ in the TLiF.6(UF,, ThF,) solid solution. This means that the fluoride deficiency in the salt was shared randomly, as expected, for loss of fluorine from the frozen state bhbut not for loss from the liquid state. Gamma spectrometry of the samples chosen for petrographic examination revealed a preponderance of ruthenium toward the lower part of the fuel, vhen compared with the zirconium-niobium activlly Lhal is gencrolly assumed to be uniformly dispersed with the ZrF) in the fuel. Cerium activity was fairly evenly distributed with the zirconium-niobium through- out the fuel, and cesium, along with zirconium-niobium, was found on mctal surfaces exposed to the gas phase. Data from Capsule 36. The analyses of the gas from irradiated capsule 36, a duplicate of capsule 24, are more directly representative of the con- tents of the capsule because an essentially all-metal collection systcm, comprised largely of prefluorinated nickel, was used to transfer and hold the samples, and because a surplus of sample was available for mass spec- trometry. Also, the analysis was made promptly. This is pertinent - because samples allowed to sit in incompletely conditioned apparatus showed evidence of reaction of fluorine. The analytical data are presented in Table 5.9. The quantity of fluorine was too great to have been present at operating temperatures. The amount of CFj agrees well with thal found in duplicate capsule 24. Also, the xenon appears to have been held up, as if by XeF), while the krypton is present in roughly the expected yield. The gamma spectrometer was able to detect Kr85 but no other activity through the nickel walls of the sample containers used for the gas from capsule 36. Efforts to improve this analysis are under way. An attempt was made to remove the contents of capsule 36 by melting and to provide representative ground samples, all under an inert atmosphere, in a manner suitable for chemical analysis and, in particular, for a deter- mination of the reducing power that is expected as a conscquence of the loss of fluorine. An apparatus for accomplishing such a sequence of opera- tion in a hot cell worked well with capsule 3A, an unirradiated control sample, and at first appeared to have given satisfactory results with capsule 36. The melting-out operation yielded, however, only about half the capsule contents, and the melted-out salt was green. This might be explained by segregation of the reduced products in the high-melting heel. 115 Table 5.9. Analyses of Cover Gas from Irradiated Capsule 36 Volume of gas space in capsule at room temperature: 4.6 cm? Volume of gas space at operating temperature: 2.6 cm’ Volume of gas removed from capsule: 188 cm? Gas Quantity (vol %) Volume (cm®) Fa 83.5 156 CF, 10.0 19 O, 2.5 4.7 Xe 0 0 Kr 0.3 0.56 He 2.8 5.3 CO+Ne 0.1 0.19 COz 0.7 1.3 A 0 HF (2) aProbably present in small amounlts from reaction . with adsorbed water. Data from Control Capsule 3A. Capsule 3A, which was an unirradiated control for capsules 24 and 36, was subjected to thermal cycles similar to those in-pile. It was punctured and 4.5 em3 (STP) of 99.7% helium was collected. The remainder of the gas was 0.05% (CO + Ne) and 0.02% COo. The absence of argon and the presence of 1 atm of helium may be explained by the fact that final closure involved welding crimped capillary tubes. Discussion of Results Experimental evidence from previous in-pile tests lends confidence to the view that high-pressure fluorine is not generated by molten fission- ing fuel. There is no evidence of severe corrosion on metal exposed to the gas phase above fissioning fuel or on metal wet by fuel. Likewise, no severe damage was observed in similarly exposed graphite. More fluorine was found than should have been produced by the decay energy liberated during the 5 min required for cooling to essentially room temperature. When the reactor was shut down, the power produced in capsules 24 and 36 dropped from 840 w to 14 w in S min, 7 w in X hr, 1.6 w in 100 hr, and 0.19 w in 95 days (calculated from the Way-Wigner equation). Assuming that hall of the energy released during a 95-day cooling period was absorbed in the fuel, a G value of only 0.035 would have been required to produce fluorine found in capsule 36. No final choice of a theory explaining why fluorine is not always observed can be made. A difference in crystallization behavior ic a 116 possibility. The results from capsules 24 and 36 are probably quite anoma- lous in comparison with anything that has been seen before in earlier ex- periments. They point to a problem that requires attention but which does not appear to be a serious threat to the integrity of the MSRE. ' 117 6. CHEMISTRY Phase Equilibrium Studies The System LiF-Belo-7Zr¥k, The proposed simplification of the MSRE fuel mixture by the omission of ThF, led to a more detailed study of the ternary system LiF-BeFs-7ZrFg4, with a view toward prescribing suitable solvents to contain the 0.15 mole % UF4 now needed for criticality. 'T'he phase diagram of the solvent csystem did not change significantly,; but new int'ormation on liquidus contours within the system established the peritectic point involving the solid phases LiF, 6LiF-® In each crystal, the Li ions are surrounded by tetrahedra of fluoride ions, and these tetrahedra share corners and edgbs in such o manner as to form continuous sheets. The Rb* and Cs* ions are situated -between the sheets and bind them together. ' Stabilization of Fluorite Na¥YF, Solid Solution. The equilibrium iriversion temperature for Lhe pure NaYl'y crystoals inverting from the high-temperature fluorite form to the low-temperature hexagonal form is 691°C. During continuing investigations of the phase relationships in the system NaF-YFg5, it was discovered that at nearly saturating concentrations of YFs in the fluorite cubic solid solution, NaYF4-YFs, s01id solutions consisting of approximately 65 mole % YFg appear as equilibrium phases to temperatures below 4T75°C. Hence the cubic form of NaYF, is stabilized against inversion to the low-temperature form to the extent of approximately 15°C for each mole percent of dissolved YF5. Oxide Behavior in Molten Fluorides — Dehydration of LiF A marked decrease in the corrosion of grade-A nickel containers used for the HF dehydration of molten LiF was achieved by using a hydrogen atmosphere at all times when heat is applied. Previously, LiF intended for the production of ultrapure crystals was melted under a heljum atmosphere and dehydrated with a mixture of Ho, and HF. Under these conditions, both discolored and clear LiF was produced. The discolored areas contained the impurities found in grade-A nickel, such as copper, iron, manganese, "and silicon, as well as nickel, in the 10° ppn range. Material produced using a hydrogen atmosphere throughout did not contain discolored areas. No copper, iron, silicon, or nickel was detected by emission spectroscopy. Only 10 ppm Mn was found, and the total impurity was <300 ppm. “G. D. Robbins and J. H. Burns, "X-ray Diffraction Study of CsgZ7rF7," ORNL-TM-310 (August T, 1962). . SPaper to be presented at the Southeast Regional Meeting of the American Chemical Society, Gatlinburg, Tennessee, November 1962. 123 Physical Properties of Molten Salts Density of LiF-BeFs-ZrF4-UF4 The density of a proposed MSRE fuel, LiF-BeFs-ZrF,-UF, (66.85-29- 4-0.15 mole %) was determined from the bouyancy of a platinum plummet suspended in the melt. The volume of the plummet, corrected to the high temperatures, was known within *0.1%, and the change in weight of about 2.5 g could be measured to about 1 mg. The fuel was prepared in a 200-g batch and purified by treatment with 40 g of NH.HF, to remove oxide and hydrolysis products. Temperatures were measured with a calibrated platinum vs rhodium thermocouple in a well that pro- jected downward into the melt from the side at an angle. The thermo- couple was calibrated at the National Bureau of Standards. A stirrer was operated manually to ensure that the salt melt was originally uniform and was not used further. Protection from the atmosphere was maintained by flowing argon. The data for LiF-BeFs- ZrFs-UFs (66.85-29-4-0.15 mole %) fell on the straight line d =2.478 — (5 x 107%)t where t is in °C, with a precision of 0.2%. At 663°C, the fuel temperature at the outlet of the core, the density is 2.147 g/em®. Unexplained behavior which indicated that the plummet and stirrer had frozen in place was noted at 542°C, approximately 100°C above the melting point. Points obtained at higher temperatures both before and after the sticking effect was noted fell on the same straight line. Subsequent petrographic examination of the fuel disclosed no deposits of oxide, so the data were tentatively accepted as valid. Further checks to egtablish this point are in progress. Estimation of Densities of Molten Fluorides Densities of many molten fluoride mixtures, especially those of MSRE importance, can now be estimated to within 3%. The method involves, first, calculating the molar volumes of the mixture at 600 and 800°C from the equation Vnix =Z NV, 1 where N. is the mole fraction of component i and V, is the empirical molar volume of component. i given in Table 6.2. Tfie average molecular weight of the mixture is then divided by V . at 600 and 800°C to obtain the density. By assuming that den51%y is linear with tempera- ture, within the range of interest, it is a simple matter to obtain the density equation for the mixture in the form dt =a+ bt , 124 Table 6.2. Empirical Molar Volumes of Fluorides Molar Volume (em?®/mole) At 600°C - At 800°C Lip 14.0 14.7 Bng , . . 23.6 24 .4 ZrFy, . 46 49 THF, ' 47.8 49.0 UF, present up to 8 mole % 58 o 60 UF, present as 8 to 100 mole % 48 - . 50 where d, is the densgity at temperature t (in °C) and a and b are con- stants. ' | By way of illustration, the density equation of the tentative MSRE fuel mixture LiF-BeFs-ZrF,-UF, (66.85-29-4-0.15 mole %) is evaluated below: ' At 600°C, | V . = 0.6685(14.0) + 0.29(23.6) + 0.04(46) + 0.0015(58) miX _ 18.13 cm3; At 800°C, | | V. = 0.6685(14.7) + 0.29(24.4) + 0.04(49) + 0,0015(60) mix 3 = 18.95 cm’; . Molecular Weight = 0.6685(26) + 0.29(47.01)-+ 0.04(167.22) + 0.0015(311) = 38.17 g; 00 o0 . 18,13 18.95 —4 3 o %00 =800 - P~ 500 = — 4.6 x 10™* g/em®.°C . Therefore 4, = 2.381 — (k.6 x 107%)¢ in units of g/em®. Details of how Table 6.2 and the percentage reliability were obtained are reported elsewhere.® 65, Cantor, "Reactor Chem. Div. Ann. Prog. Rep. Jan. 31, 1962," ORNL-3%262, p 38. 125 Graphite Compatibility Removal of Oxide from Graphite Moderator Elements Since the unclad graphite moderator of the MSRE is a possible source of oxide contamination that could conceivably cause ZrOs to form in the fuel salt mixture, treatments that favor the removal of adsorbed gases from the moderator are included in the initial startup sequence. Some desorption is expected when the reactor core is heated to about 650°C in an atmosphere of flowing helium, and more can occur when a barren fluoride salt mixture is circulated for prenuclear testing. Accordingly, an investigation of the characteristics of gas evolution from graphite under conditions simulating these startup operations is in progress. The experimental program was divided into two parts. In one set of experiments, gas evolution from graphite blocks placed in a circulating stream of helium at 650°C is being studied. The second part of this program is concerned with the removal of adsorbed gases from graphite immersed in a molten-fluoride mixture. Gas Evolution During Helium Purging. Initially the reactor core of the MSRE will be heated to about 650°C at a maximum rate of approxi- mately 30°C/hr while preheated helium is circulated through the core at about 75 liters/sec. Accordingly, pieces of graphite (6 to 12 in. in length) were canned in a manner that provided the same cross- sectional configuration for flow as an MSRE moderator element, and gas entry and exil ports were provided at eithcr end. This assembly was mounted vertically in a tube furnace as part of the circuit of a gas-circulating loop. The loop was constructed so that portions of the gas stream emerging from the graphite block could be diverted through a MEECO Model W Electrolytic Water Analyzer and a gas-sampling system. The circulated gas stream was passed through a magnesium perchlorate drying column and a cold trap to remove moisture. Based on a uniform gas velocity through the MSRE reactor core, a flow rate of about 2.5 liters/min through the experimental loop was assumed to approximate gas purge conditions around each moderator element. A simple programming device was used to control temperatures. The data on evolution of moisture from a block of AGOT graphite and from an as-received piece of ETL graphite are presented in Figs. 6.3 and 6.4. The predominant characteristics of the gas evolution were the removal of physically absorbed moisture at temperatures below 100°C and the removal of chemisorbed moisture over the temperature interval 200 to 400°C. The additional amount removed between 400 and 600°C should be relatively inconsequential to reactor operations. In several experi- ments, AGOT graphite specimens were pretreated with water wvapor by circulating moist helium through the loop. The data from subsequent degassing, presented in Fig. 6.5, provide further illustration of the characteristic bimodal behavior. 126 UNCLASSIFIED ORNI~1 R-DWG 74358 OFF SCALE | ‘ >1000ppm GRAPHITE WEIGHT: 6954¢ 140 VOLUME : 448 cm3 GAS VOLUME : 2.96 liters (STP) E GAS FLOW RATE : 2.5 liters /min . '] 8420 : TEMPERATURE RISE : 30°C /hr = 2 2 M L T 400 b4 . 2 I o b 2 80 x = - w Q & o B0 (3] S \\;o a o ‘% 20 o ' \ & h 7 R E \( = 4] [« 0 0 100 200 300 400 500 600 700 TEMPERATURE (°C) Fig. 6.3. Removal of Moisture from AGOT Graphite by Helium Purging. UNCLASSIFIED ORNL-LR-DWG 74359 OFF SCALE = 1000 ppm GRAPHITE WEIGHT :1.122 kg 140 * VOLUME: 620 cm?3 GAS VOLUME: 2.88 liters {STP) E FLOW RATE: 2.5 liters /min o TEMPERATURE RISE : 30%C/hr - 120 - —_— = > i o T Z 100 = > = g & 80 > L o S < 60 a o % o O C:: 40 ° AO-M w . "-—"'Uo '— | T = o G 20 / d 5\0/ Q (4] o 100 200. 300 400 500 600 700 TEMPERATURE (°C) Fig. 6.4. Removal of Moisture from ETL Graphite by Helium Purging. i 127 UNCLASSIFIED ORNL-LR-DWG 74360 T 1 1 T ' OFF SCALE GRAPHITE WEIGHT: 523 g >1000 ppm VOLUME : 315 cm3 GAS VOLUME : 2.66 liters (STP) GAS FLOW RATE : 2.5 liters/min 400 TEMPERATURE RISE: 30°C/hr | | | | 100 3 0 L LOfl{/ Efi\“”*“‘““*°°“‘ ) 100 200 300 400 500 600 700 TEMPERATURE (°C} 500 WATER-VAPOR CONCENTRATION IN HELIUM (ppm) Fig. 6.5. Results of Helium Purging of AGOT Graphite to Remove Moisture That Had Been Added by Contact with Moist Helium. Data on evolution of other gaseous impurities from AGOT graphitc, as determined spectroscopically, are presented in Fig. 6.6. Since no purification train, except for moisture removal, was included in the helium circuit, these data demonstrate the accumulation and recombination of impurities in the system and thus are not directly applicable to the MSRE startup conditions. Experiments are in progress to study various gas flow rates and thermal conditions. Also the removal ¢of air and carbonaceous gases will be followed under conditions that better simulate MSRE operations, and the results will be compared with data obtained by other experi- mental techniques.” Gas Evolution by Immersion in a Molten Salt. When the primary circuit of the MSRE is filled with the barren fluoride mixture for prenuclear test operations, additional adeorbed gas may be released from the unclad graphite moderator. Also, the salt mixture is expected to remove the oxide film from INOR-8 and to permit this source of fuel contamination to be flushed from the system. A laboratory demonstration of these processes was attempted in the following manner. An experimental flush salt, LiF-BeFs (66-34 mole %), was prepared in a copper container by conventional techniques. Treatment with HF and hydrogen was continued until the analyzed oxide content of the melt was less than 200 ppm. A portion of this melt was transferred into a L. G. Overholser and J. P. Blakely, "Reactor Chem. Div. Ann. Prog. Rep. Jan. 31, 1962," ORNL-3127, p 125. 128 UNCLASSIFIED ORNL-LR-DWG 74361 5 ! : | 25 GRAPHITE WEIGHT : 695¢ . 20 VOLUME : 418 cm3 / GAS VOLUME : 2.96 liters (STP} . /o' 4 — FLOW RATE: 2.5 liters/min TEMPERATURE RISE: 30°C/hr / ( v/ CO, AND O, FOUND IN HELIUM (mole %) ~N W \ ) N N \ OO& * o MASS 28 (CO+ No) MATERIAL FOUND IN HELIUM {mole % 0 100 200 300 400 500 600 700 TEMPERATURE (°C) Fig. 6.6. Gas Evolution from AGOT Graphite by Helium Purging. closed vessel of INOR-8 alloy that contained graphite, as described below. A connecting gas manifold permitted a known volume of helium to be circulated over the melt and sampled for spectrographic analysis. Provisions were made to sample the melt periodically without serious alteration of the gas-phase constituents. Simulation of reactor conditions- was based on duplication of the ratio of salt volume to graphite volume and the ratio of graphite - surface areas. Right circular cylinders of AGOT graphite, 2, 1, and 1/2 in. in diameter, were held in a positioning rack that was submerged in the molten-salt mixture. Agitation was achieved by slowly revolving the graphite holder. ' ' The experimental data obtained thus far in this program are shown graphically in Fig. 6.7. A possible conclusion from these results is that the evolution of gas from the graphite should be of little consequence to MSRE operations. Only a few tenths of a cubic centimeter of carbon monoxide escapes into gas space per 100 em® of graphite per 100 hr, and there is 1little effect on the oxide content of the salt. The initial rapid rise in oxide content resembles, qualitatively, the expected dissolution of oxide scale from the INOR-8, but in general the values for oxide content are puzzling. Behavior of CF, in Molten Fluorides Additional attempts to measure the solubility of CF, in MSRE melts brought the total to three experiments at 600°C and two experiments at 800°C in which the solubility was apparently below the limit of detection J 129 UNCLASSIFIED ORNL-LR-DWG 74362 3.0 CONCENTRATIONS OF Op,CHa, AND H, IN GAS PHASE —1 700 WERE NEGLIGIBLE 2.5 | VOLUME OF SALT: 1371 cm? VOLUME OF GRAPHITE: 1070 cm3 —| s00 VOLUME OF GAS:™ 4.6 liters (STP) - ° 2.0 500 ® // —1 400 P 1.5 A MA%>//,. A/ T 28, ° ///// A -~ l .,/’///’ B r“ ® \\A A N ””,/’ OXIDE —! 200 Y A EVOLVED GAS FOUND tN HELIUM (mole %) OXIDE FOUND IN MOLTEN - SALT MIXTURE {(ppm) / o/ s e ® l.-—l,r 0 L—_—.l | A 0 o] 50 100 150 200 250 300 350 TIME (hr) Fig. 6.7. Removal of Adsorbed Gas from AGOT Graphite Immersed in LiF-BeF, (66-34 mole %) at 600°C. of about 10~ moles of CF, per cm® of melt per atmosphere. The experi- mental equipment, described previously,® was recalibrated with argon, and the data were found to be in agreement with eariier measurements with that gas.® Analyses by mass spectrometry of a mixture of CF. and helium that was recirculated continuously at 700 or 800°C through LiF-BeFs-ZrF,-ThF,-UF, (70-23-5-1-1 mole %) to which uranium turnings had been added indicated a slow loss of CF4 at 700°C that reached = rate of, perhaps, 4% per day at 800°C. Enough uranium had been added to convert 20% of the uranium in the melt to UFsz, but, for reasons not fully understood, analyses for determining the reducing power of the melt gave consistently negative results, that is, less than 0.1 wt % U3+. The experiment. was interrupted by trouble with a gas line before the rate of reaction at 800°C could be established with certainty. 8y. R. Grimes, N. V. Smith, and G. M. Watson, J. Phys. Chem., 62: 862 (1958). °G. M. Watson, R. B. Evans III, W. R. Grimes, and N. V. Smith, J. Chem. and Eng. Data, T(2): 285 (1962). 130 Tn static experiments at 700°C involving 15 g of LiF-BeF, (66-3h mole %) containing 1.45 wt % US" (added as UFs crystals), no evidence of reaction of CF4 with the melt was found when a one-third excess of CF4 was used in the cover gas during three trials in a 155-cm® reaction vessel. The possibility that CF, might react metathetically with oxides at 600°C was also investigated, but the results were inconclusive. In experiments now being prepared, CF4 will be brought into con- tact with a reduced melt at a graphite interface to see whether the rate of reaction will increase, and an attempt will be made to measure the hydrolysis rate of CF,4. Analytical Chemistry Determination of Oxygefl in MSRE Fuel Mixtures Work continued on the determination of oxygen in salt mixturecs by the inert-gas carbon-reduction method. A graphite crucible tapped to accept a threaded plug that, in turn, screws onto a graphite spindle is used to contain the salt. The plugged capsule facilitates the loading of the sample within the dry box and subsequent rapid weighing. The quartz tube that is used to provide an envelope for the inert gas was modified by enlarging the diameter of the central expanded section and by inserting an inner quartz shell. These modifications were intended to lengthen the useful life of the quartz tube. Standard calibration curves were obtained by using metallic tin standard samples of a known oxygen content that were supplied by the Laboratory Equipment Corporation. The oxygen content of reagent-grade LiF was assayed as a test determination. The value obtained was 250 * 40 ppm based on five determinations. The method was further tested on samples of LiF-BeFs-ZrF,-ThF,-UF, that were also analyzed by the KBrF, method'® and the "triton" activation analysis methods [0*©, (H3,n)F'®]. The measured oxygen concentrations are listed in Teble 6.3. Table 6.3. Results of Comparative Oxygen Analyses of MSRE Fuel Mixture Oxygen Content (ppm) Sample - ' No. Neutron Activation Inert Gas Fusion KBrFy Method Method Method 4285 952, 963 804, 793, 706 897, 568, 550 4282 416 - 626, 668, 597 430, 425 4987 330, 386 747, 745, 605 404, 604, 518 (¥ 131 Although the comparison is not as good as anticipated, the results are in general agreement. Rather wide spreads are observed, however, in the results of a single method. This was noted previously and is possibly the result of (1) inherent errors in the methods used when applied to fluoride salts, (2) inhomogeneity of the sample, or (3) contamination during sampling and handling in the dry, inert-gas box. The latter has been proved to be a definite cause of error in that the results vary by a factor of 5 in some cases when the sample is deliberately exposed to the atmosphere. It is not possible at this time to eliminate either of the other aforementioned possibilities of error, but efforts are continuing to ascertain the sources of error. Determination of Oxygen in Graphite Several methods were tested for the determination of oxygen in graphite. The methods used to date include (1) fluorination with KBrF,,'° (2) inert-gas fusion, (3) vacuum fusion, and (4) neutron activation with 1l4-Mev neutrons. The graphite examined was high-purity material, and the samples varied in weight from 50 to 250 mg. In the initial tests the quantity of oxygen obtained by means of the first three procedures was essentially equivalent to the representative blanks for the procedures. Based on the weight of "sample used, the oxygen concentration in the graphite was less than 50 ppm. Considerably larger samples and attendant modifications of the apparatus would be required to establish lower limits of sensitivity. For example, if l-g samples were used, the oxygen concentrations as low as 10 ppm could be deterwined. Currently, efforts are being directed toward use of the neutron- - activation technique using 1h4-Mev neutrons generated by the Cockroft- - Walton generator. This method is nondestructive and capable of a high sensitivity that depends only on the condition of the tritium target. - Oxygen (0*®) in the graphite undergoes an n,p reaction upon irradiation with high-energy neutrons to produce N1®, which has a 7.3-s half-life. The gamma activity of the N'€ is counted using a gamma spectrometer and a multichannel analyzer. Special pneumatic transfer of the sample to the counter is necessary to achieve the ultimate sensitivity. Only cursory results have been obtained thus far. It is planned to extend this technique to oxygen determinations in other materials. Homogenization of Radiocactive MSRE Fuel Samples Equipment was designed to remove and homogenize radioactive MSRE solid salt samples preparatory to analytical manipulations.'! The 10G. Goldberg, A. S. Meyer, Jr., and J. C. White, Anal. Chem., 32; 31k (1960). 1M, J. Gaitanis, C. E. Lamb, and L. T. Corbin, "Homogenization of Molten Salt Reactor Project Fuel Samples," ORNL-TM-291 (August 15, 1962) . 132 equipment includes a copper pulverizer-mixer and a copper sampling ladle that contains the solidified fuel. The sampling ladle is placed in the pulverizer-mixer, which is agitated on a mixer mill. The fuel is fractured out of the ladle, pulverized into a homogeneous powder, and transferred to a storage bottle. The homogenized fuel sample is then available for analysis. 133 7. FUEL PROCESSING Work on the detailed design of the fuel-processing system, which consists of a fluorination system for uranium recovery, was started. Equipment in addition to that described previouslyl is now included for treatment of the salt with a mixture of hydrogen and hydrogen fluoride for oxide removal. A simplified flowsheet of the system is presented in Fig. T.1. ‘ : UNCLASSIFIED ORNL-LR-DWG 73994 NaoF ABSORBERS AT 1BO°F 50, CONDENSER o PREHEATERS SAMPLER SAMPLE fl POT FUEL OR FLUSHING SALT REACTOR FROM DRAIN TANKS Hp-HF MIXTURE OR FLUORINE S0, F, L~ TO LIQUID- 1 : WASTE TANK TO OFFGAS FUEL ' CAUSTIC FILTER STORAGE SCRUBBER AND TANK STACK , N NoF TRAP AT 750°C| - Fig. 7.1. MSRE Fuel-Processing System. In the proposed system, oxide removal is accomplished by sparging the molten salt with an Hp-HF mixture. The Hp, H20, excess HF, and some volatile fission products, principally niobium, tellurium, and iodine, 1"MSRP Semiann. Prog. Rep. Feb. 28, 1962," ORNL-3282, pp 13k-5. 134 are evolved. The TS50°F NaF trap removes the bulk of the niobium. The HF-Hz0 mixture is condensed, sampled, and analyzed in order to follow the progress of the treatment. The sample pot is emptied, after sparg- ing, to the caustic tank, where iodine and tellurium are collected. After sufficient decay, the caustic is sent to the liquid-waste tank and the ORNL waste system. - Fluorination of the fuel salt volatilizes some fission products, in addition to UFg, but an over-all decontamination factor of >107 is ex- pected. Additional decontamination from ruthenium, zirconium, and nio- bium is obtained as the UFg passes through the T750°F NaF trap. The uranium is collected as UFg-+3NaF in the 180°F NaF absorbers. These are portable units that will be moved to the Volatility Pilot Plant for desorption and cold trapping. Excess fluorine will be disposed of by reacting it with SOz at LOO°F, A relatively inert gas SO F. is formed and can be safely sent through the Fiberglas filters and discharged to the stack. . 13 135 8. ENGINEERING RESEARCH ON THERMOPHYSICAL PROPERTIES OF SALT MIXTURES The mixture LisC03-NazC03-K2C0s (30-38-32 wt %) has been proposed for use in out-of-pile development studies relating to the MSRE on the basis of similarity in properties to those of the MSRE fuel salt. Use of this mixture will ease the handling problems because it is essentially noncorrosive to stainless steel without a protective atmosphere, Meas- urements of the enthalpy and viscosity of this salt are reported here. Enthalgl Experimental data on the enthalpy of the LizC03-NazC03-K=C0s (30-38- 32 wt %) salt in the liquid state are presented in Fig. 8.1. Over the temperature span 475 to 715°C the data can be represented by the equation Hy - Hypep = 18.10 + 0.413 & UNCLASSIFIED ORNL-LR-DWG 73992 350 5 i S 300 —&7 a & 2 Ve = J° = 1 o Z 250 | A 5 o] g o Al o7 < =i ) T 200 ol = _1} | I 150 300 400 500 600 700 800 TEMPERATURE (°C) Fig. g.1. Enthalpy of a LigCO3—Na2003-—K2003 (30—38—32 wt %) Mix- ture. where the enthalpy, H, is in cal/g when the temperature, t, is in °C. A linear fit was selected in view of the unusually large scatter in the data (6 = 6.1 cal/g). The derived heat capacity is 0.413 cal/g-°C. The meas- urements will be repeated with a freshly prepared sample of the salt mix- ture in the hope of reducing the experimental scatter and establishing the effect of temperature on the enthalpy and heat capacity. 136 Viscosity The viscosity of the 1i>C03-Naz2C03-Kz=CO3 (30-38-32.Wt %).mixture was determined by efflux-cup measurements.l The kinematic viscosity (v = p/p) in the temperature range 460 to 715°C can be expressed as v = 0.024 ¢*818/T 2 where v is in centistokes and T is in °K. The density of the mixture has not been measured; however, assuming ideal solution, the mixture density (based on individual component data) is estimated to be p = 2.212 - 0.00039 t , where the density, p, is in g/cms and. the temperature, t, is in °C, The absolute viséosity'of the salt, calculated using this estimated density, is shown in Fig. 8.2. 1"MSRP Quar. Prog. Rep. Jan. 31, 1959," ORNL-2684, p 65-67. * UNCLASSIFIED ORNL-LR-DWG 73993 30 T0 &0 \ 25 NN — 20 ~ Q — v - 'g L 40 N, . £ o . E. \ —15 8 - > > 30 - - = - @ N a 8 \ 4w 8 l_f,_l - > 20 |~ - —~— | 0 i e vl "800 ST (¢ 1000 100 1200 1300 TEMPERATURE {°F) : Fig. 8.2. Predicted Viscosity of a Lip,C0O3-NasC03-K5C03 (30-38-42 wt %) Mixture. _ . f AN 137 Previous reports in this series are: ORNL-2799 ORNL-2890 ORNL-2973 ORNL-3014 ORNL-31.22 ORNL-3215 ORNL-3282 Period Ending July 31, 1959 Period Ending October 31, 1959 Periods Ending January 31 and April 30, 1960 Period Ending July 31, 1960 Period Ending February 28, 1961 Period March 1 to August 31, 1961 Period Ending February 28, 1962 THIS PAGE WAS INTENTIONALLY LEFT BLANK - 139 OAK RIDGE NATIONAL LABORATORY MOLTEN SALT REACTOR PROGRAM AUGUST 1, 1962 R. B. BRIGGS, DIRECTOR ] MSRE OPERATIONS MSRE DESIGN MSRE PROCUREMENT AND CONSTRUCTION S. E. BEALL® R E. 5. BETTIS 3 R e 8 . B. MCDONALD R ENGINEERING INSTRUMENTATION AND CONTROLS P. H. HARLEY R R. L. MOORE 18 :. : ;:1‘17:!: ; S. 1. BALL 1Bc W. €. ULRICH R 2. R BROWN 8¢ PROCIIREMFNT AND FARRICATION e ; 0, H. BURGER 15¢ 4 HWESTSIX . T.M.CATE 18 C. K. MCGLOTHLAN R L. V. WILSON D. G. DAVIS 18C F. N. ROUSER R P. G. HERNDON s 1. M. TEAGUE R 8. J. JONES 18C J. W. KREWSON e C. E. STEVENSON 18C J. R. TALLACKSON 1ac R. WEIS cvs MECHANICAL DESIGN = W. M. BROWN R | CONSTRUCTION R R 8. M. WEBSTER R oy N. E. DUNWOODY* E&M W, C. GEORGE R ¢ ¢ RURTY R J. P. JACKSON E8M " R. H. JONES R ANALYSIS R. S. JACKSON E&M A R.KERR " ). P.JARVIS E&M - R ; |—— P. M. HAUBENREICH R ). W, FREELS FAM J. 0. MICHOLEOH r )R ENGEL R &b huch im L. E. PENTON R B. E. PRINCE R o C. A. ROBERTS R - W. ROBINSON R F. L. ROUSER R . G. STERLING R J. A WATTS R SCHEDULING FUEL PROCESSING — . W. GOOLSBY EaM R. B. LINDAUER . T C. H. LAMASTER E&M SITE PREPARATION G. B. CHILD E&M W, L. DERIEUX EaM N, E. DUNWOODY* E&M g' & ;'25::3" E:: INSTRUMENTATION AND CONTROLS Ve SALLEE on INSTALLATION T. A. WATSON 3] J. L. REDFORD 18C A D. sCOTT COMPONENT DEVELOPMENT SPIEWAK* ExY METALLURGY A. TABOADA K. v. COOK* W. H. CONK 4. H. DEVAN* R. G. DONNELLY" R. G. GILLILAND* J. 7. VENARD C. H. WODTKE J. L. GRIFFITH* M. . A. REDDEN* IXTTXTTTEX X FLOW MODELS ] R.J. KEDL* B. J. 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C- [§Z1; mc- [_]; PNT C- C] LTDD pBL_1;BK [1; Und|:| (J ACCELERATORS [(J ADMINISTRATIVE [ AEROSOLS [(J ANALYTICAL [0 BIBLIOGRAPHIES O BIOMED [0 BLAST EFFECTS [0 CERAMICS JZXCHEMISTRY (] COMPUTERS [0 CORROSION O cOsMIC RAYS O CRITICALITY [J CRYOGENICS O DECONTAMINATION (] DIRECT ENERGY CONVERSION (0 DOSIMETRY - 0 ECOLOGY (0 ELEMENTARY PARTICLES [0 ENGINEERING [0 EQUIPMENT J FALL-OUT ] FEED MATERIALS [0 ¥ISSION PRODUCTS (] FLUID FLOW (J GEOLOGY ] HEALTH & SAFETY ] HEAT TRANSFER O INSTRUMENTATION [J 1ISOTOPE APPLICATION [C] ISOTOPE SEPARATION C] LABORATORIES | (6-62) [J LUBRICANTS | MAGNETOHYDRODYNAMICS (] MATERIALS O MATHEMATICS [J MEETINGS %@ETALLURGY ] METEOROLOGY [J MODERATORS [] NUCLEAR EXPLOSIONS (] PARTICLE PHYSICS (] PHYSICS E“PLASMA PHYSICS [0 PLOWSHARE ' 4 qu PROGRESS REPORTS ] PROPULSION SYSTEMS O RADIOBIOLOGY (] RADIOACTIVE WASTE J RADIATION EFFECTS J RAMJETS [CJ] RARE EARTHS '] RAW MATERIALS [0 REACTOR CONTROL ] REACTOR COOLING ] REACTOR ECONOMICS @ REACTOR ENGINEERING [J] REACTOR FUELS [J REACTOR HAZARDS [fignEACTOR MATERIATS (J REACTOR PHYSICS (O REACTORS J REACTORS—~AHR - ] REACTORS-BWR REACTORS— FFR ] REACTORS—-FSR [J REACTORS—GCR [[] REACTORS—HWR [J REACTORS-LMCR [ REACTORS—LMFR [J REACTORS—-OMR [ REACTORS—PWR (J REACTORS—SGR [ REACTORS—AIR [J REACTORS—CENTRAL [(J REACTORS-DUAL PURPOSE [] REACTORS—MARITIME [ REACTORS—NAVAL (] REACTORS—PACKAGE [(CJ REACTORS—-POWER _ 0 REACTORS-PRODUCTION (J REACTORS—PROPULSION [ REACTORS-RESEARCH [(J REACTORS—SPACE [J REACTORS—TEST [J REFRACTORIES [J ROCKETS [ SATELLITES (] SEPARATION [C] SHIELDING ] STRUCTURES (J THERMONUCLEAR 7 TRANSLATIONS (] WEAPONRY O USAEC Divitlon of Technicol Information Exterlon, Ock Kidge, Tenneuse Ly W W HMuoWLwWwLuw SYoBEIG S N~ R Sl HCEN S LSNP OHNEOONEEEADEOREQ WEEEQPEPNEONUPEETR S . 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