| e (j.] b dote R o T T A g ALy aw ; o e L LA . SN . e oo v T ' * o - L Ay s e . s l; [ J Ll TS ¥ ey A e ey P L ) i ol - R A I MARTINMARIETTA ENERGY SYSTEMS LiIBRAR R oRNL 2551 Reactors—~Power T1D-4500 (13th ed., Rev.) /I\ 3 445k D350L1LEL 9 February 15, 1958 . Ve ., "‘ T L Cq 109 ,,,,, REFEiRE oeiion B oStk e 4 e LM AR Y TR e MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JUNE 30, 1958 ¥ you wish scmeohé é!se to see Hhi document, send in name wflh doc LHY- T hbrory wnli nrronge “, OAK RIDGE NATIONAL LABORATORY operated by UNION CARBIDE CORPO'RATION for the U.S§. ATOMIC ENERGY COMMISSION Office of Technical Services U. 5, Department of Commerce Printed in USA, Price Mcenfs. Available from the ‘} | Washington 25, D. C. \ LEGAL NOTICE l This report wos prepared as an account of Government sponsored work. Neither the United States, | nor the Commission, nor ony person acting on behalf of the Commissiom: \ A, Makes any warranty or representation, express or implied, with respect to the accuracy, comgleteness, or usefulness of the information contained in this report, or thot the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or i | B. Assumes any lichilities with respect to the use of, or for damayges resulting from the use of ony 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 i contracter of the Commission to the extent that such employee or contractor prepares, handles ' or distributes, or provides cccess to, any information pursuant to his employment or centract | with the Commission. | | ] ORNL-2551 Reactors~Power TiD-4500 (13th ed., Rev.) February 15, 1958 Contract No. W-7405-eng-26 MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT For Period Ending June 30, 1958 H. G. MacPherson, Program Director DATE ISSUED Fad SEP 1 G958 OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the ARTIN MARIETTAENERCY S8 e A 3 wy5k 0350616 1 TEMS LIBRARIES 0, KA i M S MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT SUMMARY PART 1. REACTOR DESIGN STUDIES critical inventories required for reactors fueled Desi di with U235, In comparison with corresponding 1.1. Design Studies U237 reactors, the critical inventories of plutonium- A conceptual design of a power reactor, desig- nated the ‘‘interim design reactor,’”’ was prepared. This two-region, homogeneous reactor uses molten salt fuel which is circulated by a single fuel pump located at the top of the core. The net power output is 260 Mw, and the reactor has a thermal capacity of 640 Mw, The total cost of the power produced by this reactor is estimated to be 8.9 mills/kwhr, of which 2.5 mills/kwhr is fuel cost. The core of the reactor was designed to have a volume approximating that of an 8-ft-dia sphere, and the shapes of sections taken through the walls can be expressed as simple algebraic functions. This latter criterion assures smooth shapes, easily calculated volumes, and dimensional reproducibility. An analysis of the weights of various portions of the system was made to aid in construction planning and cost estimating. It was estimated that the empty equipment would weigh 684,000 Ib, the fuel salt would weigh 102,358 [b, the blanket salt would weigh 194,769 |b, and the sodium would weigh 100,188 |b. A detailed evaluation of the fabricability of the reactor was made, and it was established that of the pressure-vessel industry could be used. It is recommended that the initial reactor be fabricated entirely in one shop so that tolerances need not be so rigid for fitup of parts as they would be if parts were supplied by various vendors, Design work was compieted on the fuel pump proposed for the interim design reactor. The motor and pump rotating assembly are completely replaceable remotely as a cartridge unit. The bearing is a salt-lubricated orifice-com- hydrostatic bearing, and the upper bearing is a hemispherical orifice-compensated hydrostatic type that uses the pressurized helium purge supply for lubrication and support. conventional practices lower pensated Nuclear calculations were continued, and some were obtained for reactors fueled with The initial regeneration ratios 1.03 at critical inventories of U233 that were much less than the results U233 agnd with plutonium, obtained ranged up to fueled reactors were about one-third as great. The required plutonium concentration was found to be well below the solubility limit in salts of interest. A comparative study of the various gases that might be used as cover gases for molten-salt reactor systems revealed that helium and argon are, at present, the only suitable gases. Helium is less expensive than argon, but it is not available outside the United States, Test rigs were designed for experimental studies of three types of bearings. These bearings, which are designed for molten-salt application, are (1) a hydrodynamic bearing with conventional and sleeve, (2) o hydrostatic, orifice- compensated bearing mounted on the hub of the impeller with stationary pockets supplied with high-pressure molten salts, and (3) a hydrostatic bearing with rotating pockets, journal 1.2, Component Development and Testing Development tests of salt-lubricated bearings are under way with an INOR-8 journal and sleeve being tested in the hydrodynamic bearing test rig. This initial bearing failed when the thrust load was increased from 100 to 125 Ib with the shaft operating at 1200 rpm and the molten salt at a temperature of 1200°F, Other journal and sleeve materials are being studied, A rotating-pocket hydrostatic bearing was fabricated for testing. Tests of conventional organic-liquid-lubricated A sump pump in which the lubricant is Dowtherm A has operated satis- factorily for 1620 hr with a molten sailt as the pumped fluid at a temperature of 1200°F. The shaft is 2600 rpm. The oil-lubricated pump rotary assembly that is being operated in a bearings were continued. speed gamma-irradiation field at the MTR has accumu- lated a gamma-ray dose of 5.9 x 107 r. The bulk of the Gulfcrest 34 lubricant is external to the radiation field, and thus the exposure of the oil has been only 108 r, Calculations were made of the conditions re- quired for operation of a gas-lubricated bearing. ihi A small hemispherical bearing is to be used to test the validity of the calculations., Test operation of a NaK pump with a labyrinth and split-purge seal arrangement was terminated when the passages to the labyrinth became con- stricted. Material that contained carbon was found to have blocked the passage. Face surfaces of the lower seal, which had operated 3386 hr, were free of NaK and were in good operating condition. A bellows-mounted seal being subjected to an endurance test in a NaK pump continued to operate Over 5100 hr of operation has Since there are no elastomers in satisfactorily. accumulated. this type of seal, it may be suitable for operation in a radiation field. Components of electric motors are being tested for service at high temperatures in a radiation field, since the successful deveiopment of salt- lubricated bearings would make totally submerged canned-rotor pumps applicable to molten-salt systems. Investigations are under way of high permeability magnet steels, insulating materials, and current-carrying conductors. Tests were made to determine the length of time required to bring a typical fused-salt piping system to a temperature of 900°F by preheating a portion of the system with electric-furnace elements and transmitting the heat to the remainder of the system by forced circulation of helium With about 49% of the surface heated, ¢ hr and a minimum helium pressure of 98 psig were required to reach 900°F. Screening tests of mechanical joints for remote separation of system components were continued. within the piping. None of the joints tested have given any evidence of leakage during tests with molten salts, Joints are being assembled for tests with sodium, Experimental remote maintenance work on a NaK pump was completed, and the results indicated the feasibility of remote maintenance work on accessible components of a molten-salt system, A three-dimensional viewing system is being developed. High-quality welds were produced in preliminary emphasized the need for a good viewing system, An engineering layout was prepared for a remote remote welding experiments, which maintenance demonstration facility, and detailing and fabrication of the components are under way. Combination heater and insulation units designed for remote application and removal were fabricated for testing. Commercially available expansion joints were ordered for testing. If expansion joints can be used in fuel and coolant circuit piping, the extra space and fluid inventories required for thermal expansion loops could be avoided. Processing of the data obtained from a heat transfer coefficient test in which fuel 130 (LiF- Ber-UF4, 62-37-1 mole %) was circulated in an available heat exchanger test facility was completed. The data are presented in comparison with similar data for other fluids. A molten- salt-to-air radiator was designed and is being fabricated for a molten-salt-to-molten-salt heat transfer test, Operation of forced-circulation and thermal- convection corrosion-testing loops was continued, Improvements are being made to the test stands to assure operational reliability. Out-of-pile tests of components of the in-pile loop being prepared for insertion in the MTR were continved, Difficulties with gravity-filling of the system because of the high surface tension of fuel 130 have delayed the work, New filling techniques are being developed. 1.3. Engineering Research Preliminary values were obtained for the vis- cosities of the salt mixtures NaF-Ber-UF4 (53-46-1 mole %), LiF-BeF -UF , (53-46-1 mole %), and LiF-Ber-UF4 (62-37-1 mole %) over the temperature range 500 to 900°C. The enthalpy and heat capacity of the mixture LiF-BeF,-UF, (62-37-1 mole %) were established in the temper- ature range 100 to 800°C. Studies were initiated of the thermal conductivity, surface tension, and thermal expansion of the beryllium-containing fluoride salts, Initial measurements of the heat transfer coefficient for LiF-Ber-UF4 (53-46-1 mole %) flowing through a heated Inconel tube have indicated that this salt behaves, with respect to heat transfer, in the same manner as ordinary fluids, Hydrodynamic studies were continued with small- scale glass models of proposed reactor cores, In the straight-through flow model it was found that the inlet high-velocity flow essentially short- circuited the core and passed directly from the entrance to the exit without appreciable spreading. The remainder of the core was filled with slowly rotating fluid that had extremely low velocities along the sphere wall. The concentric system with annular inlet flow exhibited a number of peculiarities which can be associated with the shortness of the annulus. The main fiow was down along the sphere surface located 180 deg from the inlet elbow and up along the back surface at the 0-deg position. in the equatorial plane at the center of this fiow. Extension of the central pipe in the concentric A tapering vortex existed pipe-entrance system showed increased velocities at the bottom of the sphere. 1.4. Instrumentation and Controls Inconel-sheathed Chromel-Alumel thermocouples with magnesium oxide insulation and hot-junction closure welds made by the Heliarc welding process are being tested for endurance and stability., In 10,000 hr of exposure to sodium at 1500°F, only two of 38 thermocouples have failed because of weld closure deficiencies. Drifts from initial temperature readings are within £0.75%. Test facilities were prepared for investigating the suitability for molten-salt reactor service of the resistance-type fuel level indicator, Design work is under way on modifications required to improve a commercial mechanism for use in switching low- level transducer signals. Various types of pressure transducer are being evaluated for molten-salt service, 1.5. Advanced Reactor Studies A conceptual design was prepared of a 5-Mw experimental reactor in which molten salt fuel would be circulated by thermal convection, This simple, reliable system could be converted to a 50-Mw pilot plant by adding a fuel pump and in- creasing the capacity of the heat dump. The 5-Mw reactor could be constructed of components already developed. It would demonstrate the feasibility of continuous operation of a molten- salt reactor, provide in-pile corrosion data, and serve as a mockup to develop and demonstrate maintenance procedures, The 50-Mw system would be sufficiently similar to a large- scale power-producing plant to lead directly to design and construction of a large power plant, remote The possibility of a thermal-convection reactor of approximately 600-Mw thermal output was also investigated. |t was found that a fuel inventory of 1775 13 would be required, which is to be compared with the 530 ft° estimated for a reactor system in which the fuel is circulated by a pump. The heat exchange equipment that would be required for gas cooling of a molten-sait reactor was studied, Helium, steam, and hydrogen were the gases considered. For a given set of con- ditions, hydrogen was the most effective, but helium and steam were reasonably comparable. An optimization of the size of the tubing to be used in the heat exchanger gave a value of 0,5 in. Larger tube diameters led to excessive tube lengths, and smaller diameters led to large numbers of tubes in the matrix, PART 2, MATERIALS STUDIES 2.1, Metallurgy Metallurgical examinations and corrosion evalu- ations were made of specimens from several Inconel and INOR-8 thermal-convection loops in which various fluoride mixtures were circulated. One of the Inconel loops, which were operated at a maximum hot-leg temperature of 1250°F in order to determine the corrosive effects of various fluorides under MSR temperature conditions, gave results which contradict the corrosion postulate that ThF4 should effecta lower corrosion rate than comparable additions of UF ,. No explanation for the increased attack by the ThF ,-containing mixture is readily available, INOR-8 thermal-convection loops that were operated under MSR temperature conditions for 1000 hr with various fluoride mixtures showed no attack, and one INOR-8 loop that was operated for more than 6300 hr was found to have widely that ranged to a maximum depth of 0.75 mil at the hottest point. The initial results for INOR-8 loops are favorable. Seven scattered subsurface voids Specimens for studies of the effect of carbu- rization on the mechanical properties of Inconel and INOR-8 were prepared by exposure in a sodium- graphite system, since carburization takes place slowly if at all in a molten-salt system, Control specimens were given the same heat treatment in an argon atmosphere, INOR-8 was found to have been more heavily carburized than Inconel by the sodium-graphite system. Tensile tests showed that carburization increased the tensile strength and yield strength of Inconel and reduced its ductility, The [NOR-8 specimens were found to have a lowered tensile strength, slightly in- creased and greatly reduced yield strength, ductility, Studies of INOR-8 and Inconel! in salt- graphite systems are being planned. Preliminary tests have shown brazing alloys with high gold and silver contents to have promising Therefore long-term corrosion data are to be obtained by insertion of samples in the hot legs of thermal- convection loops. corrosion resistance to molten salts. Good correlation has been found among tensile property data for INOR-8 obtained at ORNL, Haynes Stellite Company, and Battelle Memorial Institute, Data are now available on yield strength, tensile strength, ductility, relaxation, and Young’s modulus as functions of temperature. Preliminary creep data have been obtained, and extensive creep tests are under way. An investigation of the influence of composition variations on the creep-rupture strength and the microstructure of INOR-type alloys was completed, A general consideration of all the data obtained favorably supports the composition selected for the alloy INOR-8, Embrittlement studies have indicated that aging in the temperature range from 1000 to 1400°F has no significant effect on the properties of INOR-8, Five air-melted heats of INOR-8 were prepared by Westinghouse, and about 20,000 [b of finished products will be supplied from these heats. The first shipment of seamless jubing was received, and it was found to be of excellent quality. Experimental studies of bearing materials are under way. Flame-sprayed INOR-8 coatings were successfully bonded to INOR-8 journals, Similar molybdenum coatings cracked severely and sepa- rated from the INOR-8 wupon thermal cycling. Molybdenum rods sprayed with molybdenum are now being tested. Additional welding studies have further indi- cated that the weldability of INOR-8 is satis- Sound welds can be made, and the weld characteristics of the material are factory. deposition comparable with those of Inconel or the stainless steels, Tests of all-weld-metal specimens of INOR-8 and Inconel have shown that INOR-8 has a slightly higher ultimate tensile strength than that of Inconel, a significantly higher yield strength, and a markedly lower high-temperature ductility, A weld made in a 10-in.-dia Inconel pipe with a 5/B-in. wall by a semiremote welding process being developed for the PAR project by Westinghouse was examined. Radiographic examination showed vi the weld to be completely sound; there was no evidence of porosity. Examinations of the cast-metal seals of two flanged joints that were tested with molten salts It was found that slight oxidation had impeded wetting. The joint with a silver-copper alloy seal appeared to be less subject to non- wetting than the joint with a pure silver seal. Studies are under way of an internal tube weld- ing procedure being developed by the Griscom- Russell Company that would be applicable to the attachment of tubes to tube sheets in heat ex- changers if back-brazing were impractical, Such a procedure may be applicable to the fabrication of the large heat exchangers that will be required for molten-salt reactors. were made, 2.2. Radiation Damage Apparatus is being assembled for in-pile tests of the corrosion of INOR-8 by the molten salts of interest. An electrically heated mockup of a loop for operation in the LITR is nearing com- pletion. Parts for the in-pile model have been fabricated. Final examination of an Inconel loop that circu- lated o molten salt in a vertical hole in the LITR showed the corrosion to be the same as that which would have been expected in the absence of radiation. A new fuel salt sampling method in which the salt is melted out was found to be as satisfactory as the previous method of drilling into the salt to obtain a sample. Further preparations were made for the instal- lation and operation of a forced-circulation loop in an ORR facility. Studies of bearings for use in the loop pump are under way, and the motor of the used in previous in-pile loops is being pump redesigned. Inconel capsules for testing the stability of graphite in contact with molten-salt fuel were shipped to the MTR for irradiation ot 1250°F. INOR-8 capsules were fabricated for similar tests, 2,3, Chemistry Phase equilibrium studies of LiF-BeF, systems and/or ThF, were continued. A mol ten-salt breeder-reactor fuel with a liquidus temperature of 440 + 5°C and with no more than 36 mole % BeF, is available in the LiF-BeF,- ThF -UF , system. Studies of the LiF-BeF ,-ThF, containing UF, system have shown the existence of three eutectics with temperatures in the range 360 to 429°C, Studies of the LiF-ThF ,-UF, system have shown, as was expected, extensive formation of solid solution, Additional data were obtained on the solubility of PuF, in alkali fluoride=beryllium fluoride mixtures, The data indicate that the solubility of PuF, in LiF-BeF, mixtures is at a minimum for mixtures containing about 63 mole % LiF and that it is at a minimum in the NaF-BeF, system for mixtures containing about 57 mole % NaF. Data for the solubility of PuF, in an LiF-BeF, (63-37 mole %) mixture containing 1 mole % ThF, indicate that the addition of ThF, does not appreciably affect the solubility of PuF, in this solvent, Further experimental measurements were made of the solubilities of the noble gases in molten salt mixtures, Data were obtained for the solubility of argon in NaF-KF-LiF (11.5-42-46.5 mole %) at 600, 700, and 800°C and of helium in LiF--Bel:2 (64-36 mole %) at 500 to 800°C., The trends of the data were the same as those previously observed with mixtures containing ZrF,. [t has been demonstrated that the solubilities of HF in LiF-BeF, and in NaF-ZrF, mixtures are about the same when the alkali fluoride content is low, As the alkali fluoride content is increased, how- ever, the solubilities of HF in the two mixtures differ markedly. Studies of the solubilities of fission-product fluorides in molten alkali fluoride—~beryllium fluoride systems were continued. Data were obtained for the solubility of Ce F, over the temper- ature range of 450 to 700°C for Lil:-BeF2 and NaoF-BeF, mixtures containing 50 to 70 mole % alkali fluoride. It was found that the solubility passed through a minimum at about 62 to 63 mole % alkali fluoride in both solvents, The chemical reactions of oxides with fluorides in LiF-KF are being studied in an investigation of the chemical separation of solutes in fluoride mixtures by selective precipitation as oxides, The characteristics of BeO as a precipitating agent are being studied. The activity coefficients of NiF, dissolved in a molten mixture of LiF-BeF. (62-38 mole %) are being determined. Data obtained at 600°C gave calculated activity coefficients of 2347 and 515 with respect to the solid and liquid standard states, respectively. It is known, however, that the assumed melting point for NiF, is in doubt. The solubility of NiF, in LiF-BeF, (61-39 mole %) was measured and was found to be independent of the amount of NiF, added. A series of experiments for rechecking data obtained at high temperatures on the diffusion co- efficients for chromium in nickel-base alloys and to extend the data to temperatures below 600°C is being planned. A depletion method is to be used to check the high-temperature data, and a constant-potential method will be used for the low-temperature experiments, A study of the vapor pressures of the CsF-BeF, system was made in order to obtain information on the effect of composition on the thermodynamic activities in fuel mixtures containing BeF.,. Deviations from ideal behavior were observed that were strongly dependent on the size of the alkali cation, Studies of fused chlorides as heat transfer fluids were initiated. The mixtures KCl-ZnCl,, LiCl- ZnCl,, ond LiCl-RbCl are being investigated. Experiments are under way to study the satu- ration of graphite with an inert salt whose melting point is somewhat higher than proposed reactor fuel temperatures as a possible method for pre- venting the graphite from absorbing molten-salt fuel., Graphite rods that were soaked in and completely penetrated by an LiF-MgF., mixture arenow soaking in LiI=-Bel:2-Ul:4 (62-37-1 mole %) at 1200°F, Alteration of a production facility to provide for the large-scale processing containing materials was of beryllium- nearly completed. vii CONTENTS SUMMARY ettt e e s et s r et b et et se e et e em e et be bR b r s e bbbt b 1.1. 1.2. 1.3. PART 1. REACTOR DESIGN STUDIES DESIGN STUDIES ..ottt s et bt et o sd e ne e nb b e e eee Interim Design Reaetor. .ot e s st e be e s sb e et sa bbb e s Reactor Core Configurafion ........ccovcerieriirerieiercercnireceeceee e ettt bbb easser et s b ebasneas Weight Analysis of Reactor System .......ccoociieiiiriiiiin et s e Evaluation of Fabricability of Reactor Vessel ..o Outer Blanket Shell ...t ettt ettt INNEr €ore Shell ... ettt sttt et st s bbb er e et sas bbb Blanket System Pump Housing ..o Fuel Pump Design ...t et et ettt ettt et st aaas NUClEar CalcUltions ....oieii et e et ereee e st ee e e aea s e snaseese et e gessensan smseneentanns Modifications of Oracle Program Sorghum for Calculational Analyses of Molten-Salt REACIOrs ........cciiiiiiiiiiiece et e et rans sreaa et earesan e et eebanean Analyses of Reactors Fueled with U233 oot Analyses of Reactors Fueled with Plutonium ..o Argon as a Protective Atmosphere for Molten Salts .. .....cccooiniiiiiiciii e Bearing Tester Designs ...ttt e cme e e st e b e s e Hydrodynamic Bearing Tester ...t e Hydrostatic Rotating-Pocket Bearing Tester ..o e Hydrostatic Stationary-Pocket Bearing Tester. ... COMPONENT DEVELOPMENT AND TESTING ....cocooiieiiiiriirecrecrect et et Fuel Pump Development ...t sttt eae s sbe st e n e et ene e Development Tests of Salt-Lubricated Bearings .......cccooiiimimineciniie e Development Tests of Conventional Bearings .........cc.coooiiiiiiiiiiiiic e Gas-Lubricated Bearing Studies ..........cooooiiiiiic et sttt MEChaniEal Seals oo et ertr b e e a s a et eresresenaaene snereenen Radiation-Resistant Electric Motors for Use at High Temperatures...........c.oooeiiiiiiiiiiiiienn. Piping Preheating Tests ..ot st bbb bbb Development of Techniques for Remote Maintenance of the Reactor System.......coccoviinivirennn, Mechanical Joint Development .. ... et a e e e Remote Manipulation Techniques ... ..o et s Remote Maintenance Demonstration Facility ..o e Heater-Insulation Unit Development .......ccoccoiiiiiii et st Evaluation of Expansion Joints for Molten-Salt Reactor Systems ..o, Heat Transfer Coefficient Measurement ...........ccoooiiiiiiiiis i et e Design, Construction, and Operation of Materials Testing Loops .....cccooivieiiereiviiicecec Forced-Circulation LLoops . .. oottt ra e et n s e eee b e en e beanae s P il LL00PS oottt bbb b r e r e be bt e rean e eheebben st eeabeas ENGINEERING RESEARCH .ottt s e e Physical Property Measurements ...ttt e ettt s et st sr e e VS OSIY o oieiieiei ettt ete e ket e h e e h oo ekt s e sr st abes bt s ehe 2R At e e et aen sme e ee et e reerae e e e rereere Thermal Conductivity ..ottt ettt e e e e ere e se it ste et ebbr e eaes —_— D0 O NNONON O W W 1.4. 1.5. 2.1. 2.2. Enthalpy and Heat Capaeity .....ooeioieriieicecce e et eeeer ee s ev e e s e v s e e es s s e esanssans 38 SUPFACE TERSTON ittt ettt et ee s s e es e e eee s ee e aenens 39 TREPMA] EXPANSION .....ovoeceiii ettt et e e ee e e ee e ee e e e e e e e s s et e er et eres e e eesee e 39 Hydrodynamic Studies of MSR Core ... ..ot eee e e et eenesseeeeee e, 39 Molten-Salt Heat Transfer STUAIEs ..ot s e s et e aetesesene e s nesn s 43 INSTRUMENTATION AND CONTROLS ..ottt eeees e eea e evns e s eesena s eeeeeenes 48 Endurance and Stability Tests of Sheathed Thermocouples ..........cccouoeomieioreoeeeeeee oo, 48 Resistance-Type Fuel Level INdicator . ..ot eeeees e e s e eee s ee s 48 SCANMNING SWITERES (.ot sttt et s s s e ettt e e e s se et eeaeeraene 48 PressUre TraNSAUCErS ..ottt ettt et s st ee s ees s e et e se st et s e nraeseeeens 48 ADVANCED REACTOR DESIGN STUDIES ..ottt sttt sttt et e 49 An Experimental 5-Mw Thermal-Convection Reactor ......ccueeeeeieicieeeeieeeee e et e eaeeenee e 49 A 600-Mw Thermal-Convection ReaCtOr .........c.coiiieiuiiiiecteieeieeeeeee ettt e 51 Gas-Cooled Molten-Salt Heat EXChan@er ..............ocoooviiiiieioeee e e eer e ean e 52 PART 2. MATERIALS STUDIES MET ALLURGY ottt et et ettt e e eea et e e s eeere et eeeeeereensenenas 57 Dynamic Corrosion StUdI@S ..ottt et st eeen et se et er e e ee s e et ereeenn e 57 Inconel Thermal-Convection Loop Tests ..ot er e 57 INOR-8 Thermal-Convection Loop TestS ..ooeoiiioieeoe oo eeeeeeeteees e e oot eees s e aees e 58 General Corrosion SHUGIES ...t et et esns e senenen et esaeannes 59 Carburization of Inconel and INOR-8 ...ttt e e n e 59 Brazing Alloys in Contact with Molten Salts ..o, 62 Mechanical Properties of INOR-8 ..ot e 64 B rICAHION SHUAIES (it et ettt et s et saet ettt st et e sn e rer et are e 66 Influence of Composition on Properties of INOR-8 ... . e 66 High-Temperature Stability of INOR-8 .......... ..ottt et 67 Status of Production of INOR-8 (Westinghouse Subcontract 1067) ...ooovviviiieiiivn e 68 Status of Production of Seamless Tubing (Superior Tube Company Subcontract 1112) ............ 68 Bearing Materials ...t e ettt b e eb et e e e 69 Welding and Brazing StUdies ...ttt v e st st e r et e 69 Weldability Evaluations .. ..o oiiuieiceee et ettt ee e e et en et et en et ese s sn et asbab e tnaneras 69 REMOte Walding ..o vttt ettt e e tbeeresre e e s sstsasssaasaseesseaasssaasatansserbeaenbesataeeaniras 70 Joint Development ... e et b et e e s s e eneee et 71 Component Fabrication ....o..iiccieiiiieieeeiis e eereees st e e et sv e st eeab e s srmeseebee s st e e seteensseaastn b baeesnnesess 72 Material and Component INSPECHION .. ..oiiiciiiis ettt et e bt eb st em et 76 RADIATION DAMAGE .ottt ettt et seesre sbe b e e saesiesen e e sas e saes 78 In-Pile Dynamic Corrosion Tests . e e s 78 INOR-8 Thermal-Convection Loop Assembly for Operation in the LITR ..o 78 LITR Forced-Circulation Loop Examination .......c.ccceiiiiiiiiiiini s e 79 ORR Forced-Circulation Loop Development ............coocviiiiiiiiiivniiiiie et 81 In-Pile Static Corrosion Tests ... serre s ete e ste e s ere e s r s see e ebes b bban o 81 2.3, CHEMISTRY Lottt bbbt hbe et eaeee e ee e raea et mteste et e s es s eesseeeeenene e Phase Equilibrium STUies ...ttt ettt e a et eee e et v e e s et e e esesseaseeeeanes Systems Containing UF ; and/or ThF , ..o Solubility of PuF, in Alkali Fluoride~Beryllium Fluoride Mixtures ..........ccocconnucvuinarnnenes Fission-Product Behavior ...t oo er e st en et Solubility of Noble Gases in Molten Fluoride Mixtures ..........cccooooviiiimeveecieeeeeeeeeeeeeereeereen Solubility of HF in LiF-BeF, Mixtures .....ccoccoiiiciiiciiniiec it necenenessecieseseceisessisecens Solubilities of Fission-Product Fluorides in Molten Alkali Fluoride~Beryllium Fluoride SOIVENTS ..o e ettt s sttt Chemistry of the Corrosion Process ..ot et crt sttt e s e e ne Activity Coefficients of NiF, in LiF-BeF , oo, Solubility of NiF, in LiF-BeF, (61-39 Mole %) ......cccoovviii e Experimental Determination of Chromium Diffusion Coefficients in Molten Salt—Incone! Systems ...t bt ee e Vapor Pressures for the CsF-BeF, System ..o Fused Chlorides as Heat Transfer FIuids ..o eve s Permeability of Graphite by Molten Fluoride Salts .........o.ococviiiiiiciceeeree e, Preparation of Purified Materials ...ttt st e eerese e e Preparation of CrF o oo e see e e et s e st e vt ns e ene e ans Production-Scale Operations ...ttt eeae s resserete e et ere e e saan e seessreeneerenee Experimental-Scale OPerations ... et ee e s et e et e e e e v e re s e eeeeeae e e e eseeas Transfer and Service OPerations ...........iccoiiriieiiisice ettt st e s ess s e e eeseestsbeesteststea sue xi Part 1 REACTOR DESIGN STUDIES 1.1. DESIGN STUDIES H. G. MacPherson Reactor Projects Division INTERIM DESIGN REACTOR Conceptual design studies of a power reactor have led to an ‘‘interim design reactor,’”’ which is described in a report entitled Molten Salt Reactor Program Status Report. Since this report will not be issued for general circulation until September, a brief description of the interim design reactor will be given here, together with some design information that is not included in the status report. The interim design reactor is a two-region, homogeneous, molten-salt reactor with a single fuel pump at the top of the reactor core, as shown in Fig. 1.1.1. The gross electrical output is 275 Mw, and, since 15 Mw is required in the plant, the net power output is 260 Mw. The net over-all plant efficiency is about 40%, and thus the reactor has a thermal capacity of 640 Mw. The molten salt fuel used in the reactor core is initially a mixture with the following approximate composition: 61.8 mole % Li’F, 36.9 mole % BeF,, 1.0 mole % ThF,, and 0.3 mole % UF,. The volume of the system is such that the initial critical inventory is about 600 kg of U235, Fuyel reprocessing is initiated at the end of one year in small batches equivalent to reprocessing the entire fuel charge once per year. With this system of reprocessing, the inventory of U235 builds up to about 900 kg at the end of the first year and remains approximately constant for the next 20 years. The inventory of U233 gradually builds up to about 300 kg at the end of 20 years. The blanket salt is a mixture of 71 mole % LiF, 16 mole % BeF,, and 13 mole % ThF,. The average conversion ratio for this reactor is 0.53 over the 20-year period. An off-gas system is provided in a side stream of a circulating fuel circuit. The fuel is purged of xenon and krypton by mixture with a helium purge gas. The purge gas is circulated through cooled charcoal beds of sufficient size to absorb the xenon and krypton, and the helium is recircu- lated to provide a continuous purge. The circulating fuel, which is pumped from the top of the reactor, is divided into four streams that lead to four primary heat exchangers. The four streams return to the reactor core and enter tangentially at the bottom to provide a swirling motion to the fuel as it rises in the reactor core. Heat is interchanged between the fuel and the steam in two sodium circuits in series, the first being radioactive because of delayed neutrons and the second being nonradioactive. For this reactor system a fuel cost of 2.5 mills/kwh and a total power cost of 8.9 mills/kwh have been estimated. The pertinent features of the interim design reactor are given in Table 1.1.1. REACTOR CORE CONFIGURATION J. Y. Estabrook W. S. Harris The shape of the core of the interim design reactor is somewhat arbitrary; however, certain criteria were adhered to in its definition. First, the volume of fuel in the core was to at least equal, but not greatly exceed, that of an 8-ft-dia sphere, that is, 268 ft3. Second, the shapes of sections taken through the walls of the core were to be expressed as simple algebraic equations in order to assure smooth shapes, easily calculated volumes, and dimensional reproducibility. The fuel inlet to the core was to consist of four 10-in. sched-40 pipes bent slightly to impart a rotational component to the fuel as it progressed through the core. The four inlet pipes determined the minimum diameter of the bottom of the reactor. The fuel exit was fo consist of an 18-in.~dia pump inlet sufficiently removed from the main body of the core to make the fuel surrounding the pump definitely separate from the critical mass of the reactor. Between the bottom and the large central spherical portion of the core, a conical shape was introduced, as shown in Fig. 1.1.1. The included angle of the cone (60 deg) is probably the most arbitrary part of the core shape and it will remain so until hydraulic experiments are performed. A large angle tended to minimize the fuel inventory, but it may bring about intolerable flow separation. The curve joining the spherical portion of the core with the pump inlet was determined by sketching in a curve, noting its point of tangency (40 in. above the center of the sphere) and then calculating the slope at that point (1.62088). MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 286364 \ ! ‘ BLANKET SECTION A-A \ PUMP FUEL PUMP MOT‘OR MOTOR ‘ BLANKET EXPANSION TANK SIPHON DRAIN A=T | LN FUEL LINE TO | 1~ HEAT EXCHANGER | IJI’ : | FUEL EXPANSION TANK Fig. 1.1.1. Reactor Vessel and Pump Assembly. Tuble ].]ll. PERIOD ENDING JUNE 30, 1958 Reactor Plant Characteristics Fuel Fuel carrier Neutron energy Moderator Primary coolant Power Electric (net) Heat Regeneration ratio Clean (initial) Average (20 years) Blanket Estimated costs Total Capital Electric Refueling cycle at full power Shielding Control Plant efficiency Exit fuel temperature Steam Temperature Pressure Second loop fluid Third loop fluid Structural materials Fuel circuit Secondary loop Tertiary loop Steam boiler Steam superheater and reheater Active=core dimensions Fuel equivalent diameter Blanket thickness Temperature coefficient, (Ak/k)/°F Specific power Power density Fuel inventory Initial {(clean) Average (20 years) Critical mass (clean) Burnup >90% U235F4 (initially) 62 mole % LiF, 37 mole % Ber, 1 mole % ThF4 Intermediate LiF-BeF2 Fuel solution circulating at 23,800 gpm 260 Mw 640 Mw 0.63 ™~ 0.53 71 mole % LiF, 16 mole % Ber, 13 mole % ThF4 $69,800,000 $269/kw 8.88 mills/kwh Semicontinuous Concrete room walls, 9 ft thick Temperature and fuel concentration 40:6% 1210°F at approximately 83 psia 1000°F with 1000°F reheat 1800 psia Sodium Sodium INOR-8 Type 316 stainless steel 5% Cr, 1% Si steel 2.5% Cr, 1% Mo steel 5% Cr, 1% Si steel 8 ft 2 ft —~(3.8 T 0.04) x 103 ~ 1000 kw/kg 80 kw/liter 600 kg of U233 ~ 900 kg of U23S 267 kg of U233 Unlimited MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT The curve had the appearance of part of an ellipse with its center about 2 in. above the entrance to the pump and perhaps 30 to 50 in. off to one side of the core centerline. The general equations of an ellipse having the ordinates shown in Fig. 1.1.2 were used to determine the distance () off the core centerline: b2x2? 4 aly? ~ 2ba2y +h2a? ~g2p2 -0 d b2 7% 1.62088 . dx aZ(y___b) By substituting the values of x and y shown in the sketch into the general equations, it was possible to solve for b without determining the coefficients a and b; b was found to be 39.091. The determi- nation of a and & was then made by simply transferring the origin so that it was at the center of the ellipse. The general equations were then x2 y2 —_t =1, a? b2 dy b2 x a2y The calculations gave a = 34.15 and & = 30.14. When the result was checked for fit with the spherical portion of the core, it was found that a better match was obtained by slightly changing UNCLASSIFIED ORNL-LR—-DWG 31190 SLOPE AT THIS POINT =1.62088 ¥ \ h 24.678 in. » — 2 in —-——————— 28 in.—————————» Fig. 1.1.2, with Pump Inlet. Curve Joining Spherical Portion of Core b to 39.078. The mismatch in slope at the point of tangency with the sphere was only about 0° 17, which is well within manufacturing tolerance for shapes of this size. WEIGHT ANALYSIS OF REACTOR SYSTEM J. Y. Estabrook W. S. Harris The interim design reactor was used as a basis for preliminary rough estimates of the weights of the various components of such a system. mation on the weights is pertinent to remote handling and maintenance problems, plant layout, and construction planning and costs. For these estimates it was assumed that sodium would be used as an intermediate heat exchange medium between the fuel and steam and that boilers of the Lo&ffler type would be used. The data obtained are presented in Table 1.1.2. Infor= EVALUATION OF FABRICABILITY OF REACTOR VESSEL E. J. Breeding Fabricability was a major consideration in the layout of the interim design reactor. Since the first reactor does not need to be designed for mass production, it was not considered to be essential to maintain interchangeability in vessel parts. it will be permissible for the fabricator to fit the parts, within reasonable limitations, to the tolerances he can expect with his equipment and methods. This implies that the entire reactor vessel should be fabricated in one shop adequately equipped to accomplish the job. If parts were to be fabricated by various suppliers, the dimensional tolerances would have to be somewhat closer to assure fitup upon assembly. An arrangement of shapes that appear to be fabricable according to conventional practice of the pressure-vessel industry is shown in Fig. 1.1.3. Comments on the numbered sections are presented below. Outer Blanket Shell Section 1 is a conventional flanged section designed as a support to which the fuel pump assembly can be bolted. Section 2 is a transition ring to provide a satis- factory welding arrangement and stress distribution in the neck region. This section can be machined as part of section 1 or can be welded to section 1. PERIOD ENDING JUNE 30, 1958 Table 1.1.2, Estimates of Weights of Components of a Molten-Salt Reactor System Weight (Ib) Empty equipment Reactor vessel, including fuel and blanket salt pumps and motors 111,100 Fuel-toesodium heat exchangers (4) 62,300 Blanket salt-to=sodium heat exchanger 5,400 Fuel system superheaters (4) 279,000 Blanket system superheater 23,000 Fuel system reheaters (4) 50,000 Fuel piping 18,000 Blanket salt piping 4,000 Sodium piping 52,000 Sodium pumps and motors 80,000 Total 684,000 Fuel inventory in reactor, including fuel in pump and expansion spaces (325 £+ at 122 !b/ffa) 39,650 In piping (290 #3) 35,380 In heat exchangers (224 ff3) 27,328 Total 102,358 Blanket salt inventory In reactor, pump, and expansion space (920 13 at 201 1b/Ft3) 184,920 In heat exchanger (29 #3) 5,829 In piping (20 ft3) 4,020 Total 194,769 Sodium inventory In fuel-to-sodium heat exchangers (160 #3 at 49.5 Ib/f‘ra) 7,920 In blanket salt-sodium heat exchanger (26 ff3) 1,287 In superheaters (768 ff3) 38,016 in reheaters (240 fts) 11,880 In piping (830 ff3) 41,085 Total 100,188 Section 3 is a 2:1 ASME elliptical shell head with the center section removed. It is welded to section 2. Section 4 is a cylinder rolled from plate and joined with one longitudinal weld. Section 5 is a standard hemispherical shell head with a flued opening. Section 6 is a hemispherical shell head cut and flared to mate with section 7. Section 7 is a tapered conical section rolled from plate and joined with one longitudinal weld. Section 8 is a 2:1 elliptical shell with a flued nozzle for joining with the blanket salt inlet pipe. Inner Core Shell Section 9 is a conical, dished section with a flued opening. Section 10, which is the pump barrel housing, can be either a straight section of pipe or a cylinder rolled from plate and longitudinally welded. Section 11 is an elliptical shell head. MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED ORNL—LR—DWG 31191 BLANKET PUMP AREA FUEL PUMP AREA é secTion (1) g L S LinmomT— P e 2 ///‘ 777”_" | @77 77 ‘ | ///// \:\‘_\----—::-".'.'.‘ L. Ttttz é I R CORE SHELL %, /x ——BLANKET SHELL ';’//,/ e tT LoDl - — A 2 / ////////////////I//,/// //// ) Y 7 0 Fig. 1.1.3, Molten-Salt Reactor Yessel, See text for discussion of fabricability of numbered sections. Section 12 is a special shape with reverse flues. Section 13 is a cylindrical section rolled from plate or made from straight pipe. Section 14 is a plate with reverse flues. Section 15 is a forged neck or a machined plate weldment. Section 16 is a hemispherical shell head with a flared opening. Section 17 is a hemispherical shell head cut and flared to mate with section 18. Section 18 is a tapered conical section rolled from plate and longitudinally welded. Section 19 is a tapered conical section rolled from plate and longitudinally welded. The fuel inlet pipe sections, the thermal sleeves, and the outer shell section 7 will be a separate subassembly. Section 20 is an elliptical shell head. Blanket System Pump Housing Section 21 is a standard flanged section combined with a nozzle section of straight pipe or a standard flange machined on a cylindrical section. Section 22 is a standard elliptical head with a flued nozzle opening. Section 23 is a cylindrical rolled-plate section with the pump-discharge thermal sleeve welded in as in section 19, Section 24 is an elliptical shell head. Section 25 is a special shape fabricated to form a transition section between the vessels. FUEL PUMP DESIGN W. G. Cobb The fuel pump proposed for the interim design reactor is shown in Fig. 1.1.4. Fuel is fed directly from the reactor core into the volute, which is suspended from the roof of the expansion tank and is submerged in the fluid. The drive motor is completely canned and is located in @ clean portion of the purge gas system. The motor and pump rotating assembly are completely replaceable remotely as a cartridge unit. Fuel salt and blanket salt flow through and around structural parts for high-temperature cooling. Radiation shielding of the motor is provided by solid shield materials located below and around PERIOD ENDING JUNE 30, 1958 the motor compartment. Temperature control of the motor and shield materials is accomplished by circulation of a low-temperature coolant through passages lining the motor compartment. The pump and motor rotating parts are mounted on a common shaft that is supported by a salt-lubricated orifice-compensated hydrostatic lower bearing, which uses the impeller suction shroud as a journal, and by a compound thrust and radial bearing located immediately below the motor rotor and supported on the cooled thermal barrier. The upper bearing is a hemispherical orifice- compensated hydrostatic type that uses the pressurized helium purge supply for lubrication and support. An approximate bearing configuration is shown in Fig. 1.1.5. Thrust capacity is provided by the lower continuous circumferential pocket. Gas is admitted through several orificed ports into the feed grooves at the upper and lower pocket extremities to provide stability. Radial load capacity is provided by the multiple pockets surrounding the upper regions of the hemispherical journal. Each pocket has an orifice-compensated supply. The feed grooves in the thrust pocket and the radial pockets have depths of approximately 3 mils. The thrust-pocket depth would probably not exceed 1 mil. Instability interaction of the thrust and radial forces is eliminated by using a continuous circumferential bleed groove which exhausts to the motor cavity. It is expected that the high gas velocity through the inner portion of the thrust bearing will prevent diffusion of the primary fission products into the motor cavity. The elimination of liquid lubricants avoids the radiation damage and system contamination problems encountered with their usage. Approxi- mate gas conditions for bearing operation are given below: Gas supply pressure 50 psia Motor cavity pressure 1642 psia Shaft annulus pressure 19.2 psia Thrust 1000 [b Radial thrust 1000 Ib Shaft ennulus flow 1495 scfm Total thrust bearing flow 640 scfm Totel radial bearing flow 5¢5 scfm Total flow to motor cavity 9455 scfm Total gas flow 11.5 scfm MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED F-2-02-054-7083A FLOW RATE, 25,000 gpm HEAD, 71 ft SHIELD RETAINER 7 | SHAFT SPEED, 700 rpm SEALING HEAD — / POWER REQUIRED, 1000 hp / HANDLING LUG ELECTRICAL CONNECTION | LOW-TEMPERATURE MOTOR COILS COOLANT CONNECTION [l sHiELD Lock HELICAL COOLANT PASSAGES—— ' ROTOR GAS CIRCULATION ANNULUS LAMINATIONS | e E STATOR—- e ¥ 1 HYDROSTATIC ' GAS CONNECTION T O BEARING I NUCLEAR RADIATION SHIELD ‘ GAMMA SHIELD SUPPORT RING NEUTRON SHIELD—— HIGH-TEMPERATURE COOLANT JACKET — PURGE GAS FLOW MAXIMUM SALT LEVEL s N HIGH-TEMPERATURE ¥ 7 COOLANT PASSAGE - BYPASS FLOW CHANNEL SYSTEM EXPANSION VOLUME SALT NORMAL OPERATING SALT LEVEL FLOW VOLUTE (DISCHARGE TO REAR) | \2 e HYDROSTATIC SALT BEARING INLET IMPELLER— SCALE Q 1 2 3 FEET Fig. 1.1.4. Fuel Pump for Interim Design Reactor, 10 0.0015-in, CLEARANCE 45.75-in. RADIUS PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED ORNL-LR-DWG 34192 GAS SUPPLY || PLENUM -—T 0.0017-in. CLEARANCE SHAFT ANNULUS FLOW Figc 1.]-50 Bearing of Pump Shown in Fig. 1.1.4. NUCLEAR CALCULATIONS Modifications of Oracle Program Sorghum for Calculational Analyses of MoltensSalt Reactors L. G. Alexander The Oracle program Sorghum, a 31-group, zero- dimensional, reactor ‘‘burn-out’* code, was extensively modified as required for analyses of molten-salt reactors. Provision for automatic insertion of the thorium cross sections corre- sponding to the partial saturation of the resonances was made, with the option available in the input format. Similarly, a fuel option in the input was added to provide a choice between U233 and U235 as the isotope whose concentration is adjusted to bring the reactor to the critical condition. The output edit was modified to give the following: time in years; concentration of key fuel, in atoms /cm?3; total inventory of fissionable material, in kg (including U233 in blanket system); integrated net burnup, in kg (initial inventory plus integral of feed rate less total inventory at time T); | 0" \ | RADIAL BEARING ( POCKET / BLEED SLOT . / 7 THRUST POCKET Hemispherical, OrificesCompensated, Gas-Lubricated Bearing Designed To Be Used as Upper inventories of all nuclear species in core and blanket, in kg; fraction of neutrons causing fission in each of the three fissionable isotopes (U233, U233, Py?39); neutron absorption fourteen species; average n; average v; regeneration ratio; and neutron balance. These items are punched at each edit. A subroutine was also provided which edits simultaneously, via the console typewriter, the time, critical concentration of fuel, total inventory of fissionable material, regeneration ratio, and balance. A routine that provides for automatically beginning the core processing cycle at the end of the initial period (usually one year) was aiso added. Several provisions for detecting defective problems were made, including the console type-out just described, a subroutine to dump the memory on drive 1 periodically so that no more than 10 min of computing time is lost in event of machine error, and iteration overflow that stops the computation in cases where the critical calculation does not converge in 31 trials. The modified program appears to be working satisfactorily. ratios for MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT Analyses of Reactors Fueled with U233 L. G. Alexander The investigation of the initial-state nuclear characteristics of two-region, homogeneous, molten- fluoride-salt reactors fueled with U232 was con- tinved. The data required for comparison of reactors fueled with U233 or U235 qgnd having thorium concentrations in the range from 0 to 1 mole % ThF, in the fuel salt were obtained, and the results are plotted in Fig. 1.1.6. The comparison is based on reactors fueled with U233 or U235 having a basic core salt containing 31 mole % BeF, and 69 mole % LiF and a blanket salt containing 25 mole % ThF, and 75 mole % LiF. The regeneration ratios ronge up to 0.95 at inventories of less than 350 kg of U233 for g 600-Mw system having external fuel volyme of 339 ft3, In the interim design reactor described above, the fuel salt composition was 37 mole % BeF2—63 mole % LiF plus UF, and ThF,, and the blanket salt had a composition of 13 mole % ThF ,-16 mole % BeF 5=71 mole % LiF. The performance of reactors fueled with U233 and having these fuel and blanket salts is shown in Fig. 1.1.7. The UNCLASSIFIEDR ORNL-LR-DWG 31193 1.2 TOTAL POWER. 600 Mw EXTERNAL FUEL VOLUME: 339 ft° FUEL SALT: 31mole % BeF, + 69mole % L|F+ UF, + ThF, 1.0 |- BLANKET SALT: 25mole % ThF, + 75mole % LiF 0.8 ’ O 5 35 @ u?> ENVELOPE N g | = 0.8 & o l = w B € o4 ThF, IN FUEL SALT {mole %) o 0 A 0.25 o8 ; o 0.50 A 0.75 . 4 [ NOTE: CORE DIAMETERS INCREASE TO THE LEFT | 0 | ] 0 100 200 300 400 500 600 CRITICAL INVENTORY (kg) Fig. 1.1.6, Initial-State Nuclear Performance of Two-Region, Molten \ w \ o | > L Bfi ‘ | L Z 06 | | Q ‘ l , | I u! i ; 1 | = — NUMBERS ON CURVES ARE Q%H*_J CORE DIAMETER IN FEET , 0.4 4‘ i { | ‘ | 4 ThF, IN FUEL SALT | — i ‘ } {mole %) | 0.2 P A 0.25 J‘_ | A 0.50 | | L o 0.75 |L L } . { 5 l \ 0 [ 1 1 1 | ‘ 0O 20 40 60 80 100 120 140 160 180 200 CRITICAL INVENTORY (kg OF U233 Fig. 1.1.7. Initial-State Nuclear Performance of Two-Region, Homogeneous, Molten-Fluoride-Salt Re- actors Fueled with U233, regeneration ratios range up to 0.9 at critical inventories of U233 that are about one-fifth the critical inventories required for reactors fueled with U235, It was clear from the trend of the curves in Fig. 1.1.7 that the performance would be improved by increasing the thorium concentration, and therefore three reactors with 4 mole % ThF, in the fuel salt were studied. The results, presented in Table 1.1.3, give regeneration ratios that slightly exceed 1.0. |t seems reasonable that a further siight increase could be obtained by adjusting the diameter and the thorium concen- tration. The long-term performance of the UZ233.fyeled with an 8-ft-dia core was studied with the use of the Sorghum code. An extract from the results is shown in Fig. 1.1.8. As may be seen, the regeneration ratio fell rapidly from the initial value to about 0.96 during the first year of operation, while the inventory rose about 200 kg from the initial 560 kg. At this time processing of the core fluid was started at a rate sufficient reactor Tflble 1.]-3- PERIOD ENDING JUNE 30, 1958 Initial-State Nuclear Characteristics of Two-Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U233 and Having 4 Mole % ThF4 in Fuel Salt Fuel salt: 37 mole % BeF2 + 63 mole % LiF + UF4 + T!"lF4 Blanket salt: 13 mole % ThF4 + 16 mole % BeF2 + 71 mole % LiF Total power: 600 Mw (heat) External fuel volume: 339 f43 Core diameter, ft 6 8 12 U233 iy fuel salt, mole % 0.324 0.226 0.140 U233 gtom density* 12.14 8.48 5.27 Critical mass, kg of U233 149 246 479 Critical inventory, kg of y233 595 559 869 Neutron absorption ratios** U233 (fissions) 0.8715 0:8660 0.8785 U233 (4,) 0.1285 0.1340 0.1215 Be-Li«F in fuel salt 0.0887 0.0999 0.1274 Core vessel Li«F in blanket salt 0.0856 0.0616 0.0320 Outer vessel Leakage Th in fuel salt 0.6409 0.7734 0.9228 Th in blanket salt 03725 0.2517 0.1306 Neutron yield, i 2.19 2.21 Regeneration ratio 1.013 1.025 1.038 *Atoms (X 10! 9)/cma. **Neutrons absorbed per neutron absorbed in u233, to hold the concentration of fission fragments constant. The regeneration ratio was approximately stabilized, but the inventory continued to creep up to about 950 kg in 15 years. Meanwhile, the integrated net burnup, that is, total purchases less inventory, had reached 250 kg. The net burnup rate averaged about 16 kg per year. Analyses of Reactors Fueled with Plutonium D. Baxter Two cases of two-region, homogeneous, molten- fluoride-salt reactors fueled with plutonium were studied. The cores contained no thorium and were 6 and 8 ft in diameter, respectively. The results of the analyses are presented in Table 1.1.4. In comparison with corresponding UZ2335-fueled reactors, the critical inventories of these reactors are about one-third as great, and the required plutonium concentration appears to be well below the solubility limit in the salt mixtures of interest. The regeneration in the blanket was substantial; it amounted to about 40% of the fuel burned. The U233 formed in the blanket should be effective in compensating for ingrowth of fission products. 13 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 31195 1.10 | CORE DIAMETER: 8ft FUEL SALT: 37 mole % BeF,+ 63mole % 1.05 FLiF+UF, +TnF, o TOTAL POWER: 600 Mw {HEAT) %r_: PLANT LOAD FACTOR: 0.8 = EXTERNAL FUEL VOLUME: 339 ft3 o £ 1.00 ' @ w 2 w o w @ \ \ 0.95 —4 — rCORE PROCESSING BEGINS 0.90 ’ 1000 I /—— WENT \ 800 ‘ “(Ofli'/ /r/ 600 I B N g |+ CORE PROCESSING BEGINS ND I 400 t | o L 200 auRNY [ <0 Net A 0 0 2 4 6 8 10 12 14 16 TIME (years) Fig. 1.1.8. Long-Term Nuclear Performance of Two- Region, Homogeneous, Molten-Fluoride-Salt Reactors Fueled with U233 and Having 4 Mole % ThF4 in Fuel Salt, 14 ARGON AS A PROTECTIVE ATMOSPHERE FOR MOLTEN SALTS L. A. Mann Helium has been used almost exclusively at ORNL as the protective or ‘‘blanket’ gas for molten salts in experiments in both radiocactive and nonradioactive environments. Argon has been used occasionally, and other gases have been considered as possible alternates for use with particular designs, materials, and operating conditions (for example, krypton, xenon, neon, hydrogen, nitrogen, carbon dioxide)., =~ Several factors must be considered in selecting the cover gas. The major considerations are, of course, availability and cost of gas of the required purity. In establishing the specifications for the gas, it is necessary to study the effects of the gas on the container and other contacted materials, as well as the effect of the gas on the salt and the salt on the gas. If the gas is to be used in a radiocactive environment, determinations must be made of the activity that will be induced in the gas and the effect the activity will have on the system. For molten-salt reactor use, the requirement of ability to strip krypton and xenon from the fuel is an added factor. The scarcity and high cost factors immediately eliminate krypton, and xenon from con- sideration for use in large systems. The nonnoble gases can be evaluated only in terms of specific applications. The possible chemical activity of the nonnoble gases requires that a determination be made of the compatibility of the gas with all materials contacted under the proposed operating conditions. neon, A review of the various factors has indicated that only helium and argon are suitable for molten- salt reactor application. Cylinder argon costs, at present, about 6.5 cents per standard cubic foot, compared with about 4 cents for cylinder helium. Pipe line helium costs about 2.5 cents, but it is anticipated that the price will increase in the near future. Helium has the disadvantage that no considerable sources of helium have been found to exist outside the United States. Calcu- lations have indicated that the activity produced in the A40 isotope of natural argon by neutron PERIOD ENDING JUNE 30, 1958 Table 1.1.4. Initiol-State Nuclear Characteristics of Two-Region, Homogeneous, Molten-Fluoride Reactors Fueled with Plutonium Total power: 600 Mw (heat) External fuel volume: 339 3 Fuel salt: 37 mole % BeF2 + 63 mole % LiF + PuF3 Blanket salt: 13 mole % ThF4 + 16 mole % BeF2 + 71 mole % LiF Core diameter, ft 6 8 Thorium in fuel salt, mole % 0 0 Pu in fuel salt, mole % 0.045 0.013 Pu atomic density* 1.55 0.460 Critical mass, kg of Pu 19.7 13.7 Critical inventory, kg of Pu 78.1 31.1 Neutron absorption ratios** Pu (fissions) 0.6043 0.6291 Pu {n,y) 0.3957 0.3709 Li-Be-F in fuel salt 0.1404 0.3093 Core vessel 0.1282 0.1459 Li-Be-F in blanket salt 0.0266 0.0233 Outer vessel 0.0047 0.0034 Leakage 0.0070 0.0033 Th in blanket 0.4625 0.3530 Neutron yield, 1 1.76 1.84 Regeneration ratio 0.463 0.35 *Atoms (X 10-]9)/cm3. **Neutrons absorbed per neutron absorbed in y233, capture is negligible compared with that of the Some of the materials being considered for krypton and xenon that will escape from the fuel in a molten-salt reactor. BEARING TESTER DESIGNS L. V. Wilson Hydrodynamic Bearing Tester A test rig was designed for experimental studies of hydrodynamic bearings and bearing materials in molten salts. The apparatus consists primarily of a model PK centrifugal pump which was modified so that a journal could be mounted on the shaft at the normal impeller position, as shown in Fig. 1.1.9. The bearing is flexibly supported, and radial force may be applied to it through the load column attached to the load beam. The force for the bearing is applied to the load beam by an air cylinder mounted on the spool piece. The bearing torque is measured by changes in power input to the motor that drives the pump shaft. bearings have coefficients of thermal expansion that are approximately one-third the coefficient of INOR-8. Since a stress problem will exist when the bearing assembly is raised to operating temperatures, a compensating-column type of journal mounting is being investigated. In the design illustrated in Fig. 1.1.9, the journal is mounted on eight radial load columns made of the same material as the journal. These load columns are located near the centerline of the shaft and near the axial center of the bearing to keep the differential expansion to a low enough value to be absorbed by the elastic bending of the load columns. The bearing load on the journal is transmitted to the shaft by the columnar reaction of the load columns. Results of operating of this test rig are described in Chapter 1.2 of this report. MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT LOAD BEAM—— | PIVOT BELLOWS — AN LOAD COLUMN Fig. ]-]ogu 16 ® > \!_/ . UNCLASSIFIED ORNL-LR-DWG 31186 B:T If !1 | ! | / "LI‘ / [ } ! / “.‘ ! .*': { ( / [ = / _ i ! E 7 b L = | . | ! ‘ i i B = = > = == SUPPORT PINS L 7 v N N - NI e BEAR!NG/ L g JOURNAL TO FILL AND DRAIN TANK Apparatus for Experimental Studies of Hydrodynamic Bearings, Hydrostatic RotatingsPocket Bearing Tester A rig designed for water tests of a hydrostatic orifice-compensated bearing is shown in Fig. 1.1.10. The bearing pockets and orifices are located on the rotating member and the pressure supply is provided by the centrifugal pumping action of the supply channels to the orifices. The rotating element is relatively rigid, and the stationary element, which is mounted on two ball bearings, TACHOMETER BEARING POCKET PRESSURE TAF’/ AIR CYLINDER IMPELLER PERIOD ENDING JUNE 30, 1958 is free to move in the direction of the load, which is applied by means of an air cylinder. Motion of the stationary element relative to the rotary element is measured by dial indicators. The stationary element is a Plexiglas shell, reinforced with cutout steel pipe, to permit observation of beoring operation. Pressure taps to measure pressure distribution are located in the Plexiglas around the periphery of the bearing. UNCLASSIFIED ORNL-LR-DWG 31197 COOLING WATER i J U COOLING COIL DIAL INDICATCR Fig. 1.1.10. Apparatus for Experimental Studies of Hydrostatic Rotating-Pocket Bearings, MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED ORNL—LR—DWG 31198 COOLANT QUTLET COOLANT INLET LEAKAGE CUTLET GAS INLET PRESSURE TAPS __—0ORIFICE .____—JOURNAL PUMP TANK BEARING POCKETS BEARING BYPASS FLOW VOLUTE IMPELLER Fig. 1.1.11. Apparatus for Experimental Studies of Hydrostatic Stationary-Pocket Beorings. 18 Bk by e s S P nma e Al S et Hydrostatic StationarysPocket Bearing Tester A rig designed for testing hydrostatic stationary- pocket bearings in water and then in molten salts is shown in Fig. 1.1.11. A model PK pump was modified for this application by removing the lower ball bearing and seal, replacing them with a labyrinth seal, and mounting the journal of the hydrostatic bearing on the impeller hub. The lower bearing load is carried by the hydrostatic bearing, PERIOD ENDING JUNE 30, 1958 which is supplied with high-pressure fluid from around the outer periphery of the impeller. The fluid flows from the impeller to the bearing orifices, into the bearing pockets, and out to the expansion tank. The load on the bearing is produced by the radial hydraulic unbalance in the volute acting on the impeller. The bearing load and the direction of the load are measured by the pressures in the bearing pockets and their resultant vector. 19 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT 1.2. COMPONENT DEVELOPMENT AND TESTING H. W. Savage W. B. McDonald Reactor Projects Division FUEL PUMP DEVELOPMENT W. F. Boudreau A. G. Grindell Development Tests of Salt-Lybricated Bearings P. G. Smith W. E. Thomas L. V. Wilson Hydrodynamic Bearings. — Construction of equip- ment for testing hydrodynamic bearings in salt mixtures was completed, and cold and hot mechani- cal shakedown tests were performed (see Chap. 1.1 for description of equipment design). The first test was run to obtain a preliminary evaluation of tolerable bearing loads, and the bearing consisted of an INOR-8 journal and sleeve, since this material was known to have excellent corrosion resistance to salt mixtures. While dry, the bearing was subjected to one thermal cycle from room tempera- ture to 1200°F and return. Heating required 26 hr and cooling 12 hr, and the journal was rotated in the bearing one-quarter turn each hour by hand. An attempt was then made to operate the bearing in fuel salt 130 (LiF-BeF ,-UF,, 62-37-1 mole %) at 1200°F with a shaft speed of 1200 rpm. The bearing operated for about 20 min ot an applied load of up to 100 Ib. When the load was increased to 125 |b, the power consumption of the drive motor increased rapidly, and within a few additional minutes the load became sufficient to stall the The test was interrupted at this point, and the shaft and bearing assembly were removed for examination. Results of the examination are not yet available. A second similar bearing, modified to assure better filling, is operating satisfactorily. Tests are planned in which fuel 130 will be used as the lubricant and some of the operations ex- pected for the fuel pump in a molten-salt reactor will be simulated in order to obtain bearing per- formance data for comparison with the results of calculations based on theoretical considerations. The calculations related bearing clearance and minimum film thickness to bearing load, speed, and salt viscosity and were the basis for selecting testing clearances of 0.003 and 0.005 in. between radial journal and sleeve. Calculations of the motor. 20 journal power required and of Sommerfeld’s number were also made for bearings having either of the two clearances specified. The INOR-8 bearing tested had a radial clearance of 0.005 in. Other design and experimental investigations are under way to determine @ suitable means for mounting refractory metal or cermet bearing journals and sleeves (molybdenum and tungsten carbide with 12% cobalt binder) to the basic structural material. The design must take into account the differences in thermal expansion between the materials, and imposition of stresses sufficient to distort the bearing material must be avoided. Hydrostatic Bearings. — Calculations have been made to determine load capacity, eccentricity, bearing liquid flow, bearing pressure distribution, journal speed, orifice size, bearing clearance, and supply pressure for tests of hydrostatic bearings. Fabrication of the test equipment (described in Chap. 1.1) is nearly complete. Preliminary water tests will be performed to obtain the bearing load and pressure distribution with respect to test loop flow, flow resistance, pump speed, and power. Rotating-Pocket Hydrostatic Bearings. — The test apparatus described in Chap. 1.1 for evaluating the rotating pocket hydrostatic bearing,! with water as a lubricant, was completed. The test bearing consists of an aluminum journal 11.351 in. in diameter and 2.5 in. in height. There are twelve pockets, 0.090 in. deep, equally spaced around the periphery, with all lands, both peripheral and axial, 0.375 in. in length. The bearing is con- structed from Plexiglas pipe, bored to 11.375 in. ID, to give a radial clearance of 0.012 in. between journal and bearing. Each pocket on the journdl is supplied with pressurized lubricant by one vane of the impeller. Each vane consists of a 3/4-in.-dic1 drilled hole and takes its inlet from the 4-in.-dia impeller eye. The test-bearing impeller is designed to deliver a 50-ft head (with water) at the orifice face when rotating at 1300 rpm. IB. W. Kinyon, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL.-2474, p 16. Development Tests of Conventional Bearings D. L. Gray W. E. Thomas Organic-Liquid-Lubricated Bearings. — The high- temperature sump pump being used to conduct an experimental evaluation of the behavior of Dow- therm A {eutectic mixture of diphenyl and diphenyl oxide) as a lubricant has continued to operate satisfactorily for a period of 1620 hr. The pump is circulating fuel 30 (NaF-ZrF4-UF4, 50-46-4 mole %) at a temperature of 1200°F; the shaft speed is about 2600 rpm; the temperature of the Dowtherm A supplied to the bearings is maintained at 180°F, with approximately a 5°F rise through the test bearings. Upon completion of 1000 hr of operation, the pump was disassembled and the test bearings were inspected. The double-row angular-contact ball bearings that were submerged in the lubricant during operation showed no detectable wear or imperfections. The diameters of the aluminum bearing and the Inconel journal were measured and the diametral clearance was found to have increased by 0.001 to 0.0045 in., as measured at room temperature. Since 0.0045 in. is the design diametral clearance, this initial wear may be attributed to ‘‘wear in"' rather than to poor lubri- cation. It was also noted that the O-rings (buna N material) had increased in size upon exposure to Dowtherm A, but the elasticity of the material had not been affected greatly. No leakage was observed from the O-rings during operation. Bearing and Seal Gamma Irradiation. — By the end of the quarter the pump rotary assembly being operated in the canal at the MTR as a bearing test under gamma irradiation had accumulated a gamma- ray dose of 5.9 x 10° r. Since the bulk of the Gulfcrest 34 bearing lubricant is external to the radiation field, the oil has accumulated a gamma- ray dose of about 10% r. Samples of oil totaling 931 ¢m?, collected at various time intervals from the lower-seal-leakage catch basin and from the bulk oil supply, as well as four different helium samples, have been received from the MTR for analyses. The rotary element is operating under the following conditions: shaft speed, 4000 rpm; total lubricating oil flow, 4 gpm; temperature of oil inlet to bearing housing, 143°F; lower seal helium purge flow, 60 liters/day; upper seal helium purge flow, 500 liters/day. PERIOD ENDING JUNE 30, 1958 The barrier heater used to simulate gamma heating in the lower seal area completely shorted out on March 3. The oil temperature level dropped only 5 to 7°F, and the loss of the heater was not considered sufficient cause for terminating the fest. For seven days during March, the test assembly operated without gamma irradiation while necessary maintenance was performed on other MTR canal tests. A power failure in April stopped test operation for approximately 20 min. During this power failure, some 750 cm® of oil leaked from the lower seal oil purge line before the helium purge flow was regained. Gas-Lubricated Bearing Studies D. L. Gray Gas-lubricated bearings are being considered for use in molten-salt pumps. In one design,2 the gas bearing is hemispherical. A study of the literature has indicated that this type of bearing is probably feasible, but it is known that gas- lubricated bearings require very small running clearances and are subject to self-induced vibra- tions. Calculations have indicated that a thrust bearing of 8.5-in. spherical radius, supported with helium, as proposed for a 5000-gpm molten-salt pump, will require a minimum supply pressure of 45 to 50 psia. A gas flow of 5 scfm will be re- quired to maintain @ minimum clearance of 0.001 in. in this bearing. For a 25,000-gpm pump, a thrust bearing of 15.75 in. spherical radius that is de- signed for a constant clearance between journal and bearing at equilibrium conditions will require a 5-scfm helium flow into the pump tank end a 4-scfm flow into the motor cavity, at a supply pressure of 40 psia, in order to provide a minimum clearance of 0.001 in. A small hemispherical bearing approximately 8 in. in diameter will be used to test the validity of these calculations. Mechanical Seals D. L. Gray Labyrinth and Split-Purge Arrangement. — Oper- ation was terminated of the NaK pump which was modified, as previously described,? to include a 2W. G. Cobb and M. E. Lackey, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 18. 3p. G. Smith and L. V. Wilson, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 18. 2] MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT labyrinth in the pump shaft annulus between the seal region and the pump bowl and a purge gas inlet midway of its length so that a portion of the helium purge gas moved upward in the shaft annulus past the lower seal region and into the catch basin and the remainder of the purge gas moved down the shaft annulus and into the gas volume of the pump tank. The pump had operated for 3386 hr, but, during the last three weeks of the test, it was observed that the pump tank pressure responded slowly to changes in gas supply pressure. [t was deduced that the small holes drilled through the heat barrier shaft sleeve to the labyrinth had become constricted. Efforts to clear the passages were unsuccessful, and the test was terminated when communication from gas supply to the pump tank became impractical. A post-test inspection of the pump rotary assembly revealed that the lower surface of the heat barrier was covered with NaK and oxides. Brown- and black-colored material had practically filled the gas supply holes which fed the labyrinth, circum- ferential, V-groove plenum. This material was later determined by chemical analysis to contain approximately 7% carbon, and thus it was evident that lubricating oil or graphite had played a part in the formation of the constriction. Approximately 0.064 in. of the lower seal graphite nose piece had been worn away during the 3386 hr of service. Both face surfaces of the lower seal were free of NaK and, although worn, appeared to be in excellent operating condition. Bellows-Mounted Seal. — By the end of the quarter, the bellows-mounted seal being subjected to an endurance test in a NaK pump had accumu- lated more than 5100 hr of operation. This modified Fulton-Sylphon type of seal is being tested without gas purge protection from the vapors of the NaK being pumped. The main helical spring was removed from the standard seal and replaced with 24 small- diameter helical springs equispaced in a circle around the bellows and concentric with it. There are no elastomers in this seal, and therefore it may prove to be suitable for a pump that is to operate in a radiation field. As mentioned previ- ously,? the seal leakage during the first 2000 hr of operation averaged approximately 25 cm®/day, but it then decreased to approximately 8 cm®/week 4s. M. DeCamp and W. E. Thomas, MSR Quar. Prog. Rep. fan. 31, 1958, ORNL-2474, p 20. 22 and has remained at that rate despite two shut- downs required to correct deficiencies in parts unrelated to the test. Radiation-Resistant Electric Motors for Use at High Temperatures S. M. DeCamp The successful development of molten-salt- lubricated bearings would permit consideration of totally submerged canned-rotor pumps for molten- salt service. Such a pump would require an electric drive motor that would operate reliably at high temperatures and would be resistant to damage by A preliminary study has indicated the need for high permeability magnet steels suitable for 1250°F operation, current-carrying conductors of acceptable strength and electrical conductivity at 1250°F, and electrical insulation for use between the conductors ond the magnet iron or steel that would not be damaged by high temperatures. An investigation of electrical insulating materials was initiated. At present, ceramics and glasses appear to be the most promising materials for this application. Several wire manufacturers are de- veloping insulated wire for missile electronic components, and some of their products may be suitable for the motor being considered. A contract has been given to the Louis Allis Company for the production of six test coils suit- able for use at 1200°F. These coils will be insu- lated with a coating of metallic oxides that appears to have good radiation and thermal stability. The first test coils, which are expected in September, will be tested at the desired service temperature, and if they have satisfactory thermal stability they will be tested for nuclear radiation stability. Work performed by others in the development of motors for use at high temperatures has indicated that loss of strength and high resistance of the conductor will be problems at the contemplated operating temperature of 1200°F. Nickel and copper alloys are being considered for use as conductors. At 1200°F, the Curie point of common electrical steels is approached, and it may be necessary to use cobalt steels. radiation. Piping Preheating Tests D. L. Gray Tests were made to determine the length of time required to bring a typical fused-salt piping system to a temperature (900°F) substantially above the salt freezing temperature by preheating a portion of a closed forced-circulation test loop with electric-furnace elements and transmitting the heat to the remainder of the loop by forced circulation of helium within the system piping. The preheating tests were made in a loop that included a pump, piping, a venturi, and throttling valves. About 1200 |b of Inconel was used in constructing the loop. Forty-nine per cent of the total loop surface was subjected to radiant heating from the flat, ceramic-covered, radiant-wire heating elements. The loop helium pressures and pump speeds used in three tests are given below: Loop Pressure Pump Speed (psig) (rpm) 14 3475 98 3450 10 0 Fig. 1.2.1, Mechanical Joint Test Loop. PERIOD ENDING JUNE 30, 1958 The required temperature of 900°F was attained in a period of approximately 9 hr in the test for which a system pressure of 98 psig was used. It was not possible to reach the desired temperature in either of the other tests. DEVELOPMENT OF TECHNIQUES FOR REMOTE MAINTENANCE OF THE REACTOR SYSTEM E. Storto W. B. McDonald Mechanical Joint Development A. S. Olson Further screening tests were conducted on the three types of mechanical joints, described previ- ously,® that are being developed to facilitate remote removal and replacement of reactor com- ponents. The loop shown in Fig. 1.2.1, which 5A. S. Olson, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 20. M UNCLASSIFIED PHOTO 31474 (1) Joint being tested, (2) Calrod heaters and shim stock binding, (3) clamshell heaters, (4) loop frame, (5) molten-salt pump, (6) salt sump, (7) insulation support trough, (8) insulation. 23 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT was used for these tests, is capable of circulating molten salts at temperatures up to 1500°F, and controlled thermal cycling conditions can be imposed. An assembled joint is shown in Fig. 1.2.2. The joints are subjected to the following typical test procedure: 1. joint assembled and leak checked while cold (leakage specification, <1 x 10~7 c¢m® of helium per second), 2. joint welded into loop and loop operated through 50 thermal cycles between 1100 and 1300°F, 3. loop drained and joint removed intact, 4. joint leak checked, 5. joint separated, remade, and leak checked. None of the joints fested have given any indi- cation of leakage during operation, and the leakage rates found before and after the tests have been satisfactory. Three joints that were photographed upon separation after testing are shown in Figs. 1.2.3, 1.2.4, ond 1.2.5. The freeze-flange joint shown in Fig. 1.2.3 was separated without difficulty, the backup seal ring was easily removed, and the frozen salt was readily cleaned from the flange faces. The in- dented-seal flange joint shown in Fig. 1.2.4 was also separated without difficulty, but the gasket stuck to one-half the joint and had to be machined Fig. 1.2.30 sleeve, (3) groove for backup seal ring. 24 Freeze-Flange Joint Separated After Testing. UNCLASSIFIED PHOTO 31519 o ¥ Fig. 1.2.2. After Removal from Test Loop. Assembled Indented-Seal Flange Joint UNCLASSIFIED | PHOTO 31525 | (1) Frozen salt between flange faces, (2) insert o — - e — e e T e e 1 = = PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED PHOTO 31132 —I T l ¥ l ¥ ' ¥ l I ' ¥ l T I v l_' l i l T | v ‘ O 1 2 3 INCHES Fig. 1.2.4, Indented-Seal Flange Joint Separated After Testing. UNCLASSIFIED PHOTO 31423 Ew R Fig. 1.2.5. Cast-Metal-Sealed Flange Joint Separated After Testing. 25 e e e, S e e —— MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT from the flange face. The cast-metal-sealed flange joint (Fig. 1.2.5) was heated to above the melting point of the sealing metal in order to effect sepa- ration, and heavy oxidation of the sealing metal resulted. Considerable force was required to separate the flanges, and it was impossible to remake a tight joint without reworking both sealing faces. The freeze-flange and indented-seal joints are being reassembled for testing with sodium as the circulated fluid, since sodium may be used as the secondary coolant for a molten-salt reactor system. All parts of the freeze-flange joint are being re- used. A new gasket will be used in the reassembled indented-seal flange joint. A new freeze-flange joint for use in 4-in. pipe has been fabricated and assembled, as shown in Fig. 1.2.6. Details of the design of this joint may be seen in Fig. 1.2.7. The 4-in. joint will be installed in an existing salt-circulating loop for testing. Provision has been made for continuous leak monitoring of the joint during high-temperature operation of the loop. Remote Manipulation Techniques C. K. McGlothlan Experimental remote maintenance work on a NaK pump which had been used in a high-temperature UNCLASSIFIED PHOTO 31405 Fig. 1.2.6. New Freeze-Flange Joint for 4-in. Pipe. 26 PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED ORNL-LR-DWG 31199 GAP (~v g in.) NICKEL-PLATED SOFT COPPER CLAMP - BACKUP SEAL RING AIR CHANNEL FOR COOLING & @@ o “o REINFORCING WEB RING INSERT TO PROVIDE LABYRINTH FOR SALT LEAKAGE TN / | e ——— T SALT STREAM - Lh______/ WELD OF FLANGE TO LCOP (4-in. PIPE) THIN SECTION TO PROVIDE TEMPERATURE GRADIENT FOR FORMATION OF FROZEN SALT SEAL & SCREEN INSERT IN FREEZE ANNULUS TO REDUCE SALT HOLDUP VOLUME AND PROVIDE BASE FOR FROZEN SALT CAKE PUMP OUT TUBE FOR LOCATION QF GAS LEAK OR TO BUFFER GAS LEAKAGE CLAMP LUGS Fig. 1.2,7. Design Details of Joint Shown in Fig. 1.2,6, 27 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT system was completed as previously planned.‘s The results of this work’ showed that remote maintenance is definitely feasible for a pump or any other readily accessible component in a fused- salt system. Good vision is essential to the performance of maintenance operations by remote manipulation, and direct viewing through cell windows, indirect viewing with periscopes, and closed-circuit tele- vision viewing have been considered for this application. The use of television is contingent on the availability of radiation-resistant cameras and the development of stereo viewing. A review of work in progress at ORNL and in industry has given reasonable assurance that video cameras éc. K. McGlothlan, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 24. 7w. B. McDonald, C. K. McGlothlan, and E. Storto, Remote Maintenance Experimental Work on a Reactor System Pump, ORNL CF-58-4-93 (April 23, 1958). ¢ % = can be constructed that will operate satisfactorily for extended periods in gamma radiation of up to 1 x 10° r/hr. Since no good stereo television system was found to be readily available, an experimental optical stereo system was designed and constructed. Stereo pair photographs were used to simulate television monitor screens acti- vated by a pair of television cameras remotely located. The three-dimensional viewing achieved with this system was very satisfactory, and there- fore specifications were issued for the purchase of a matched pair of closed-circuit television systems, the monitors of which will be substituted for the stereo pair photographs. A study of the feasibility of remote welding by mechanical manipulation was initiated. A pair of Argonne model 8 manipulator arms was set up to handle a standard Heliarc welding torch and filler rod, as shown in Figs. 1.2.8 and 1.2.9. Two welders succeeded in producing high quality welds & UNCLASSIFIED | PHOTO 31477 St e Fig. 1.2.8. Remote Welding of Plate. 28 PERIOD ENDING JUNE 30, 1958 UNCL ASSIFIED PHOTO 31479 Fig. 1.2.9. Remote Welding of Pipe. with this apparatus on standard Inconel plate and pipe test coupons. The welds passed x-ray and dye-penetrant tests. A specimen of pipe opened by a leak was also repaired. Both welders reported that they had no difficulty in becoming accustomed to the apparatus (Figs. 1.2.10 and 1.2.11), but they both stated that inadequate viewing was a severe handicap. The best results have been achieved by viewing through binoculars, but this method results in severe eyestrain. Experiments are now being conducted on the development of camera lens filtering and special illumination to permit remote viewing of the welding process with television. Remote Maintenance Demonstration Facility An engineering layout has been approved for a remote maintenance demonstration facility (Fig. 1.2.12), and detailing and fabrication of the com- ponents are proceeding. An order has been placed for a General Mills manipulator, and specifications for a closed-circuit television viewing system were issued, as stated above. An area has been cleared for the facility in the experimental engineering building (Building 9201-3), and design of the manipulator runway is essentially complete. Heater-Insulation Unit Development Pilot model heater-insulation units® (Fig. 1.2.13) were fabricated and are to be tested on a loop circulating molten salts. The units are constructed on the clamshell principle and are designed to be applied to pipe sections and removed from them by remote manipulation. Each unit will serve to preheat its pipe section preparatory to filling of the system and will insulate it during high-temper- ature operation. Joints between units are covered with clamshell inserts that contain only insulation. 8A. L. Southern, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 25. 29 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT | | _ = g Fig. 1.2.11. Setup for Television Viewing of Remote Welding. 30 — e UNCL ASSIFIED | PHOTO 31475 HUNCL ASSIFIED PHOTO 31474 ) UNCLASSIFIED < ORNL-LR-DWG. 31482 Fig. 1.2.12. Remote Maintenance Demonstration Fa- cility. EVALUATION OF EXPANSION JOINTS FOR MOLTEN.SALT REACTOR SYSTEMS R. L. Senn A test program was initiated to evaluate expansion joints that are commercially available, with respect to their usefulness in molten-salt reactor systems. If expansion jeints can be used in fuel and coolant circuit piping, the extra space and fluid inventories required for thermal expansion loops can be avoided. Six expansion joints were ordered from three different vendors. Each vendor will supply one Inconel and one stainless steel unit. The Inconel joints are to be tested with the molten salts of current interest, and the stainless steel joints are to be tested with sodium. A test stand was designed and is presently being built which will provide for simultaneous testing of three expansion joints. Each joint will by cycled through its maximum axial traverse once every 2 hr for 1000 cycles. The cycle period is sufficient to allow relaxation of stresses introduced into the bellows by each half cycle. The joint will be filled with the appropriate fluid and held at 1300°F and 75 psi during the test. PERIOD ENDING JUNE 30, 1958 UNCL ASSIFIED QRNL-LR-DWG, 31559 Fig. 1.2.13. Pilotps'and changes D. B. Trauger in operating procedures. The major improvements 3. A. Conlin P. A. Gnadt to be incorporated in new loops and test focilities are shown schematically in Fig. 1.2.16. A loop that incorporates all the improved features has been fabricated and is shown in Fig. 1.2.17 on its portable dolly with one half of the cocler box removed. Changes of operating loops are being made as they are shut down for disassembly or for repairs. 34 Out-of-pile tests were conducted which demon- strated adequate heat-removal capacity of the salt- to-air heat exchanger designed for use with the in-pile loop that is to be operated in the MTR.'3 The temperature region near the salt freezing point I3D. B. Trauger, J. A, Conlin, and P. A. Gnadt, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 33. GE Table 1.2.1. Forced-Circulation Loop Operations Summary as of May 31, 1958 Composition Approximate Maximum Minimum Maximum Hours of Loop LOOP Number of Flow Reynolds Wall Fluid Fluid Operation at Designation Moter!a| Circulated Rate Number Temperature Temperature Temperature Corditions Comments and Size Floid* ©Pm) a1 V100°F (°F) (°F) CF) Given CPR Inconel, 3/8 in. sched 40, 122 1 5,000 1250 1095 1190 8801 Repairs being made on pump impeller 1in. OD x 0.035 in. wall, Sodium ~7 97,700 1085 1135 3, . . 7 in. 0D x 0.035 in. wall B 9344.-1 Inconel, ]/2 in. CD, 123 2 3,250 1300 1100 1210 5125 Loop shut down for 90 hr to change motor 0.045 in. wall 9354-3 INOR-8 84 2.75 1200 1070 1150 3983 Loop froze once because of power loss; Hot leg, % in. sched 40 4,500 thawed successtully Cold leg, ]/2 in. OD, 5,400 0.045 in. wall 9344.2 Inconel, l/2 in, OD, 12 2.5 8,200 1200 1000 1100 2956 Loop froze because of power loss; 0.045 in. wall repaired and restarted 9377-1 Inconel, ]/2 in. OD, 126 2.0 1,600 1300 1100 1175 3413 Loop froze because of power loss; to be 0.045 in. wall repaired and restarted 9377-2 Inconel, I/2 in. OD, 130 2.0 3,000 1300 1100 1210 3055 Loop failed when frozen because of 0.045 in. wall power loss; terminated 4/6/58 9354-1 INOR-8, 1II/2 in. OD, 126 2.5 2,000 1300 1100 1210 2106 Hestelloy pump bowl replaced with 0.045 in. wall INOR-8 pump bowl; no incidents during this period 9377-3 Inconel, ]/.z in. OD, 131 2.0 3,400 1300 1100 1215 1925 Loop froze once because of power loss; 0.045 in. wall thawed successfully 9354-2 INOR-8, 1/2 in. OD, 12 2.0 6,500 1200 1050 1140 1052 Hastelloy pump bowl replaced with 0.045 in. wall INOR-8 pump bowl and loop restarted 2/26/58; loop subsequently froze because of power loss; failure occurred during aftempt to restart 9354-5 INOR-8, 3/8 in. OD, 130 1.1 2,200 1300 1100 1205 501 Special loop containing graphite semples; 0.035 in, wall started 5/8/58 9354-4 INOR-8 130 Special loop containing INOR-8 inserts; Hot leg, 3/8 in. sched 40 fuel salt not yet available Cold leg, % in. OD, 0.045 in. wall *Compesition Composition Composition 84: Composition 122: 123: Nal":-ZrFA-UF4 (57-42-1 mele %) NaF-BeF2-UF4 {53-46-1 mele %) Nc:F-LiF-BeF2 (27-35-38 mole %) 12: NaF-LiF-KF {11.5-46.5-42 mole %) Composition 126 LiF-BeF2-UF4 (53-46-1 mole %) Composition 130: LiF-Ber-UF4 (62-37-1 mole %) Composition 131: Li|:-BeF2-UF4 (60-36-4 mole %) 8561 ‘0€ INNI ONIANI goiy3d MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT EMERGENCY CONDITIONS WHICH WILL AUTOMATICALLY PLACE LOOP ON ISOTHERMAL OR PRE HEAT CONDITIONS: 1., LOW TEMPERATURE. . HIGH TEMPERATURE. « LOSS OF TRANSFORMER POWER. . LOSS OF BOTH CLUTCH SUPPLIES. . LOSS OF PUM?P DRIVE MOTOR. . COMPLETE POWER FAILURE (DIESEL GENERATOR STARTS). . LOSS OF BUS DUCT SUPPLYING LOOP EQUIPMENT (DIESEL GENERATOR STARTS). ~N O RN MPROVEMENTS INCORPORATED [N REVISEL CORROSION LOOFP DESIGN: DOUBLE SHEAVES ON PUMP DRIVE, USE OF STEEL CORE ''v** BELTS. MAINTAINED CONTACT ON PUMP MOTOR START BUTTON. USE OF AUTOMATIC RESET ON TRANSFORMER POWER. ELIMINATION OF SLOW RESPONSE HEATER ELEMENTS ON COOLER COIL BY APPLYING DIRECT RESISTANCE HEAT AUTOMATICALLY 1IN EVENT OF AN EMERGENCY CONDITION. &, USE OF A LID ON CCOLER COIL DUCT WHICH DROPS AUTOMATIC- ALLY IN THE EVENT OF AN EMERGENCY CONDITION, @» «OO UNCLASSIFIED ORNL-LR-DWG 29715 LOOP CN ISOTHERMAL OR LOW HEAT CONDITIONS CONSISTS OF: *A. HEATER LUGS "' AND "'D'* DISCONNECTED BY 1600 AMP BREAKERTQ PROVIDE RESISTANCE HEAT THROUGH LUGS A" AND "'C!' TO ENTIRE LOOP EXCEPT COCLER COIL. B. RES|ISTANCE HEAT APPLIED DIRECTLY TO COOLER COIL BY 10 KVA POWER SUPPLY. C. COOLING AIR OFF, D. LID DROPPED ON COOLER DUCT TO ENCLOSE COOLER COIL. *NOTE: ALTERNATE METHOD, IN THE EVENT CF FAILURE OF THE 110 KVA POWER SUPPLY, AUTOMATICALLY CONNECTS CLAM SHELL HEATERS (ITEM 10) T PROVIDE SUFFICIENT HEAT TO PRE- VENT FREEZE UP. @ INSTALL A 1600 AMP CIRCUIT BREAKER TO AUTOMATICALLY SWITCH TO LOW HEAT CONNECTIONS ON LOOP PIPING IN AN EMER- GENCY. 8. IMPROVED OPERATION OF MAGNETIC CLUTCH SPEED ALARM BY USE OF HIGHER QUALITY METER RELAY AND CIRCUIT CHANGES. 9. INCREASED FUNCTION OF LOW TEMPERATURE ALARM TO INCLUDE AUTOMATIC ACTIONS FOR PREVENTION OF FREEZE UPS. PROVISIONS MADE FORTHE USE OF CLAM SHELL HEAT AUTOMATIC~ ALLY IN THE EVENT OF A FAILURE OF MAIN 110 KVA TRANSFORMER POWER. (1) PROVISIONS MADE FOR USE OF EMERGENCY DIESEL GENERATOR, IN THE EVENT OF A COMPLETE POWER FAILURE, FOR ISOTHERMAL OPER- ATION OF CORROSION LOOPS, Fig. 1.2.16. Improved Forced-Circulation Cortosion Testing Loop and Test Facilities. was investigated extensively in these tests, since low temperatures can be expected during a reactor scram. |t was determined that a control system can be designed which will permit the loop to survive a reactor scram without danger of freezing and without close operator supervision. The salt in the heat exchanger was deliberately frozen and remelted several times during the course of the experiment. 36 Prototype pump tests revealed that gravity-filling of the system will be difficult because of the surface tension of fuel 130, and two other methods of filling are being studied. One method is to fill the pump sump with small increments of fuel. This method is not compatible with filling the loop after insertion in the reactor as was done previously with other loops. Accordingly, tests are being run with the prototype pump loop to determine the PERIOD ENDING JUNE 30, 1958 URCU ASSIFIED . PHOTO 31454 L G Fig. 1.2,17, New Forced-Circulation Loop. feasibility of freezing and melting the loop. This method would permit a complete loop checkout prior to its shipment to NRTS for installation in the MTR and would eliminate the uncertainties associated with filling after complete assembly when minor repairs could no longer be made. The second method utilizes the gravity-filling system, with the filling lines modified so that the pump impeller and loop are filled before a pump sump level is established. This method was demonstrated successfully with water, The in-pile loop assembly has been completed to the point where the pump can be installed. Further assembly work will be delayed, however, until the new filling techniques are tested. All parts have been fabricated and the shielding and bulkhead sections are complete. 37 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT 1.3. ENGINEERING RESEARCH H. W. Hoffman Reactor Projects Division PHYSICAL PROPERTY MEASUREMENTS W. D. Powers Viscosity The viscosities of three beryllium-containing fluoride salt mixtures were experimentally de- termined over the temperature range 500 to 900°C. As previously described,’ the measurements were made with the use of precalibrated efflux-cup viscometers. The scatter of the data was above average (110% in one case), and therefore the data must be considered to be preliminary. Mean viscosity values at three temperatures are given in Table 1.3.1. Some of the observed deviations of the data might be explained in terms of the surface tension of the beryllium salts. There is considerable “humping’’ of the fluid surface, and thus the determination of the cup efflux time is extremely difficult. Further measurements of the three mixtures listed in Table 1.3.1 and of other salts are being made, and possible refinements of the measuring technique are being studied. Thermal Conductivity A variable-gap apparatus is being fabricated with which to measure the thermal conductivities of beryllium-containing fluoride salt mixtures. Contamination by the beryllium salt requires that the apparatus be expendable, and, hence, it must be relatively small and inexpensive. On the ]W. D. Powers, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 37, other hand, a small system increases the un- certainty in the measurement and control of the heat flow through the test sample. It is believed that the design chosen represents a satisfactory compromise between the factors of cost, ease of operation, and experimental accuracy. The apparatus will consist of two solid stainless steel cylinders, each 4 in. in diameter and 41/2 in. long, arranged so that their vertical axes are coincident. The lower cylinder will be an integral part of the sample container, while the upper cylinder will be supported and separated from the lower cylinder by a set of support rings attached to the sample container. These rings will permit adjustment of the sample thickness in 0.010-in. steps between 0 and 0.14 in. Heat generated by cartridge heaters located within the upper cylinder will flow downward through the test sample to a heat sink located below the lower cylinder. The magnitude of the heat flow will be determined by calculation from the measured temperature gradients in the two cyl- inders. Enthalpy and Heat Capacity The measurement of the enthalpy and heat capacity of the mixture LiF-BeF,-UF, (62-37-1 mole %) was completed. Values for these proper- ties may be represented by the following equations: Solid (100 to 300°C) Hp ~ Hygog = ~9.6 + 0.293T + (2.18 x 10~7) T2 c, = 0.293 + (4.36 x 10~5) T Table 1.3.1. Viscosities of Beryllium-Containing Fluoride Salt Mixtures Salt (mole %) Viscosity (cp) LiF NaF BeF UF, At 600°C At 700°C At 800°C 53 46 ] 15.2 7.4 4,2 53 46 1 20.5 10.8 6.4 62 37 1 12.2 7.2 4.7 38 Liquid (471 to 790°C) Hp ~ Hygoc = 24.2 + 0.4887 + (2.52 x 10~5) T2 c, = 0.488 + (5.04 x 10~ T In the expressions above, H = enthalpy in cal/g, c, = heat capacity in cal/g-°C, T = temperature in °C. Surface Tension Recent experiences in pumping a beryllium-con- taining mixture through a small in-pile loop and observafions made during viscosity measurements generated interest in measurements of the surface tensions of mixtures in this class of fluoride salts. An apparatus is therefore being fabricated with which to measure surface tensions by the maximum-~bubble-pressure method. The maximum pressure necessary to form o bubble at the end of o small-diameter tube immersed in the salt melt will be determined, and the surface tension will then be calculated from the formula 5 r y = p2 ’ where 8p is the pressure difference between the inside and outside of the formed bubble and r is the inside radius of the tube. The maximum- bubble-pressure method was chosen for this in- vestigation because (1) the experimental apparatus is simple and presents no design and fabrication difficulties, (2} instrumentation for measuring the pressure is available and can be readily calibrated, (3) the theory of the method is well developed and on a sound basis, and (4) o new surface is formed with each bubble so that, with a clean tip on the bubbler, surface contamination should not be a problem. Preliminary studies made with water to provide operafional experience gave the following results: Diameter of Tube Surface Tension (in.) (dynes/em) 0.026 68 0.032 70 0.046 72 The generally accepted value for the surface tension of water is 72 dynes/cm at 24°C. The PERIOD ENDING JUNE 30, 1958 difference between the accepted value and the values is believed to be due to in the measurement of the inside An error of 0.001 in. would be sufficient to account for the observed dis- crepancies. Measurements with the salt mixture NaNO,-NaNO,-KNO, (40-7-53 wt %) are currently in progress. experimental inaccuracies tube diameter. Thermal Expansion In evaluating the nuclear characteristics of a reactor, it is necessary to know the quantity of fuel present in the core at any time. For a homo- geneous reactor, such as the MSR, the fuel density, and hence the amount of fuel, is de-’ pendent on the core temperature. This relationship between the core temperature and the fuel density enters the nuclear calculations as the coefficient of thermal expansion. Variations in the values derived for this coefficient for a number of beryllium-containing mixtures from density meas- urements at ORNL and at the Mound Laboratory have indicated the need for a direct measurement of the thermal expansion of the salt mixture. An apparatus has been designed and is being fabricated with which to obtain the thermal expansion by measuring the difference in height of the fluid in the two legs of a U-tube when the legs are maintained at different temperatures. The liquid levels in the two legs will be de- termined with adjustable metallic probes inserted through the Lucite top of the inert-atmosphere box containing the apparatus. The probes will be hollow, with only the contact tip solid, since the large thermal gradient which will exist along the probe would make the length of a solid rod both variable and indeterminate. Low-expansion quartz rods will be inserted in the tubes; and the movement of the upper ends of these rods, as determined by fixed micrometers, will be used to establish the liquid-level difference in the U-tube. The initial ‘*zero’”” will be obtained by maintaining the two legs at the same constant temperature. HYDRODYNAMIC STUDIES OF MSR CORE F. E. Lynch Hydrodynamic studies of three proposed molten- salt reactor cores were continued with the use of the phosphorescent-particle flow visualization technique. The three cores being investigated 39 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT differ only in the associated entrance-exit systems which provide (1) straight-through flow with diametrically opposed entrance and exit (MSR-1), (2) concentric entrance-exit flow with fluid entering through the inner pipe and exiting through the outer annulus (MSR-2), and (3) concentric entrance- exit flow with fluid entering through the annulus and exiting through the inner pipe (MSR-3). The physical details of these models and the charac- teristics of flow of the second type mentioned above were described in the previous report.? The results of full-field observations made with a planar light source for the straight-through flow model (MSR-1) are given in Fig. 1.3.1. The inlet Reynolds modulus (based on the inside diameter of the inlet pipe) was 74,000, which corresponds to a flow rate of 1.38 ft/sec. It may be seen 2k, E. Lynch, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 35. UNCL ASSIFIED ORNL—LR--DWG 31200 T -1 /// | ! VELOCITY NEXT i TO THE WALL 1S £ VERY LOW @ /PARTICLE MOTION IN THIS REGION 1S VERY UNSTABLE Fig. 1.3.1, Flow Patterns Observed Within an lllumi- nated Plane of the Straight-Through Core Flow Medel {MSR-1). 40 that the main high-velocity flow short-circuited the core and passed directly from the entrance to the exit with essentially no spreading. The remainder of the core was filled with slowly rotating fluid that exhibited low-velocity flow upward along the sphere surface and extremely low-velocity flow near the core exit. This rotating mass displayed some instability (perhaps as- sociated with gross pressure fluctuations in the system), with smaller eddies appearing as indi- cated by the dashed lines of Fig. 1.3.1. In- stantaneous velocity profiles for this same flow are shown in Fig. 1.3.2 at the indicated axial positions along the core. The flatness of the velocity profile through the central jet is apparent at all positions. The distortion of the exciting beam and the profile, particularly noticeable in (@), (b), and (f) of Fig. 1.3.2, resulted from the curvature of the sphere. Results were also obtained for flow in the third type of model described above (MSR-3). A photograph of the experimental system is shown in Fig. 1.3.3. The central pipe was flared outward at the top of the core so that the fluid entering through the outer annulus was directed along the sphere surface. For this system, the inlet Reynolds modulus (based on the diameter of the inlet side arm) was 70,000, which corresponds to a flow rate of 1.3 ft/sec. The gross velocity patterns obtained through visual observation of particle motion within an itluminated plane of MSR-3 are shown in Figs. 1.3.4 to 1.3.6. The complexity of the flow is only partly indicated by these figures. The ob- served patterns were generated, not by the annular entrance per se, but rather by the manner in which the fluid entered the annulus, as is evident in Fig. 1.3.4, which shows that the fluid flowed past the central pipe rather than parallel to it. Lengthening the annular region or installing turning vanes should correct this condition. Such changes are to be studied in future tests. The inset circle of Fig. 1.3.4 outlines the nature of an off-center rotating eddy formed by the asymmetrical inlet flow. When observed from above (see Fig. 1.3.5), this eddy was found to consist of two separate vortices, one located 120 deg from the inlet and the other slightly less than 90 deg. Both tapered toward the center and joined intermittently. The outer face of the eddy did not contact the sphere surface. At Ly Fig. 1.3.2, Instantaneous-Velocity-Profile Photographs in Straight-Through Core Flow Model (MSR-1). UNCLASSIFIED ORNL-LR-DWG 31201 8561 '0f ANNI 9NIGNT @0i¥d3d MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT Fig. 103130 Designed for Fluid Entrance Through Outer Annulus, Concentric Entrance-Exit Model (MSR+3) higher velocities, the vortices combined and the eddy assumed a position perpendicular to the inlet flow., Other aspects of the flow in this geometry are shown in Fig. 1.3.6, in which the plane of illumination includes the center of the sphere. The inlet flow imparted a general rotation to the entire fluid mass. Further, it was observed that at the exit the flow split and some of the fluid recirculated through the core. A number of stagnant regions were noted. One was found off the trailing edge of the inlet flare at a position 180 deg from the inlet where the incoming and recirculating flows combined. Another, found within the inlet flare region, was due to bypassing of the inlet flow. This stagnant region extended around the annulus to 90 deg on each side of the entrance. A third stagnant region was found close to the sphere center in the region of the eddy. As is also shown in Fig. 1.3.6, it was found that the axis of the eddy rotated. A series of instantaneous photographs obtained for this geometry are presented in Figs. 1.3.7 and 1.3.8, which were velocity profile 42 UNCLASSIFIED ORNL-LR—DWG 3{202 STAGNANT REGION — —t——— / s — e e o —— / - - . FLOW N T / - — . - - - g \A—FLOW PATTERN | WITHIN A ZONE OF THE SPHERE Fig. 1.3.4, Flow Patterns Within the Annular Entrance and a Zone of MSR-3 As Observed in an llluminated Plane. UNCLASSIFIED ORNL—LR—~DWG 34203 AXIS OF ROTATION COF / EQUATORIAL EDDY ! CENTRAL PIPE OUTER LIP OF FLARED CENTRAL PIPE INLET SIDE-ARM AXIS OF ROTATION OF / EQUATORIAL EDDY- Fig. 1.3.5. Flow Patterns Within MSR-3 As Observed from Above. UNCLASSIFIED ORNL-LR-DWG 31204 INTERMITTENT STAGNANT REGION WHICH FLUCTUATED AS INDICATED BY ARROWSj ~STAGNANT REGION EXTENDED ARQUND THE ANNULUS TO 30 deg ON EACH SIDE OF INLET -EDDY ROTATED ABOUT AX[S SHOWN BY DASHED LINES. THE AXIS CF ROTATION OF THE INTER- MITTENT STAGNANT REGION WAS THE QUTER EDGE OF THE EDDY. Fig. 1.3.6. Flow Patterns Within MSR=3 As Visualized in an llluminated Plane. taken at positions 90 deg apart. The downward flow along the sphere wall in the region opposite the inlet is apparent in Fig. 1.3.8, while Fig. 1.3.7 gives evidence of the upward surface flow in the region closer to the inlet. Further, Fig. 1.3.8, (&) through (d), shows clearly the eddy described in Fig. 1.3.5. Recent studies by Taylor? of flow in a sphere with a central pipe entrance and an annular exit have shown that the depth of penetration of the inlet stream is a function of the extent of projection of the central pipe into the sphere. A bubble-photograph technique was used in these studies., Taylor also investigated the effect of flaring the inlet pipe. With no projection into the sphere, it was found that the inlet stream did not reach the bottom of the sphere. This condition persisted until the projection extended to one-fourth of an inlet-pipe diameter. Con- siderable vibration was observed when flared tubes were used. The amplitude of vibration increased for tube projections of one-fourth diameter or less and also varied with the ex- pansion being greater with a 40-deg 3A. F. Taylor, Preliminary Studies of the Flow Patterns in Two Designs of Homogeneous Test Reactor Core. Part I. Re-entrant Core, IGR-TN/CA-324 (sub- ref HARD(A)/P-12) (March 23, 1956). angle, PERIOD ENDING JUNE 30, 1958 included-angle flare than with a 20-deg flare. Vibrations with the straight tube occurred only for projections of less than one-fourth diameter. The model shown in Fig. 1.3.9 (MSR-2A) was fabricated in order to study the effect of extending the central pipe in the MSR-2 model. The central pipe was extended three-fourths of a pipe diameter into the sphere (compared with one-fourth diameter for MSR-2), and the flare at the top of the sphere in the region of the join with the exit pipe was increased. Bubble photographs of the over-all flow are shown in Fig. 1.3.10. The inlet Reynolds modulus was 118,000, which corresponds to a flow rate of 3.25 ft/sec in the inlet pipe. An unsteadiness in the flow pattern may be seen by comparing the two photographs of Fig. 1.3.10. In Fig. 1.3.104, a well-defined eddy may be seen in the right-hand region of the sphere, while the left side shows a number of smaller eddies and more random patterns. In Fig. 1.3.105, taken a short time later, the patterns are reversed. |t was also noted that the three-fourths-diameter projection gave a higher velocity at the bottom of the sphere than did the original one-fourth- diameter projection, A series of instantaneous velocity-profile photographs obtained with this model at a Reynolds modulus of 108,000 (flow rate of 3 ft/sec) are shown in the consecutive exposure reproduced in Fig. 1.3.11. The points of excitation are indicated by arrows, MOLTEN-SALT HEAT TRANSFER STUDIES H. W. Hoffman A preliminary series of tests was completed with the salt mixture LiF-Ber-UF4 (53-46-1 mole %) flowing through an electrical-resistance- heated small-diameter Inconel tube. Typical results are shown in Fig. 1.3.12, where the data are presented in terms of the Colburn j-factor. It may be seen that the limited results closely agree with the correlation, j = 0.023NES'2, for ordinary fluids. Some difficulty was experienced in establishing a satisfactory heat balance for the system, and experiments are under way to diagnose and correct this difficulty. An existing system that contains a pump is to be modified so that long-time exposure data can be obtained with this salt mixture. Long-time exposure will be required in order to determine what adverse conditions, such as those which would arise as a result of film formation, will exist in this system. 43 vy (c) Fig. 1.3.7. & CAMERA POSITION ~ + _—_ ANGLE OF EXCITATION Instantaneous«Velocity-Profile Photographs in MSR-3. (d) UNCLASSIFIED ORNL—LR—DWG 31205 1Y0dIY SSIYO0dd WYHO0dd ¥yOLIvIY LAVS-NILTIOW )4 ANGLE OF EXCITATION (¢) 4 CAMERA POSITION Fig. 1.3.8, Instantaneous«Velocity-Profile Photographs UNCLASSIFIED ORNL—LR—DWG 31206 {6} (d) in MSR-3. Photographs taken 90 deg from those of Fig. 1.3.7. 2561 "0 INNT 9NIANT @ol¥3d MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED PHOTO 34310 Fig. 1.3.9. Modified MSR-2 Central-Pipe-Inlet and Annular-Exit Flow Model (MSR-2A). 46 Fig, 1.3.11, Consecutive Profile Photographs of Flow in MSR-2A. UNCLASSIFIED PHOTC 44497 Instantaneous-Yelocity- PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED PHOTO 44496 'Bi UNCLASSIFIED ORNL—LR-—-DWG 31207 2 2t Nep Mp, r © o O o T 0.002 — | COLBURN FACTOR, j S ‘ Lo ‘ ooot b o v o 1. 2 x 103 5 10% ? 3x10% REYNOLDS MODULUS, A Fig. 1.3.12. Heat Transfer with LiF-Ber-UF4 (53-46-1 Meole %). 47 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT 1.4. INSTRUMENTATION AND CONTROLS H. J. Metz Instrumentation and Controls Division ENDURANCE AND STABILITY TESTS OF SHEATHED THERMOCOUPLES C. M, Burton fnconel-sheathed, 0,250-in,-OD, Chromel-Alumel thermocouples with magnesium oxide insulation and hot-junction closure welds made by the Heliarc welding process are being tested in order to determine the expected accurate life of a thermocouple in service and to determine the effect of a liquid metal or a molten-salt environment on the integrity of the closure weld. More than 10,000 hr have been accumulated in a test in which sheathed thermocouples are immersed in sodium held at 1500°F, The temperature of the sodium bath is reduced once each week to 1300°F and to 1100°F for readings at these temperatures. Of the 38 thermocouples included in this test, only 2 have failed because of weld closure deficiencies. Although the temperature readings from the various thermocouples being tested do not correspond exactly to the reading from the calibrated standard thermocouple as a result of individual variations and of different positions in the pot of heated sodium, drifts from the initial temperature readings are within +0.75%. RESISTANCE-TYPE FUEL LEVEL INDICATOR R. F. Hyland Test facilities were prepared for an investigation of the suitability for molten-salt reactor service of the resistance-type fuel level indicator. Two Inconel elements of the type shown in Fig. 1.4.1 were fabricated for these tests without temperature compensation. Tests of these elements in fuel 130 (LiF-Ber-UF4, 62-37-1 mole %) will provide the INTERNAL STRUCTURE data needed for design of temperature-compensated INOR-8 elements and indications of the long-term stability of the elements., SCANNING SWITCHES A, M, Leppert Design work is under way on modifications required to improve the Delta switch manufactured by the Detroit Controls Corp. for use in switching low-level transducer signals. Switches of this type are useful for monitoring transducer signals for alarm purposes and for commutating, transmitting over a pair of wires, and decommutating as many as 80 signals. The modified units are to have stainless steel housings, and the individual con- tacts will be insulated with Ceramicite.,! The present units have Bakelite bodies which hoid and insulate the individual contacts. The modifications of this mercury-jet type of switch are being made to increase bearing life, reduce the mercury con- tamination, and eliminate electrostatic charges during operation, PRESSURE TRANSDUCERS J. W. Krewson A test program was initiated for determining the suitability of various types of pressure trans- ducers for use in a reactor containment cell where freedom from drift, errors, and failure would be essential, The available instruments are being tested under a single pressure condition in order to determine their static drift behavior, and other instruments are to be obtained. The relationship between temperature and driftisalso being studied. 1 Ceramicite is manufactured by Consolidated Electro- dynamics Corp. UNCLASSIFIED PHOTO 29725 Fig. 1.4.1, Resistance-Type Fuel Level Probe, 48 PERIOD ENDING JUNE 30, 1958 1.5. ADVANCED REACTOR DESIGN STUDIES H. G. MacPherson Reactor Projects Division AN EXPERIMENTAL 5-Mw THERMAL-CONVECTION REACTOR J. Zasler The history of reactor technology has indicated that the development and demonstration of a reactor concept requires the operation of o small experimental reactor and a medium-sized pilot plant, In the case of molten-salt power reactor development, it appears that the simplest and most reliable experimental reactor would be based on thermal convection of the fuel. A design study has indicated' that a 5-Mw thermal-convection reactor would be large enough to provide experi- mental data and yet be small enough to keep the chief disadvantages of the thermal-convection system - excessive fue! volume and large heat exchangers — from being major factors. Further, the 5-Mw thermal-convection reactor could be converted to a 50-Mw pilot plant by adding a fuel pump and increasing the capacity of the heat dump. Proposed layouts for the 5-Mw reactor system are shown in Figs. 1.5.1 and 1.5.2, and the dimensions and operating conditions for 5- and 50-Mw service are given in Table 1,5.1. Provisions would be included for connecting the blanket and fuel regions so that the reactor could be operated 1], Zasler, Experimental 5 Mw Thermal Convection Molten Salt Reactor, ORNL CF<58+6-66 (June 13,1958). UNCLASSIFIED ORNL-LR-DWG 31208 a i 2 .. 3 AIR REACTOR Lo - LOCK CELL L | MAINTENANCE - AREA DO MANIPULATOR - SODIUM PUMP [ TURBINE-GENERATOR ROOM {FOR FUTURE EXPANSION) 5-Mw RADIATOR FUEL-TO-SODIUM & HEAT EXCHANGER ] F @ o o» i FUEL EXPANSION TANK — ~CUTTING PLANES FOR INSTALLING e et : e 50-Mw SODIUM SYSTEM 1 O a . a ’ TR T ™ 1. """ BLANKET-TO-SCDIUM L e S b K EXCHAN’GER ‘- :1.:‘ e @l \ /é\\‘z‘\\\ i W 7 S S ";_somum DRAIN TANK 5 0 5 10 15 20 e — e n—] FEET Fig. 1.5.1. Elevation Drawing of 5«Mw Experimental Molten-Salt Reactor Plant, 49 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 31209 CONTROL ROOM - ~x TURBINE GENERATOR ROOM (FOR FUTURE EXPANSION] FUEL-TO-SODIUM & - HOT HEAT EXCHANGERE: - .1 MAINTENANCE R AREA REACTOR ¥ ! MAINTENANCE BLANKET-TO- SODIUM PUMP 5-Mw o 84 f1, "IRADIATOR ‘ AREA SODIUM HEAT CH T BLOWER 5 0 5 10 15 20 e e Fig. 1.5.2, Plan View of 5-Mw Experimental Moliten-Salt Reactor Plant, as a one-region system, The 5Mw reactor en- visioned could be constructed of components already developed. [n order to provide for 50-Mw operation, the fuel expansion tank was designed so that a sump-type fuel pump could be installed in it, and the sodium lines leading to the heat exchanger were sized to handle the flow required in the 50-Mw system, The 5-Mw reactor would serve to demonstrate the feasibility of continuous operation of a molten- salt reactor, to provide in-pile corrosion data on removable samples in hot and cold legs, to develop and demonstrate remote maintenance procedures, and, by replacing the air heat dump with a steam heat dump, to demonstrate sodium-to-steam heat transfer. When the experiment had been completed, the system could be converted for 50-Mw operation by installing a sump-type fuel pump in the fuel expansion tank, a new sodium system consisting of a 10,000-gpm pump and a sodium-to-steam heat exchanger, and a turbine-generator system. The fuel-to-sodium heat exchanger designed for 5-Mw 50 operation would be satisfactory for 50-Mw operation because of the reduction in the fuel film resistance that would occur as a result of going from laminar flow at the lower power to turbulent flow at the higher power. Calculations indicate that the bianket circuit could be designed so that thermal convection would provide adequate circulation of the salt at both power levels, but if necessary a pump could be installed, Although such a plant would not be identical with a large-scale power- producing plant, there would be sufficient similarity, especially with respect to control, corrosion, and maintenance problems, so that successful operation would lead directly to the design and construction of a large power plant. A rough estimate has indicated that the 5-Mw plant would cost about $10,000,000, and that for an additional $10,000,000 it could be converted to the 50-Mw plant, The arbitrary power levels and estimated costs would, of course, be subject to change if optimization studies were carried out for these systems. PERIOD ENDING JUNE 30, 1958 Toble 1,5.1, Dimensions and Operating Conditions for Experimental 5~ and 50-Mw Molten e Ll 0I5 . . v A ’ & - fi L M ‘I, S %A . & ok cedvd | x T - x k. = » < e Fre %g “@.' y v o - L t : " Ty, e O = " éi ¢ e-_ o e g0 P ? {’” -y i W0 ; + .?( . F‘w - v..-. h“ : " .% o « . S - S o — oy . ‘- :“: P A w R -‘. " :;. a“?'...’ < s c*a Fig. 2.1.3. Hot 5t : sy e Y & v, . - . B v e Y it oo . . - - 5 e, : ; , * 7 7 - ¥ . i v M . .. . -~ * % T ¢ . f . !y Ty e A L q‘f" w0y : 1 . . . . . ‘ . L LI . Y e e ™, -t §‘ - : . 3 . - o, - N . ey g € L . ’ - e < ”~. . s s swem » L yooot o« o« 1.003F - R W o - b - . £ . ¥ R o oAt ¥ o L £ ™ R e - * - % . i % v ’ L £ s s % - e e L e ¥ " - *~ ; g i e . v i ; . £ - N — [ e i . sy ¥ 5 ¢ ; . % « ! ¥ - : . : % . - 7 £ # 3 " e 1 * * - g * ; . § ~, \ s - Sl s Dy s e Lo e e - C * T 4 » L ) 4 v "o e o » : 00 £ # o - o F 2 ) . N e - n‘\ £ e v tE7 - ) - . e ¢ f - = 5 o ¥ * 5 ‘ £ LA * L . gP-:. et - }‘_ . L 1\ . # & " s £ * . § « P L . % U ¥ . F { { s r ~ - ki e ., 4 . * k] ! - T - . # E . 4 * . i e -~ - ‘ ) * ¥ . 008 .. N e -, » - . ™ + . T e, ' 4 . i * o e e ~ v ow w “ . . \ , ) * - % & » ¥ * . - " . : i ¥ * - " =~ g ¥ o * % & i * . - - - “q * e, * > T - - . - Al . - ¥ - T . 008 « £ -~ .. sy ! i - L7 : o - ’ . ¥ > ) 2 N ¢ ' . % AN ¢ ¢ 5 * - e , o owm ] e . ¥ e r ; - . o : - % -~ * ki - . i . e s - e F . ¢ # - - n * ; ,3’ N : A % * J s ‘ . i - . . e .o08 ’ i . ; - : # - S : | - » ! * - i * . * g * -~ 1 ~ Lo9 () ' ' | (b} “ Fign 2.].5- Specimens 0,040 in, Thick of (a) Inconel and (b) INOR-8 (Heat SP-16) After Exposure in a Sodiums- Graphite System for 40 hr at 1600°F, Etchant: copper regia, 500X. Table 2.1.4. Carbon Analyses of Millings Taken from 0.25-in.=Thick Bars of Inconel and INOR-8 Carburized in a Sodium-Graphite System for 40 hr at 1600° F Sample Carbon Found (wt %) INOR-8 Specimen Inconel Specimen Millings from cross section First 0,010 in. from surface {(0-0.010 in.) Second 0.010 in. from surface (0.010-0.02Q in.) 0.04 0.06 0.25 0.33 0.09 0.08 61 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT agreement, values are presented in Table 2.1.5. average As may be seen in Table 2.1.5, carburization of the Inconel increased its tensile strength and yield strength and reduced its ductility. The same carburizing treatment lowered the tensile strength, slightly increased the yield strength, and greatly reduced the ductility of INOR-8 (heat SP-16). Apparatys is being prepared for studies of INOR-8 and Inconel in contact with a graphite and molten salt system. Brazing Alloys in Contact with Molten Salts E. E. Hoffman D. H. Jansen Two precious-metal-base brazing alloys showed promising corrosion resistance to LiF-Ber-UF4 (62-37-1 mole %, fuel 130) in preliminary tests,3 and therefore other commercially available brazing alloys having high gold or silver contents are currently being tested at 1300°F. The compositions of the alloys being tested and their approximate brazing temperatures are listed in Table 2.1.6. The solidus points of all these alloys are well above 1250°F, the highest temperature expected in a molten~salt power reactor. In order to obtain long-term dynamic corrosion data on brazing alloys in fuel 130, a series of tfive alloys will be tested in duplicate by inserting brazed lap in the hot legs of thermal- The designations of the alloys joints convection loops. 3D. H. Jansen, MSR Quar. Prog. Rep. Jan, 31, 1958, ORNL-2474, p 59. that will be tested and their compositions are given below: Composition Coast Metals alloy Nos» 52 89% Ni—5% Si—-4% B-2% Fe Coast Metals alloy Nos 53 81% Ni—=8% Cr—4% Si—-4% B-3% Fe General Electric alloy 70% Ni=20% Cr-10% Si NO. 8] Au=-Ni 82% Au—18% Ni Copper 100% Cu The configuration that will be used for their insertion in the hot leg of a thermal-convection loop is shown in Fig. 2.1.6. The loops are ready for filling and should be in operation in the near future. A series of three identical loops is to be Table 2,1.6. Precious-Metal-Base Alloys Being Tested in Static Fuel 130 Test temperature: 1300°F Test duration: 500 hr Brazing Temperature Material Range (°F) Pure silver 1760-1900 90% Ag—10% Cu (coin silver) 16001850 72% Ag-28% Cu 1435-1650 50% Ag—33.3% Au-16.7% Cu 15251600 75% Au~20% Cu-5% Ag 1650~1850 42% Au—40% Ag-18% Cu=0.6% Zr 15081600 Table 2,1.5. Results of Tensile Tests of Carburized and Control Specimens of Inconel and INOR-8 Tensile Strength Yield Strength at 0,2% Offset Elongation Material (psi) {psi) {% in 2-in, gage) Inconel Control 75,878 20,906 47.25 Carburized 91,532 28,577 25,67 INOR-8 Control 116,338 50,542 49,50 Carburized 98,535 52,828 7.75 62 PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED ORNL-LR-DWG 29965 SPECIMEN HANGER COLD LEG—mm SPECIMEN (TEN) Fig. 2.1.6. Configuration for Tests of Brazing Alloy Specimens in the Hot Leg of a Thermal-Convection Loop, 63 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT operated for 1000-hr, 5000-hr, and 10,000-hr periods before the specimens are removed for examination. MECHANICAL PROPERTIES OF D. A. Douglas The tensile properties of INOR-8 at various temperatures were measured at this laboratory and at Haynes Stellite Company on material from similar heats that contained 0.02% carbon and at Battelle Memorial Institute (BMI) on material from a heat that contained 0.06% carbon. The ORNL data are presented in Fig. 2.1.7 and are compared in Figs. 2.1.8 and 2.1.9 with the Haynes Stellite and BMI data. As may be seen from Fig. 2.1.8, there is good correlation of the data. The BMI results were expected to be slightly higher than the others as a result of the higher carbon content. The yield strength shows the expected trend, but the tensile strength at 1250°F appears to be low. The elongations measured at rupture in the same experiments are shown in Fig. 2.1.9. The relatively brittle failure which occurred in the INOR-8 UNCLASSIFIED BMI test at 1250°F explains the low tensile strength reported in Fig. 2.1.8. The temperature dependence of Young's modulus for INOR-8 is shown in Fig. 2.1.10, Relaxation curves obtained from data taken at 1200, 1300, 1400, 1500, and 1600°F are presented in Fig. 2.1.11. Considerable relaxation occurs at 1400°F and above within the first 6 min of the (x 10%) ORNL—LR-DWG 28709R 110 10 l 17 [ 1 TENSILE STRENGTH | E } ‘ 100 - | | L 100 | i ‘ ‘ ‘ : | \ | . . 90 b—— + : } 90 ; ; | 80 —————— - ,,J_‘ - N —1 80 ‘ i - 70 : Lo -+ | - 70 a2 . ELONGATION (%) ! . 0 - T ’ 32 2 N\ = —- i -— 60 2 w 60 f ‘ I - ) o ” % 50 40 30 {20 | i i S | | ) 0 400 800 1200 TEMPERATURE (°F) 1600 Fig. 2.1.7. Tensile Properties of INOR-8 as a Function of Temperature. 64 test., At 1300°F and below, incubation periods of (107} UNCLASSIFIED 130 ORNL-LR-DWG 31214 ! T T ] T ] T T T 120 [-B~k—-+ ULTIMATE TENSILE STRENGTH—————r—F rfi . . ; ; \"“‘*\! 1 ! | ‘ | : | "o -"'"-a__‘ : ‘[ | i | —_— ~ ST S | 100 [— ORNL ; =] . - 50 ] ! >~ g HAYNES STELLIT = o~ ' . 2 : L~ 5 80 - ; f f w , € 70 : Lo L | ol ? 60 ‘ S - SN | ' 50 +——+——0.2% OFFSET YIELD STRENGTH —— | SSSSes=mmgo e[ B e —— e 30 | HAYNES .~ A RS e 20 i L | 4 o 200 400 600 800 1000 1200 1400 TEMPERATURE (°F) Fig. 2.1.8, Comparison of Tensile Properties of INOR-8 Obtained at ORNL, Haynes Stellite Company, and Battelle Memorial Institute (BMI). UNCLASSIFIED ORNL-LR-DWG 31212 68 '*jfi BMI W e o o o e - HAYNES STELLIT ELONGATION AT RUPTURE (%} Q 200 400 600 800 1000 {200 1400 TEMPERATURE (°F) Fig. 2.1,9. Comparisons of Tensile Elongations of INOR-8 Specimens Tested at ORNL, Haynes Stellite Company, and Battelle Memorial Institute, 1 hr and longer occur before creep commences. It appears that, even at 1200°F, INOR-8 will eventually relax to a stress of 2000 to 3000 psi. Creep tests of INOR-8 (heat SP-16) are under way at temperatures ranging from 1100 to 1500°F with fuel 107 (NoF-KF-LiF-UF,, 11.2-41-45.3-2.5 mole %). The data obtained thus far are plotted in Fig. 2.1.12 os creep rate vs stress for temperatures UNCLASSIFIED ORNL-LR-DWG 29238 — 1 1T 17 1T 7T 7T 1 26 YOUNG'S MODULUS n 5 na N L1 | 1 0 200 400 €600 800 1000 1200 1400 TEMPERATURE (°F) A 1L ™ o PERIOD ENDING JUNE 30, 1958 of 1100, 1200, and 1300°F. Stress-rupture and creep results obtained in tests in air at Haynes Stellite Company are presented in Fig. 2.1.13, and data obtained at ORNL in molten salts under similar test conditions are presented in Fig.2.1.14. The correlation of the data is excellent. UNCLASSIFIED ORNL-LR-DWG 31213 (X 103) 26 — o4 1200°F 25 [ 4300°F! L = ‘ B 20 1£}OO°F. f [T T : 18 i ‘ el =16 ‘ , ‘ ey @ 1500°F | T~ : : | ;;‘4 ! NG J”—rl - o1 j B : ‘ — T40 o 8 6 4 2 0 o1 1o 20 50 iC0 Fig, 2.1,10. Young’'s Modulus for INOR-8 as a Fig. 2.1.11. Relaxation Curves for INOR-8 Stressed Function of Temperature, to 0,1% Strain at Temperatures from 1200 to 1600°F. UNCLASSIFIED 50,000 OTNL—LRI—DWG 312‘;4 ' T | Loy | | L T o T 40,000 —[ l ’ ll ( ’ } J— { I ‘ | | l / 35,000’-——fl {H / ’ { | | J 30,000 %—— | 1 i ( | { { W 1100°F | { 1200°F ‘ = 25,000 b _J’_ 2 . / & 20,000 / | | /| v V] A /| 15,000 [T ’7 o L L 108 2 5 1074 2 5 1073 2 5 1078 2 5 10~ MINIMUM CREEP RATE (% /hr} Fig. 2.1.12, Creep Rates vs Stress for INOR-8 Tested at Various Temperatures in a Molten-Salt Environment, 65 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT (X 10%) UNCLASSIFIED 100 ORNL—LR—DWG 34215 50 20 STRESS(psi) 1 2 5 10 20 50 100 200 500 1000 TIME (hr) Fig. 2.1.13. INOR-8 Creep and Stress-Rupture Data Obtained in Air by Haynes Stellite Company. UNCLASSIFIED STRESS (psi) 10 yr | o2 L LT VL L | 1yr§ | 1 10 100 1000 10,000 TIME (hr) TO 1% STRAIN 100,000 Fig. 2.1.14, Tested at Time to 1% Strain vs Stress for INOR-8 Yatious Temperatures in a Molten-Salt Environment, FABRICATION STUDIES J. H. Coobs H. Inouye T. K. Roche Influence of Composition on Properties of INOR-8 T. K. Roche An investigation was completed of the influence of composition variation on the creep-rupture strength at 1500°F and the microstructure of alloys containing 10-20% Mo, 5~10% Cr, 4-10% Fe, 0.5% Al, 0.5% Mn, 0.06% C, and the balance nickel. A report has been prepared that presents the details of this investigation,4 and the results are summarized below. The compositions of the alloys studied were varied systematically in an effort to determine the effect of each element on the alloy strength. Creep-rupture tests at a stress of 10,000 psi were run on all the alloys in the annealed condition. The criteria used to evaluate the strengths of the 66 alloys were the times required to produce strains between 1 and 10%. The results could not be explained simply in terms of composition variation, since the principal factors affecting the strength of the alloys were solid-solution elements, carbide and noncarbide aging reactions, the presence of M6C-fype carbides in the microstructures, and grain size. From the standpoint of creep-rupture strength, it was possible to conveniently group the alloys according to the three concentrations of molybdenum studied: 10, 15, and 20%. It could be concluded from the chemical and microstructural analyses that the relative contribution of each of the factors to the strength of the alloy varied among the groups. The combined effects of solid-solution strength- ening by molybdenum and the increase in quantity of dispersed M C-type carbides which this element promoted in the annealed materials were the predominant factors which increased the strength of the alloys grouped by molybdenum content. The only exception noted was the 20% Mo-7% Cr-10% Fe alloy, which precipitated a noncarbide phase as a consequence of crossing a new phase boundary. This phase contributed noticeably to creep-rupture strength in the later stages of testing. The contributions of chromium and iron to the strengths of the alloys within the individual groups could not be established with certainty because of simultaneous variations in other factors affecting creep-rupture behavior. In order to obtain a better indication of the strengthening influence of chromium and iron, creep-rupture studies were conducted on low-carbon ‘‘high-purity’’ alloys. Although the analysis of the data was complicated by the presence of a limited amount of carbide precipitation and by grain-size variations, the influence of chromium was found to be significant when 5 to 10% was added to a 15% Mo-bal Ni base, with the presence of 10% chromium in the base composition resulting in the most pronounced strengthening influence. The strengthening effect of iron was interpreted as being insignificant when amounts up to 10% were added to a 15% Mo-7% Cr—bal Ni base. A general consideration of all the data obtained from this investigation favorably 41, K. Roche, The Influence of Composition Upon the 1500°F Creep-Rupture Strength and Microstructure of Molybdenum-Chromium-Iron-Nickel Base Alloys, ORNL-2524 (June 24, 1958). supported the composition specification selected for the alloy INOR-8. High-Temperature Stability of INOR-8 H. lnouye Embrittiement studies of INOR-8 in the temper- ature range of 1000-1400°F were completed for aging times up to 2000 hr. Data from tensile tests of such specimens indicate that aging produces no significant changes in the properties of the alloy. Two heats containing different carbon PERIOD ENDING JUNE 30, 1958 contents were found, however, to have significantly different properties. The lower carbon content alloy (heat SP-19, 0.06% C) behaved in a more ductile manner than the higher carbon content alloy (heat 8M-1, 0.14% C). The data obtained in the tests are compared in Figs. 2.1.15 and 2.1.16 with data for Hastelloy B, whose tendency to embrittle was previously discussed.’ SR. E. Clausing, P. Patriarca, and W. D. Manly, Aging Characteristics of Hastelloy B, ORNL-2314 (July 30, 1957). UNCLASSIFIED ORNL—-LR—DWG 29617 160 140 120 100 o AGED 500 hr m AGED 2000 hr TENSILE STRENGTH (1000 psi) A AGED 1000 hr 0o AGED 2000 hr ———HAST B — v AGED 500 hr A AGED 1000 hr ELONGATION (%) 1000 1100 1200 1300 1500 AGING TEMPERATURE (°F) Fig. 2.1.15. Room-Temperature Tensile Properties of INOR-8 and Hastelloy B After Aging in the Temperature Range 1000-1400°F for Times Up to 2000 hr. 67 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT Status of Production of INOR-8 (Westinghouse Subcontract 1067) H. Inouye Five air-melted heats of INOR-8 consisting of about 5000 |b each were prepared at the Blairsville Metal Plant of the Westinghouse Electric Corp. About 20,000 |b of finished products are to be furnished from these heats, Status of Production of Seamless Tubing (Superior Tube Company Subcontract 1112) H. [nouye The first shipment of seamless INOR-8 tubing (0.500 in. OD, 0.045 in. wall) was received during the quarter. It consisted of approximately 150 f1 of tubing which was processed by the Superior Tube Company's production department without UNCLASSIFIED ORNL—LR—DWG 29616 140 | | | | S HEAT 8M4// ———HAST B | . v ANNEALED x ANNEALED 0 ANNEALED '3420 -0 AGED 500 hr ® AGED 500 hr vV AGED 500 hr o ¢ AGED 1000 hr A AGED 1000 hr A AGED 1000 hr S m AGED 2000 hr 0 AGED 2000 hr —— £ 100 ’i%%?‘:‘« ‘ \ v mme o GED 5 sua : b e i~ =z SISO RS - N SO N lfig [ \\ - — 80 w) wl - 2 N\ = 60 =~ 40 3 = SRR /57 =z K S L~ 7 o R e S R R AT S 7 5 v vy o - . QTR = —— S —— o ’-—"T ———— ——::: .—--—-—‘ 1000 1100 1200 1300 1400 1500 TEST TEMPERATURE (°F) Fig. 2.1.16. Elevated-Temperature Tensile Properties ature Range 1000~1400°F for Times Up to 2000 hr, 68 of INOR-8 and Hastelloy B After Aging in the Temper- difficulty. This was the first production run of seamless tubing by a commercial vendor. The production involved an extrusion from a 9-in.-dia drilled billet to a 4-in.-OD, ]/z-in.-wall tube shell by the International Nickel Company (on a purchase order), reduction by Babcock & Wilcox to a 2-in.-0OD, 0.187-in.-wall tube, and tube reduction and re- drawing by the Superior Tube Company to the final size. Despite the circuitous route involving four manufacturers, the tubing appears to be of excellent quality. (The inspection of this material is described in a subsequent section of this chapter.) Bearing Materials J. H. Coobs INOR«8 Journals, — Two 3-in.-dia journals of INOR-8 were flame-sprayed with INOR-8 for use in bearing tests. One journal was built up with 5 mils of sprayed coating and the second with 20 mils. The journals were first sprayed with a light layer (~2 mils) of molybdenum to provide a rough bonding surface, and after being sprayed with INOR-8 were heat-treated at 2250°F to bond the coatings. During the heat treatment, the 5-mil coating on the first journal separated at one end. This journal was recoated after being rough-machined to provide a better bonding surface. The final coating was more than 25 mils thick and was successfully applied and finished. These results indicate that the thickness of the sprayed coating for a 3-in.-dia journal should be a minimum of 20 mils. An experiment was also run in which an INOR-8 journal was flame-sprayed with molybdenum. A sample of ]'/2-in.-dia INOR-8 pipe was threaded to provide a good bonding surface, spray-coated with about 20 mils of molybdenum, and subjected to thermal cycling. The great difference between the thermal expansion of INOR-8 and that of molybdenum resulted in severe cracking of the coating during the first thermal cycle. After 23 additional thermal cycles from room temperature to 1350°F, the molybdenum layer was partially separated from the INOR-8 base. The large cracks that developed in the coating during the first thermal cycle and the area where the coating broke away from the INOR-8 pipe may be seen in Fig. 2.1.17. Molybdenum Journals. — Several l-in.-dia speci- mens of molybdenum rod were sprayed with molybdenum in an effort to develop a suitable procedure for spray-coating molybdenum journals. PERIOD ENDING JUNE 30, 1958 Preliminary tests indicated that a surface which had been grit-blasted and etched was most suitable for bonding, and that a high-temperature heat treatment was needed for producing reliable coatings. Evaluation tests are now under way on two specimens of 21/4-in.-dia molybdenum rod. These rods were rough-threaded to provide a good bonding surface. One sample was then cleaned by bright-annealing at 1800°F, and the other was etched with chromic-sulfuric pickling solution. Both specimens were sprayed with about 20 mils of molybdenum and then heat-treated at 2250°F for 2 hr in order to bond the coatings. WELDING AND BRAZING STUDIES R. L. Heestand G. M. Slaughter Weldability Evaluations P. Patriarca E. A. Franco-Ferreira Results of tests of specimens of Haynes heats SP-16 and -19 that were machined from V-in. plate and were sent to Rensselaer Polytechnic Institute for hot ductility testing are presented in Table 2.1.7. Since weldability tests of these two alloys had indicated severe weld-metal and base-metal cracking of the SP-16 material and good weldability of the SP-19 material, the results of these tests were expected to be of value in determining whether welding techniques and materials or the base material composition was the cause of the difference in weldability. = As may be seen, the ductility of the SP-19 alloy upon heating to 2200°F was significantly higher than that of the SP-16 alloy. Further, the ductility of the SP-19 alloy was not appreciably impaired by prior heating to 2300°F, but it was impaired somewhat by prior heating to 2350°F and above. Additional weld test plates were made for mechanical property studies of welded joints and for radiographic, metallographic, and hardness examinations, as well as to obtain general information pertaining to the welding characteristics of the materials under conditions of high restraint. In general, the weldability of the INOR-8 alloys appears to be good. No weld cracking difficulties have been experienced with any of various heats, except Haynes heat SP-16. The welds have been found to be sound, in general, and the welding deposition characteristics of the material are comparable with those of Inconel or the stainless 69 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED Y-26035 Fig. 2.1.17. Temperature to 1350°F. steels. A new International Nickel Company filler metal, designated INCO Weld A, which was designed for the joining of dissimilar materials, is being tested for applicability to the joining of Inconel to INOR-8. This new filler metal is reported to be subject to aging at high temperatures {contains 3% titanium), and therefore its mechanical proper- ties both in the as-welded and aged conditions should be studied. All-weld-metal reduced-section tensile bars (0.252-in. dia) were prepared from the test plates of Westinghouse heat M-5, Haynes heat SP-19, and Inconel, and tensile tests were conducted at room temperature, 1200, 1300, and 1500°F. The results are presented in Table 2.1.8. The Inconel data were obtained in order to provide a comparison with the data for INOR-8. Comparisons of wrought sheet and weld metal of both Inconel and material from heat SP-19 are shown in Fig. 2.1.18. Although moderate ductilities were obtained at the lower temperatures for both heats of INOR-.8 70 INOR-8 Pipe Spray-Coated with Molybdenum and Heated Through 24 Thermal Cycles from Room weld metal, the ductilities of both heats were 10% or lower at 1500°F. Although the ultimate tensile strengths are approximately the same, the yield strengths of the two heats of weld metal were much greater than those of the wrought sheet. For example, the yield strength of heat SP-19 wrought sheet at room temperature was approxi- mately 44,000 psi, while it was approximately 30,000 psi at 1200°F. The ductilities of INOR-8 in both forms are comparable. An examination of the data also reveals that, in general, INOR-8 possesses a slightly higher ultimate tensile strength than that of Inconel, a significantly higher yield strength, and a markedly lower high-temper- ature ductility. Remote Welding E. A. Franco-Ferreira A test specimen was supplied to Welding Processes, Inc., Wilmington, Delaware, for semi- remote welding by a machine being developed for Tflble 2.].7- PERIOD ENDING JUNE 30, 1958 Results of High-Temperature Ductility Tests on Hoynes SP-16 and SP-19 Materials Reduction in Area Ultimate Tensile Strength Thermal Cycle (%) (psi) SP-16 SP-19 SP-16 SP19 1800°F (on heating) 40.8 63.4 53,600 56,400 2000°F (on heating) 30.8 73.8 41,300 42,800 2200°F (on heating) 1.8 72.8 3,400 30,200 2300°F (on heating) 1.1 6.0 20,000 2400°F (on heating) 0.5 0.0 1,300 1,800 2400°F cooled to 2300°F 046 0.0 600 5,600 2400°F cooled to 2200°F 0.1 1,700 2400°F cooled to 2000°F 1.8 14,000 2400°F cooled to 1800°F 4,5 17,300 2350°F cooled to 1800°F 33.2 55,200 2350°F cooled to 2000°F 30.8 44,400 2350°F cooled to 2200°F 0.3 24,900 2300°F cooled to 1800°F 5%.1 59,000 2300°F cooled to 2000°F 68.4 45,900 2300°F cooled to 2200°F 63.8 31,800 the Westinghouse PAR Project. The specimen was a 10-in.-dia Inconel pipe with a 5/B-in. wall. The weld was made with the pipe axis vertical; nitrogen was used as the backup gas; ond the arc was shielded with helium. A continuous arc time of 5 hr was required for the 59 passes of the weld. Yiews of the face and root of the sample weld are shown in Figs. 2.1.19 and 2.1.20. The large amount of root push-through was apparently the result of internal mismatch in the pipe samples rather than to an inherent defect in the welding procedure, Radiographic inspection at ORNL showed the weld to be completely sound, with no evidence of porosity. A cross section of the weld may be seen in Fig. 2.1.21. The structure is that of a normal Inconel weldment. Joint Development G. M. Slaughter property tests were made on cast silver in order to provide information pertinent to the design of a flanged joint with a cast metal seal Mechanical (see Chap. 1.2, this report). The 0.252-in.-dia tensile bars prepared from cast silver were tested at room temperature, 1200°F, and 1400°F. The test results indicated that the room-temperature mechanical properties were about the same as those given in published data, that is, 20,000-psi tensile strength, 8,000-psi yield strength, and 50% elongation in 1 in. At 1200°F, the tensile strength dropped to 2400 psi and the yield strength to 1900 psi, with a corresponding elongation of 6 to 10%. At 1400°F, the tensile strength was 1500 psi, the yield strength was 1000 psi, and the elongation was 10%. Half-sections of two flanged joints with cast- metal seals that were tested under simulated operating conditions, as described in Chap. 1.2, are shown in Figs. 2.1.22 and 2.1.23. The sedl material used for the joint shown in Fig. 2.1.22 was silver, whereas the seal material used for the joint shown in Fig. 2.1.23 was a silver-copper alloy. The examination indicated that moderate oxidation of the components had occurred during 71 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT opening and closing of the flange and that the oxidation had impeded wetting of the base metal. The joint with the silver-copper alloy seal appeared to be less subject to nonwetting than the joint containing the pure silver seal, probably because of the lower temperatures required to remelt the alloy during opening. A typical interface between the silver-copper alloy and the Inconel base metal is shown in Fig. 2.1.24. Table 2.1.8, Comparative Tensile Properties of All-Weld-Meta! Specimens of INOR-8 and Inconel (As Welded) INOR-8 Temperature Westinghouse Haynes Heat Inconel Heat M-5 SP-19 Ultimate Tensile Strength (psi) Room 116,615 116,725 95,295 121,260 115,095 93,975 1200° F 72,350 74,965 67,570 75,575 65,930 68,945 1300°F 62,855 59,075 58,740 67,500 64,080 59,015 1500°F 54,595 53,780 37,480 57,040 52,175 36,505 Yield Strength at 0,2% Offset {psi) Room 75,385 78,425 56,790 80,840 76,390 55,980 1200°F 55,580 54,595 37,790 57,240 50,525 44,100 1300°F 52,900 38,610 50,300 37,390 1500°F 50,520 44,815 34,630 51,130 44,930 34,285 Elongation (% in l-in. gage) Room 41 37 411 52 38 45 1200°F 18 19 49 18 17 41 1300°F 12 16 38 12 40 1500°F 4.5 9.5 77 5.5 10 65 72 Component Fabrication G. M. Slaughter The Griscom-Russell Company, Massillon, Ohio, is developing an internal tube welding procedure that would be applicable to the attachment of tubes to tube sheets in heat exchangers if back-brazing were impractical, as it would be for thick tube sheets or exceptionally large heat exchangers. In order to make tube-to-tube-sheet joints with the procedure being developed, the tube sheet is drilled with holes of the same dimension as the inside diameter of the tubing, and bosses are machined or otherwise formed on the underside of the tube sheet to permit a butt welding operation. Inert-gas-shielded tungsten-arc welding is then performed internally with a specially constructed rotating-electrode mechanism designed to provide a full penetration weld. A test joint is shown before and after welding in Fig. 2.1.25. The joints made by this GriscomRussell procedure have included 1]/2-in.-OD, 0.072-in.-wall to ]/2~in.-0D, 0.070-in.-wall tubing, but most of the experience has been with the larger size tubes. The primary production item has been a heat exchanger made for the Knolls Atomic Power lLaboratory of 2.25% Cr—1% Mo steel, which con- tained 76 welds. The tubes were 11/2-in. OD and 15/] -in. OD, and both sizes of tubing had 0.072-in. wafis. A power supply which consists of a 200-amp Vickers Controlarc rectifier with a program timer was used for this welding. The weld program high current, rapid current decay to the normal welding current, and slow decay at the weld termination. The tungsten electrode was seated in a modified copper collet assembly, and a stainless steel gas-focusing assembly was used to direct the shielding gas onto the weld zone. Test welds made with 1-in.-0D, 0.070-in.-wall, type 316 stainless steel tubes were obtained for examination at ORNL. Incomplete fusion of the locating lips of the headers was found in some areas of some of the tubes, but this could probably be remedied by careful determination of the optimum welding variables. Moderate oxidation of the outer surfaces of the welds was found, but this condition was probably the result of inadequate purging by the makeshift welding fixture being used. Nondestructive and metallographic ex- aminations of these welds have not been completed, and it is not yet known whether they contain included an initial e R B e PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED ORNL—LR—-DWG 31246 —B— [NOR-8 WELD METAL (ULTIMATE TENSILE STRENGTH) ~=B== [NOR-8 WELD METAL (YIELD STRENGTH) —(— [NCONE{. WELD METAL {ULTIMATE TENSILE STRENGTH) === [NCONEL WELD METAL (YIELD STRENGTH) e @e== |[NOR-8 SHEET (ULTIMATE TENSILE STRENGTH) ~=@== INOR-8 SHEET (YIELD STRENGTH) —0— INCONEL SHEET (ULTIMATE TENSILE STRENGTH)} FROM =-0-=- INCONEL SHEET (YiELD STRENGTH) INT. DUCKEL (X 103) 160 120 R - .—‘— .’% - e C 80 i e L o }._ o) 40 | O 80 I —ge— NOCR-8 WELD METAL ) —pe (NCONEL WELD METAL —@= [NOR—8 SHEET ———O—— INCONEL SHEET ;6 o L = ——— g 40 | \ . / 1 y & k A— - — T e =T SR \oven. - WELD ALL AROUND SILVER SOLDER/\‘/.s—in.DFnLL, 3 HOLES 120 ° APART FOR PINS SILVER SOLDER ASSEMBLY Fig. 2.2,6. Capsule for Testing the Stability of Graphite in Contact with MoltensSalt Fuel Under lrradiation in the MTR. 82 PERIOD ENDING JUNE 30, 1958 2.3. CHEMISTRY W. R. Grimes Chemistry Division PHASE EQUILIBRIUM STUDIES Systems Containing UF, and/or ThF, R. E. Thoma H. A. Friedman H. Insley C. F. Weaver The System LiF-Ber-ThF4-UF4. — Detailed studies are being made of the phase equilibria characteristics of the quaternary system LiF- BeF ,-ThF ,-UF, in the composition region 26 to 40 mole % BeFZ, 1 to 3 mole % UF,, and 1 mole % ThF ,. A fused-salt breeder-reactor fuel having a liquidus temperature of 440 + 5°C and containing no more than 36 mole % BeF, can be chosen from this composition region. Liquidus curves for some LiF-BeF,-ThF ,-UF , compositions in this region are shown in Fig. 2.3.1. The System LiF-BeF,-ThF,. — A partial phase diagram of the system LiF-BeF,-ThF, was pre- senfed previously,! and a more complete phase diagram based on inferences from the best thermal analysis and quenching data presently available is shown in Fig. 2.3.2. Recent results of optical and x-ray diffraction examinations of quenched R, E. Thema et al., MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 81, UNCLASSIFIED ORNL-LR -DWG 29668R x 1 mole % UF, +1mole % ThFy g0 } N\ L ® 2mole % UF, + 1mole % ThF, # . o> 3mole % UF, +1mole % Thi, | | 455 ‘o — e - —_— S | : 2LiF-BeF LIQUID | “ 2. + LiQUID . J L b o o o e x 450 | iU id i = S o] e — [ | | w445 re— T TTLIF B, TR T T T 5 | +L1Quip \ [ % ! - o LiQuio. b 440 - 2LiF BeF ‘ | + LIQUID . | ‘ 4 3 5 },7 — — 77 ——— e—— — _ I - _— ,,7{ P _I‘_ __[ ol 1] 30 3 32 33 24 35 36 37 38 39 BeF, (mole 7o) Fig., 2.3.1. Liquidus Temperatures in the System LiF'BEFz'UFA“Th F4. samples indicate that the following three eutectics occur within this fuel solvent system: Composition (Mole %) Melting Point (°C) LiF BeF2 ThF4 66 29 5 427 65 30 5 429 47 51.5 1.5 ~360 Two experimental factors preclude the immediate construction of an unequivocal polythermal phase diagram. First, LiF-ThF, solid phases do not appear in the ternary system LiF-BeF ,-ThF, as pure phases; that is, they contain BeF,. The type of solution which occurs is not, as yet, known. Since no LiF-ThF, compounds occur as pure phases in the ternary system, all Alkemade lines become areas in the polythermal diagram rather than lines and thus lose some significance relative to defining compatibility triangles. Second, no hydrolysis products have been ob- served in any LiF-BeF.,-ThF, samples, although these materials convert UF , into UD, so readily in LiF-BeF ,-ThF ,-UF, mixtures that stringent measures are required to ensure their absence. UNCLASSIFIED CORNL-LR-DWG 29666 7 LiF-SThF‘r\ L-LiF-2Thf, \. ‘ 3 LiF-ThFy,- s 2LiF-BeF,” LiF-Befp~ Fig. 2.3.2. The System LiF-BeF,-ThF,. 83 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT Previous reports '+ of results of LiF- -BeF ,-ThF quenches showed consistent liquidus and secondary phase temperatures and rarely reproducible solidus temperatures. |t can be inferred that such vari- ations in solidus values are indicative of the presence of a quaternary liquid containing hy- drolysis products in solution. The most recent quench data for mixtures with compositions close to that of the 2LiF.BeF,-7LiF.6ThF 4-LiF-2ThF eutectic and to that of 'rhe 2LiF.BeF -7L|F 6ThF ,- 3LiF-ThF, eutectic were obtcnned with somples in whlch only prehydrofluorinated BeF, was used. The results from these data are much more self-consistent than those obtained with as- received BeF . The System LiF- ThF,~UF,. — In quaternary mixtures of LiF-BeF -ThF UF4, solid solutions of LiF-ThF, and L|F UF cipitated from molten secondary phases. compounds are pre- |iquids as primary or An understanding of the UF“-ThF4 phase behavior in molten-salt reactor fuels whose compositions are chosen from the quaternary system LiF-BeF ,-ThF ,-UF, is there- fore dependent on a rellobiy accurofe L|F ThF ,- UF , phase diagram, A comprehensive study of the phase equilibria in the system LiF- ThF ,-UF, has been made, and the phase diagram for 'rhe system is pre- sented in Fig. 2.3.3. A report giving complete results of the thermal analysis and of optical and x-ray diffraction examinations of quenched samples is being prepared. The salient characteristic of the system LiF- ThF ,-UF,, as would be expected, is the extensive formation of solid solutions. In the course of the investigation detailed studies were made of methods for the precise determination of UF, and ThF, concentrations in primary phases. |t was found that concentrations precise to within +0.5 mole % could be determined with the petro- graphic microscope for UF , or ThF ,. The supplemental results obfcmed from recently completed thermal-gradient quenching experiments were consistent with the results presented pre- viously,! and liquidus isotherms could be drawn for the entire system. Also, data from the recent experiments defined the phase boundaries of the 2R, E. Thoma, Resulis of X-Ray Diffraction Phase Analyses of Fused Salt Mixtures, ORNL CF-58-2.59 (Feb. 18, 1958). . 84 3LiF-ThF, compound and established the solu- bility limit of UF, in this phase at the solidus. The compound 3LiF-UF , has no region of stability in either the binary system LiF-UF4 or the ternary system LiF-ThF -UF,. It is probably isostructural with 3LiF.ThF ,. Solubility studies of mixtures with the composition 3LiF.UF,- 3LiF-ThF , (75mole % LiF) show that 3LiF . Th(U)F, solid solutions containing as much as 15.5 mole % UF, may be formed. This solid solution is green, is biaxial negative, and has a birefringence of approximately 0,008, with an optic angle of approximately 30 deg. The indices of refraction of 3LiF-Th(U)F4 as a function of UF, are shown in Fig. 2.3.4. content Solubility of PuF, in Alkali Fluoride~Beryllium Fluoride Mixtures C. J. Barton R. A. Strehlow Further information was obtained on the solu- bility of PuF in alkali fluoride—beryllium fluoride mixtures, During the quarter, additional data were obtained for several of the solvent com- positions previously tested and for one new solvent mixture. Some of the previously reported solubility values were changed slightly as the result of repeated analyses, and, in addition, the solvent compositions were more accurately defined. All the data obtained to date are given in Table 2.3.1 and are shown graphically in Figs. 2.3.5 and 2.3.6. The values marked with asterisks in Table 2.3.1 are believed to be in- correct, and therefore they were omitted from the graphs. The data® indicate that the solu- bility of PuF, in LiF-BeF, mixtures is at a minimum for mixtures containing about 63 mole % LiF and that it is at @ minimum in the NaF-BeF, system for mixtures containing about 57 mole % NaF. Comparison of PuF, solubility data with the results of CeF, solubility determinations? showed that CeF, was slightly more soluble than PuF, for the same solvent composition and temperature. A study of the effect of CeF, on the solubility of PUF3 in LiF-BeI:2 (63-37 mole %) is under way. 3c. 4. Barten, W. R. Grimes, and R. A, Strehlow, Solubility and Stability of PuF3 in Fused Alkali Flu- oride—Bery!lium Fluoride Mixtures, ORNL<2530 (June 11, 1958). “W. T. Ward, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 94, and subsequent section of this chapter. PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED ORNL~-LR-DWG 28215R ThF, 114 / / LiF-4ThF TEMPERATURES IN °C 4 7 0 % 77 0 % LiF-2ThF, A 9 & ke P890 g /N s, o, o /N ‘ N/ 7LiF-BThF, , ROS \\ L ? R\ vo & D P595{ 65, % % E560 Ad‘ O 3LIF-ThF, B 9, 605 % / E5704 o ~ ‘ 6 C 9 5 & S 3 % / /N D 8D e A % @ % N 0 0 3% R A\E \uos 6. & % O Q \ % % /N 7 | , LiF 500~ “490 610 770 LiF- 4 UF UF, 845 4LiF-UR, 7LiF-6UF, 4 1035 Fig. 2.3.3. The System LiF-ThF,-UF,. Because there is interest in the possibility BeF, (65.4-34.6 mole %), obtained from Table of converting Th232 to fissionable U2?3® in a 2.3.1, are compared with data for PuF, solubility plutonium-fueled fused-salt reactor, a brief study in LiF-BeF-ThF, (62.4-36.6-1.0 mole %, cal- was made of the effect of 1 mole % ThF, on the culated composition). A plot of solvent com- solubility of PuF, in LiF-BeF, (63-37 mole %). position vs PuF, solubility3 indicated that the The results of the study are presented in Fig. difference in the LiF-to-BeF, ratios of the two 2.3.7, in which data for PuF, solubility in LiF- solvents would have very little effect on the 85 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT INDEX OF REFRACTION 1.540 1.530 1.520 1.510 1.500 1.480 1.480 Fig. UNCLASSIFIED ORNL-LR-DWG 31247 — MAXIMUM EXTENT OF — 3LiF - ThF, SOLID SOLUTION ¢ O o0— UF, CONTENT (mole %!} 2,3.4, Solid Solution., PuFy (mole % ) o o 0.2 oA 10 Fig. UNCLASSIFIED ORNL—LR—DWG 29605 TEMPERATURE (°C) 600 550 650 500 74.3 LiF-28.78eF, | 51,7 LiF ~ 48.3 BeF, 65.4 LiF~34.6 BeF, 56.3LiF-43.7Bef, 5 11.0 1.5 12.0 12.5 13.0 13.5 14.0 10,000 /7 (°K) 2.3.5. Solubility of PuF; as a Function of Temperature for LiF-BeF2 Solvents, 86 Indices of Refroction for 3LiF-ThF4 PuF, (mole %} UNCLASSIFIED ORNL—LR—DWG 29606 650 550 500 TEMPERATURE (°C) 600 _l_ T ] i{ 565NGF 1?5LnF 26BefF, | 1 | | | | o Gn L 57.0 NaF-43.0Bef;~ 5 0.2 L ~ ~~\4 | \ 49?NGF 50.3BeF;, \ o4 [ 10,0 10.5 11.0 11.5 12.0 12.5 13.0 13.5 10%/°K Fig. 2.3.6. Solubility of PuF; as a Function of Temperature for NoF-BeF, and NoF-LiF-BeF, Solvents. UNCLASSIFIED ORNL-LR-DWG 31220 TEMPERATURE(°C) 650 600 550 500 1.0 0.8 06 3 @ g 04 > = — @ o | ® o2 u”? o SOLUBILITY IN LiF-BeF, a (65.4 - 366m0|e%) ® SOLUBILITYIN LiF- BeF UF (62.4-36.6-1mole%) 04 0.08 10.0 10.5 1.0 1o 2.0 12.5 130 135 10%/°K Fig. 2.3.7. Effect of Addition of ThF, on Solubility of PuF3 in an LiF-BeF2 Mixture, PERIOD ENDING JUNE 30, 1958 Table 2.3.1. Solubility of PuF, in Alkali Fluoride~Beryllium Fluoride Mixtures Solvent Composition (mole %) Fi|trcfionoTemperdfure Filtrate Analysis NoF LiF BeF, v Pu lwt %) PuF3 (mole %) 49.5 50.5 552 1.18 0.22 600 1.73 0.33 600 1.79 0.34 651 2,72 0.52 57.0 43.0 538 1.17* 0.22* 552 1.63* 0.31* 600 1.35 0.26 600 1.26 0.24 609 1.26 0.24 650 1.48* 0.28* 652 2.21 0.42 706 3.40 0.66 63.4 36.6 550 1.54 0.29 598 2.43 0.46 600 2,00* 0.38* 650 4.40 0.85 51.7 48.3 463 1.02 0.16 549 2,44 0.38 599 2.96* 0.47 654 5.76 0.93 56.3 43.7 494 1.04 0.15 560 1.89 0.28 602 3.15 0.48 653 5.47 0.86 550 1.98 0.30 599 2.04* 0.31* 649 6.24* 0.95* 65.4 34.6 532 1.15 0.16 600 1.78 0.27 643 4.30 0.63 713 28.7 546 4,00 0.56 597 5.90 0.85 650 8.48 1.26 56.5 17.5 26.0 500 2.92 0.51 554 7.68 1.43 565 2.61* 0.46* 600 7.59 1.41 634 13.0 2.58 655 6.45* 1.18* *Daubtful results excluded from Figs. 2.3.5 and 2.3.6. 87 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT solubility of PuF,, but it is expected that the solubility of PuF, in pure LiF-BeF, (63-37 mole %) will be determined in the near future in order to obtain @ more direct comparison. Al- though the scatter of the data shown in Fig. 2.3.7 precludes firm conclusions, it seems likely that the addition of 1 mole % ThF, to an LiF- BeF, (63-37 mole %) mixture will not appreciably affect the solubility of PuF, in the solvent. FISSION-PRODUCT BEHAVIOR G. M. Watson F. F. Blankenship Solubility of Noble Gases in Molten Fluoride Mixtures N. V. Smith Studies were made of the solubilities of argon in NaF-LiF-KF (11.5-42-46.5 mole %) and of helium in LiF-BeF, (64-36 mole %), and numerical values expressed as Henry's law constants are presented in Tables 2.3.2 and 2.3.3 and Fig. 2.3.8. For comparison, the previously reported® solubility constants of helium and neon are also shown in Fig. 2.3.8. SN. V. Smith, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL.-2474, p 91. The trends of the data are the same as those previously observed for the solubilities of the noble gases in mixtures containing ZrF,. The solubilities follow Henry's law, increase with increasing temperature, and decrease with in- creasing molecular weight of the gas. However, an interesting difference in solubility is ex- hibited by helium in the LiF-BeF, mixture. The heat of solution of helium is considerably lower in LiF-BeF, than:in NaF-KF-LiF or in mixtures containing ZrF, ond is not approximated by the product TAS calculated from the gas law for isothermal expansion of an ideal gas from an initial concentration, C , at the temperature and pressure of saturation, to a final concentration, C o that is numerically equal to the measured solubility. The entropy of solution diminished by the entropy of expansion of an ideal gas over the same concentration range, Cg to C is considerably larger in the mixture containing BeF, than in any of the molten fluoride mixtures previously studied. The heats and entropies of solution for helium, neon, and argon in NaF- KF-LiF and for helium in LiF-BeF,, as well ®). H. Shaffer et al., MSR Quar. Prog. Rep. Oct. 31, 1957, ORNL-2431, p 41. Table 2.3.2. Solubility of Argon in NaF-KF-LiF (11.5-42-46.5 Mole %) Temperature Saturating Pressure Solubility . (°C) {atm) (moles of urgon/cm3 of melt) K x 108 x 10~8 600 1.01 0.834 0.825 1.50 1.378 0.917 2,02 1,929 0.957 Av 0.90 0,05 700 1.00 1.744 1.744 1.50 2,705 1.799 2,04 3.811 71.87] Av 1.80 £0.04 800 1.004 3.4 3.397 1.51 5.161 3.422 2.033 6.842 3.365 Av 3.40 £0.02 *K = ¢/p in moles of gas per cubic centimeter of melt per atmosphere. 88 PERIOD ENDING JUNE 30, 1958 Table 2,3.3. Solubility of Helium in LiF-BeF2 (64-36 Mole %) Temperature Saturating Pressure Solubility . °C) (atm) (moles of he lium/em? of melt) K x 10~8 x 108 500 1.009 7.46 7.39 1.521 11.46 7.53 1.996 15.08 7.56 Av 7.49 600 1.103 11.74 11.59 1.570 18.76 11.95 2,107 22,40 11.1 Av 11,55 700 1.004 14,81 14,75 1.526 23,75 15.56 1.963 28,41 14,47 Av 14.93 800 0.996 19.36 19.47 1.52 29,63 19.49 1.978 38.55 19.49 AV 19149 *K = c/p in moles of gas per cubic centimeter of melt per atmosphere. as the corresponding TAS products, are given in Table 2.3.4. As may be observed, the heats of solution in NaF-KF-LiF can be estimated to within 10% by considering the gas to be ideal and calculating the entropy of isothermal ex- pansion over the concentration interval defined by a single experiment. This is not the case, however, for helium in LiF-Ber. The fundamental significance of these observations is not imme- diately apparent, and efforts to interpret the data will be deferred until additional measurements have been made. Solubility of HF in LiF-BeF, Mixtures J. H. Shaffer The investigation of the solubility of HF in LiF-BeF, composition mixtures as a function of solvent in the range 0-55 mole % BeF, was almost completed, and, in order to avoid repetition, it is planned to withhold presentation of the results until the measurements are com- pleted. It may be mentioned that, in mixtures with low alkali fluoride content, the solubility of HF in LiF-BeF, mixtures is approximately the same as in NaF-ZrF, mixtures. As the alkali fluoride content is increased, however, the solu- bilities of HF in the two mixtures differ markedly. For example, in mixtures containing about 90 mole % alkali fluoride, the HF solubilities are about tenfold lower in the LiF-BeF, system than in the NaF-ZrF, system. This comparison is based on the inferpolation described previously” of available data for the NaF-ZrF, system. 7). H. Shaffer, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 93. 89 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT UNCLASSIFIED ORNL-LR-DWG 31221 TEMPERATURE (°C) (x1079) 800 700 600 500 1000 HELIUM IN NaF-KF-LiF 100 HELIUM IN iF-BeF, NEON IN aF-KF-LiF K [moles OF GAS/(cm® OF melt) (atm )] IN NaF-KF-LiF 104/ °K Fig. 2.3.8. Solubilities of Noble Gases in NaF-KF- LiF (11,5-42-46,5 Mole %) and in LiF-BeF2 (64-36 Mole %). Solubilities of Fission-Product Fluorides in Molten Alkali Fluoride—~Beryllium Fluoride Solvents W. T. Ward Determinations of the solubility of CeF, over the temperature range of 450 to 700°C were com- pleted for both LiF-BeF2 and NaF-BeF, solvents ranging in composition between 50 and 70 mole % alkali fluoride. It was found that the solubility of CeF, passed through a minimum at approxi- mately 62 to 63 mole % alkali fluoride in both solvents. The solubility values were somewhat in NaF-BeF, than in LiF-BeF, of cor- responding composition, as may be seen in Fig. 2.3.9, in which the solubilities in terms of weight per cent cerium are plotted as functions of solvent composition for three different peratures. The same solubility values in terms of mole per cent CeF, in the filtrates are shown in Fig. 2.3.10. Solubility values taken from these curves are listed in Table 2.3.5. less tem=- Chemical Reactions of Oxides with Fluorides in LiF-KF J. H. Shaffer As shown previously,® it is possible to effect some chemical separations of solutes in molten 83. H. Shaffer, MSR Quar. Prog. Rep. Janm 31, 1958, ORNL-2474, p 99. Table 2.3.4. Enthelpy and Entropy Changes Occurring upon Solution of Some Noble Gases in Molten Fluorides at 1000°K Solution ldeal Gas Sol vent Gas AH (cal/mole) AS (ev) Expansion TAS NaF-KF-LiF Helium 8,000 -0.25 8,250 (11.5-42-46.5 mole %) Neon 8,900 - 0.97 9,870 Argon 12,400 -0.10 12,500 LiF-BeF2 Helium 4,850 ~3.73 8,580 (64-36 mole %) 90 PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED o UNCLASSIFIED ORNL—LR—DWG 3t222 ORNL~LR—DWG 24223 ] [ { l ] | glf T 2.2 8 ! 2.0 7 / 1.8 C / -~ 1.6 _6 o S € 14 E 700°C Z'; 5 ° W E \\/ R 5 / | E Da \ = 10 = z - / 700°C o o / | 8 os o3 }r %Oooc \J/ J 0.6 > . A 600°C _| N 0.4 4 \'X 'X\ Ly 0.2 o ‘x‘*L"""‘*soo"c \&"w.. 500°C 500°C ‘ *Tex=x= 500°C o 0 ’ 50 60 70 80 50 60 70 80 50 60 70 80 50 & 70 80 LiF IN LiF-BeF, SOLVENT NoF IN NaF-BeF, SOLVENT LiF IN LiF-BeF, SOLVENT NaF IN NaF-BeF, SOLVENT (mole %) (mole %] (mole %) A Fig. 2.3.9. Solubility of Ce‘sF3 as Ce (wt %) in Alkali Fig. 2.3.10. Solubility of CeF, as CeF, (Mole %) in Fluoride—Beryllium Fluoride Solvents, Alkali Fluoride—Beryllium Fluoride Solvents. Table 2.3.5. Solubility of Cer* at 500, 600, and 700°C in Nc:F-BeF2 and LiF-BeF, Solvents as a Function of Solvent Composition Alkali At 700°C At 600°C At 500°C ali Fluoride In NaF-BeF2 [n LiF-Bo;-.F2 In NaF-BeF, In LiF-BeF, In NaF-BeF, In LiF-BeF2 in Solvent CeF, Ce CeF, Ce CeF, Ce CeF, Ce CeF, Ce CeF, (mole %) (wt %) (mole %) (wt%) (mole%) (wt%) (mole%) (wt%) (mole®) (wt%) (mole®) (wt%) (mole%) 50 3.6 1.25 5.9 1.65 1.84 0.60 2.9 0.79 0.87 0.29 1.3 0.31 55 2.8 0.91 5.2 1.40 1.33 0.41 2.3 0.61 0.58 0.19 0.86 0.22 60 2.15 0.72 4.7 1.24 0.95 0.30 2.00 0.51 0.38 0.12 0.68 0.168 63 2.13 0.74 4,66 1.18 0.82 0.26 1.97 0.48 0.32 0.093 0.66 0.162 65 2.55 0.85 5.1 1.28 0.88 0.28 2.06 0.50 0.30 0.076 0.70 0.166 70 6.2 1.95 7.7 1.9 2.4 0.77 3.2 0.79 75 1.07 *The experimental error of the measurements is estimated to be +5%. 91 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT LiF-KF by selective precipitation as oxides. Relatively sharp separations of uranium from cerium and of zirconium from cerium were demon- strated by the stepwise addition of solid CaO to solutions containing UF ,-CeF, and ZrF,- CeF, in molten LiF-KF. It was also found, however, that CeF , and BeF , coprecipitated from the LiF-KF solvent containing BeF,-CeF, when Ca0 was added. It appeared to be pertinent, therefore, before attempting investigations of LiF-BeF, solvent to determine the characteristics of BeO as a precipitating agent for uranium, zirconium, and cerium from solutions made up with the simple solute LiF-KF. [t would be desirable, of course, to use BeO as a pre- cipitating agent in mixtures containing BeF,, since it would not introduce any foreign con- stituents to the solution. The results of an experiment in which BeO was used as the precipitating agent for zirconium are presented in Fig. 2.3.11. The concentration of zirconium remaining in the LiF-KF solvent is shown as a function of time. A filtrate was obtained before addition of a 20-fold excess of BeO pellets, and filtrates were obtained at intervals after the addition. As may be seen, the precipitation of the zirconium took place in ]/2 hr or less after the addition of the BeO, but several hours were required to reach the stoichio- metric concentration of beryllium in the solvent. On the basis of this experiment alone, it is not possible to state why the concentration of beryl- lium takes some time to build up to the stoichio- metric value, but it may be surmised that zirconium oxide forms as a layer over the pellets of BeO and that it therefore takes time for the BeF, to diffuse out. It is interesting to note that the the free energies of formation of the pure solids® show a positive free-energy change of +16.6 kcal for the reaction mixtures, 2BeQ + ZrF4 \-_—" ZrO2 + QBeF2 . The partial precipitation of cerium that occurred upon addition of 20-fold excess of BeO to an A. Glassner, A Survey of the Free Energies of Formation of the Fluyorides, Chlorides, and Oxides of the Elements to 2500°F, p 6, 20, ANL-5107 (Aug. 1953). 92 UNCLASSIFIED ORNL~LR-DWG 34224 80 e 70 ¢ —T L~ ; . &0 © | 2 3 5 5 50 : » ! 5 ; o | S | ‘% 40 H i | o ] i f @ | £ | o ! % 0 ZIRCONIUM REMAINING IN SOLUTION E 30 | {DETERMINED RADIOCHEMICALLY) N @® BERYLLIUM REMAINING IN SCLUTION ,<_[ {DETERMINED CHEMICALLY ) - | | 20 ! [ ! i 10 ‘ | \ o SO~ 0 ‘ L o 0 5 10 15 20 25 TIME (hr) Fig. 23,11, Results of Reaction of ZrF, (Hf'8'F ) with BeO in LiF-KF (50-50 Mole %). LiF-KF mixture containing CeF, is indicated in Fig. 2.3.12. This experiment confirmed the previous observation® that the oxides of cerium and beryllium coprecipitate. Since it appeared in this experiment that the reaction was surface controlled, an additional 30-fold excess of BeO was added after heating the solution to 900°C. {t was surprising to find after this drastic treat- ment that the cerium had not completely pre- cipitated and that its concentration had only decreased from 57 to 36 meq/100 g of solution. No satisfactory explanation of this behavior is immediately apparent. UNCLASSIFIED ORNL-LR-DWG 31225 140 O CERIUM REMAINING IN SOLUTION ¢ (DETERMINED RADIOCHEMICALLY) 120 | ® BERYLLIUM REMAINING IN SOLUTION ___| {DETERMINED CHEMICALLY) = 100 2 5 © o 80 o Q ® S g [} E a 60 Z O— O [T i <1 ',_- £ 40 20 o 0 5 0 15 20 25 TIVE (hr) Fig. 2.3.12, Results of Reaction of CeF3 with BeQO in LiF-KF (50-50 Mole %) at 600°C. CHEMISTRY OF THE CORROSION PROCESS G. M, Watson G. J. Nessle Activity Coefficients of NiF, in LiF-BeF, C. M. Blood The activity coefficients of NiF, dissolved in a molten mixture of LiF-BeF, (62-38 mole %) are being determined by using techniques described previously, 10.11 The results of experimental measurements at 600°C are summarized in Table 2.3,6, which gives the experimentally determined partial pressures of HF and H,, the mole fractions of NiF, in the melt, and the equilibrium quotients. 1OC. M. Blood, W. R. Grimes, and G. M. Watson, Activity Coefficients of Ferrous Fluoride and of Nickel Fluoride in Molten Sodium Fluoride—Zirconium Flu- oride Solutions, paper 75, Division of Physical and Inorganic Chemistry, 132nd Meeting of the American Chemical Society, New York, Sept. 8~13, 1957. Ne, M Blood, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 105, PERIOD ENDING JUNE 30, 1958 The equilibrium quotients are also shown graphi- cally in Fig. 2.3.13 as functions of the mole fraction of NiF, in the melt. An examination of the results indicates that, within the experimental precision achieved, the equilibrium quotients are independent of the mole fraction of NiF, over a range from approximately 0.9 x 10=% to 21 x 1074, The activity coeffi- cients of NiF, in this solvent, calculated in the manner previously described, '%1! have values of 2347 and of 515 with respect to the solid and liquid standard states, respectively. Brewer's'? tabulation of thermodynamic properties was used in the calculations, and it is now known that the assumed melting point for NiF, is seriously in error. However, additional efforts to refine the values of the calculated activity coefficients will not be made until it is possible to obtain some experimental values for the free energies of formation of pure crystalline NiF, at high temper- atures (500-600°C) by measuring the equilibrium quotients by the present method at concentrations approaching saturation. An attempt to obtain such values at 500°C is under way. Solubility of Nin in LiF-—-BeF2 (61-39 Mole %) C. M. Blood Measurements were made in order to establish a concentration limit below which NiF, does not precipitate from LiF-BeF, os a complex compound or in the pure state. Experimental procedures were used in this investigation that were similar to those used previously'® to determine the solubility of FeF, in LiF-BeF,. The results are summarized in Table 2.3.7 and are shown graphi- cally in Fig. 2.3.14, The solubility of NiF, was found to be inde- pendent of the amount of NiF, added. Further- more, petrographic examination'? of powdered samples of the filtrates revealed that the satu- rating phase was pure NiF,. Accordingly, the activity coefficients of solid NiF, at saturation in this solvent are given directly by the reciprocal of the solubility expressed as mole fraction (see 12| | Brewer et al., Natl. Nuclear Energy Ser. Div. IV, 19B, 65, 110, 202 (1950). 13R. J. Sheil, MSR Quar. Prog. Rep. Jan. 31, 1958, ORNL-2474, p 107. 4R, A, Strehlow, personal communication. 93 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT Table 2,3.6. Equilibrium Quotients at 600°C of the Reaction Nin(d) + H,(g) -‘—_A Ni(s} + 2HF(g) in LiF-BeF2 (62-38 Mole %) Pressure of HF Pressure of H2 NiF2 in Melt . (atm) (atm) (mole fraction) K x 10~2 x 10~4 x 10™4 0.283 3.89 1.418 1.45 0.292 3.84 1.302 1,71 0.288 3.86 1.273 1.69 0.286 3.87 1,273 1.66 0.244 4.10 0.897 1.62 0.233 4.16 0.897 1.46 0.242 4,11 0.897 1.59 0.265 3.98 0,955 1.85 0.250 4.06 0.926 1.66 0.259 4.02 0.949 1.76 0.319 3,69 1.504 1.83 0.322 3.67 1.504 1.88 0.307 3.76 1.620 1.55 0.323 3.67 1.620 1.76 0.454 2.96 3.963 1.76 0.456 2.95 4,050 1.74 0.448 2.99 4,165 1.61 0.457 2.94 4,107 1.73 0.737 1.43 19,15 1.98 0.734 1.44 19.75 1.89 0.749 1.36 20.74 1.99 0.744 1.39 20.54 1.94 AV ].73 io.]z *Kx = PaF/(XNiF2PH2)‘ where X is mole fraction and P is pressure in atmospheres., UNCLASSIFIED (x 10%) ORNL—LR—DWG 31226 4.0 l ‘ l ‘ 35 1 K, =(1.73£0.42) x 10% . - : ! Kols)= T7.37 - ( i‘ x 30 Kyl1)=336 . o ? 7 ¥ (s) = 2347 E y (1} = 545 | 5 | | i ] e ~ e :/ ./ ‘/_/ 2 e T 7 7 7 7 7 . s = .-.‘{ e T | | — I T — o 1.5 | —« o [ [ [ |[ o ‘ . e ! ‘ ‘ ‘ 2 ‘ | ! C 10 i " ‘ i 8 -k R | | : ; | | | o0 | | | r | | 0 ’ | ! . 4 ] 2 4 6 8 10 12 14 16 18 20 22 (x10 ) Nifp IN MELT {mcle fraction} Fig. 2.3,13. Equilibrium Quotients for the Reduction of NiF, by H, in LiF-BeF, (62-38 Mole %). 94 PERIOD ENDING JUNE 30, 1958 Table 2.3.7. Solubility of NiF, in LiF:-BeF'2 (61-39 Mole %) Temperature Ni Added Solubility of NiF, Activity Coefficient of NiF ,(s) (° Q) (wt %) (mole fraction) at Saturation 694 2.85 0.0171 58.5 649 2.14 0.0128 78.1 597 1.33 0.00783 127.7 551 0.90 0.00529 189.0 499 0.58 0.00337 296.7 omNeLAsSIFIED Experimental Determination of Chromium Diffusion TEMPERATURE (°C) Coefficients in Molten Salt—Inconel Systems 5 152 20 600 500 R. B. Evans R. J. Sheil 2 & :g AH=12,000 cal /mole E 102 > = - m > -] c 5 0 o L =z 2 1072 10.0 10.5 14.0 11.5 +12.0 12.5 £3.0 13.5 10%/°K Fig. 2.3.14. Mole %). Solubility of NiF, in LiF-BeF, (61-39 Table 2.3.7). These activity coefficients com- bined with the equilibrium quotients at saturation obtained from reduction experiments will be useful in the experimental determination of free energies of formation of pure crystalline NiF, at high temperatures. The available data on diffusion coefficients for chromium in nickel-base alloys may be divided into two distinct groups. The first group consists of high-temperature data (T > 900°C) which were obtained through classical self-diffusion measure- ments based on labeled-chromium (C¢3') distri- butions within alloy specimens. > No molten salts were involved. The second group consists of data based on the exchange (1) Cro+ Cr*F2 ;B CrF2 + Cro AF°=0and K =1, which tokes place when labeled and unlabeled chromous fluorides are dissolved in a molten salt contained by Inconel. Rates of the depletion of Cr*F, with time and the corresponding Cr® distribution in the Inconel were measured at relatively low temperatures (800 to 600°C),'¢ Diffusion data at these — and even lower — temper- atures are important from the standpoint of esti- mating the to be expected in molten-salt power reactor components. A series of experiments for rechecking the data of the second type, which consist of values long-term corrosion rates obtained at four temperatures, and to extend the 15p, L. Gruzin and G. B. Fedorov, Doklady .Akad. Nauk. S.S.5.R. 105, 264 (1955). V6R. B. Price et al., A Tracer Study of the Transport of Chromium in Fluoride Fuel Systems, BMI-1194 (June 18, 1957). 95 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT data to temperatures below 600°C would be highly desirable. Since solid-state diffusion mechanisms may vary over wide temperature ranges, extrapo- lation of high-temperature data to lower temper- atures could introduce large errors in calculated diffusion rates, The experimental procedures which are to be used in future investigations are described below. Depletion Method. — If consideration is given to an Inconel-molten salt system in which the molten salt initially contains dissolved CrF, and Cr*F, and the Inconel contains no Cr°*, a random exchange will take place, as shown by Egq. i, although the net change of total chromium is zero. The combined action of the exchange reaction and the diffusional forces within the Inconel will result in @ gain of Cr*® in the Inconel and a loss of Cr*F, from the salt, If the depletion of Cr*F, activity in the salt (corrected for time decay) is measured as a function of time, a diffusion coefficient for chromium in the metal may be calculated by means of the following relationship: by=o ~ bt 22 (2) — - l—e erfc(a\/t) , t=0 where t = time, sec, bt=0=counfs/g-min {at time count is made} of a filtered salt sample taken at ¢ = 0, b, = counts/gemin (at time count is made) of a filtered salt sample taken at ¢, A [CF] Py — D, sec"], aq =— —_ v [CrF2] Psalt A/V = ratio of the salt-exposed area of Inconel to the salt volume, em™7, [Cro]/[Cer] = weight fraction ratio of chromium in Inconel to chromous fluoride (as Ce™™) in the salt, pw/psah = density ratic of metal to salt, D = diffusion coefficient, cm?/sec. Equation 2 is based on a simultaneous solution of the linear diffusion equation 2 aCC rO* J CCrO* (3) =D , ot ax2 96 and the equation resulting from a balance of the instantaneous transfer rates of labeled chromium from the salt to the metal, or (4) 0 M ) % DA a[C (0,1)] The variable x is distance within the lnconel measured in the direction of diffusion, in cm; Cc,o. is concentration of Cr°, in g/cm®; and Mc g is the amount of Cr*F, in the melt, in g. r F2 2 The boundary conditions applied to obtain this solution are: (1} the Inconel is infinitely thick in the x direction, (2) the initial concentration of Cr®* in the lnconel is zero, and (3) the concen- tration of Cr®* at the Inconel surface at any ¢t > 0 is governed by Eq. 1 and varies with time according to the relationship [Cr*Fz] (5) (G, oo =Pul e _[E:E;T . The simultaneous solution holds only when a is constant with time. This will require that the temperature and A/V be held constant for any experiment to which Eq. 2 is applied. Inconel containers have been fabricated for a series of diffusion experiments based on these considerations. The containers consist of large cylinders which enclose small cylinders. The cylinders are arranged so that the salt contacts periphery and bottom of the small cylinder and the inner periphery and bottom of the large cylinder. For given-diameter cylinders and given stirrer—thermocouple well arrangements (all components Inconel), the A/V ratio becomes independent of fluid height when the vertical distance between the two cylinder bottoms is properly adjusted. Salt samples may be removed from the container to obtain the count data without changing the value of A/V. The corrected counts are plotted vs the square root of time, and this curve is compared with a plot of the function 1 — e erfe(u) vs u, where u may be considered to be @\/t. Values of a, and then D, are easily calculated by this method. the outer Constant-Potential Method for Low-Temperature Experiments. — The results of experiments based on the depletion method indicate that the Cr*F, content of molten salt will remain essentially constant if the pot is first equilibrated, with L e g e respect to equation 1, at high temperatures for several days and then subjected to lower temper- atures. The same condition would exist if the amounts of CrF, and Cr*F, in the container exceeded the solubility of chromous fluoride at the temperature of interest. In either case, the Cr*F, concentration would remain constant under small subsequent depletions, since large amounts of Cr*F, and/or Cr°* are involved. The surface of a ]/4-in. Inconel tube or thermo- couple well subsequently immersed in such a salt would pick up labeled chromium under conditions of a constant surface potential, that is, the Cr°* concentration at the surface of the immersed specimen would be constant with time. The corresponding Cr°* transfer equation is 1/2 D¢ (6) AM o, = 24C_ o, <—-> w A rearrangement of Eq. 6 gives 2 1 [y [CrF2] 1 D = - ' 167t (z (GO rbp > b = height of the immersed specimen, where r = radius of the immersed specimen, y = total counts of the entire specimen per min at ¢, z = counts of the salt per g-min ot ¢. The measured variable y is an accurate indication of the Cr°* gained by the specimen because of the low penetrations or diffusional distances involved {x < 5 microns). From an experimental point of view, the constant-potential method appears to be the simplest of the two methods, and current plans are to concentrate activity on this method and to utilize depletion experiments only for a few high- temperature studies. The depletion method does not yield measurable depletions at low temper- atures (T < 750°C) unless long exposure times are used. Results. — The results of initial experiments with the two methods are shown on Fig. 2.3.15. The 900°C value was obtained by means of the depletion method and the 755°C value was obtained by using the constant-potential method. PERIOD ENDING JUNE 30, 1958 UNCLASSIFIED ORNL-LR-DWG 31228 - | "ZK‘TI [Cr°] TYPE OF EXPERIMENT "3 T 020 [Cro*]DISTRIBUTION IN ALLOY | (NO SALT), REF.15 -14 e 016 [Cr"F,] DEPLETION METHOO (WITH SALT) o 0.6 Cro*/[Cr*F,] CONSTANT-PONTENTIAL 5 - METHOD (WITHSALT) — | 7 | -16 60 65 7.0 75 80 85 90 95 100 105 10%/°K Fig. 2.3.15. Diffusion Coefficients for Chromium in Nickel-Base Alloys, The solvent was molten NaF-KF-LiF (11.5-42-46.5 mole %) for both cases. High-temperature data which appear in the literature ' for a nickel- chromium alloy similar to Inconel are also shown for comparison. VAPOR PRESSURES FOR THE CsF--BeF2 SYSTEM F. F. Blankenship A study of the vapor pressures of the system CsF-BeF, was made in order to obtain information on the effect of composition on the thermodynamic activities in fuel mixtures containing BeF,. The deviations from ideal behavior in systems related to the BeF,-containing fuel mixtures were found to depend strongly on the size of the alkali cation. Since vapor pressures for the Nc:l':-BeF2 system 17 it was of interest to compare the effect of substituting Cs” ions (radius 1.69 B\) for Na® ions (radius 0.95 K) Another reason for choosing the CsF-BeF, system for S. Cantor were measured previously, study was that association in the vapor phase was expected to be less pronounced, and hence the system should be more readily amenable to determinations of activities from vapor pressures. 7K. A. Sense and R. W. Stone, [. Phys. Chem. 82, 453 (1958). 97 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT The vapor pressure data were obtained by a quasi-static method, developed by Rodebush,'® which gives total pressures. The results, as presented graphically in Fig. 2.3.16, show no change with composition in the composition region from 80 to 100 mole % BeF,. Since CsF probably makes a negligible contribution to the total pressure in this region, such behavior is sug- gestive of a liguid-liquid immiscibility gap. Attempts to confirm the existence of the immis- cibility gap by examination of quenched samples have, as yet, been inconclusive. At 900°C and 30 mole % BeF,, the vapor pressure of NaF-BeF, is 3 mm Hg compared with 7 mm Hg for ideal behavior; for CsF-BeF, the vapor pressure is 1l mm Hg compared with 19 mm Hg for ideal behavior. of the curves for vapor pressure as a function The general shape of composition in both systems probably implies positive deviations in Ber-rich mixtures and negative Also, the existence of a vapor compound such as CsBeF, probably obscures the extent of the negative deviations in the middle of the CsF-BeF, deviations at lower BeF2 contents. system, A transpiration method is being developed for obtaining weight and composition data for the saturated vapor. This information, combined with the total pressure data, will yield partial pressures of the vapor components and, hence, the activities. 8%, H. Rodebush and A. L. Dixon, Phys, Rev. 26, 851 (1925}, UNCLASSIFIED ORNL -LR-0DWG 31229 200 T oo — 00— U e N _ [ T T ’*_A‘i;’,, : 7 T Y P ..Ln_&‘\ 1000°C ——= — i /.jk’ - = e e T 50 : e /'7 \ p— e, 9507C Ny Qo VAPOR PRESSURE {mm Hg} Q CsF CONTENT OF CsF-Bef, MIXTURE (male Fo) Figc 2.3-] 6- Cs F-Ber. Total Vaopor Pressures in the System 98 FUSED CHLORIDES AS HEAT TRANSFER FLUIDS R. E. Moore Tests were run in order to determine the com- patibility of Inconel and fused chlorides of interest as heat ‘-ansfer fluids. In initial tests, the eutectic mixtures 29 mole % KCI-71 mole % ZnCl, and 23 mole % LiCl-77 mole % ZnCl, and Inconel specimens were sealed in containers of fused silica and held for 90 hr at 600°C. After the test the melts were found to contain 330 and 300 ppm of chromium, respectively, and metallo- graphic examinations of the inconel strips re- vealed that there was no perceptible attack. After a subsequent test carried out under similar conditions with excess chromium metal instead of Inconel, the KCI-ZnCl, melt was tound to contain about 9.7 wt % CrCIZ, Another test with an insufficient amount of chromium metal for complete reaction produced a melt containing 6.3 wt % CrCi2 and a metallic phase that was identified by x-ray diffraction as mainly zinc metal containing some chromium metal probably enclosed in pure zinc. It seems clear that the attack on chromium by KCI-ZnCl, is so great that corrosion of lInconel, in which the activity of chromium is about 0.1, should be very serious. The reaction with chromium in the experiments with Inconel was probably far from completion after 90 hr because of the slowness of diffusion of chromium from within the alloy to the surface layers. The mixture 58.3 mole % LiCl-41.7 mole % RbCl, which appears to be more attractive from the standpoint of vapor pressure and corrosion, was also tested in contact with chromium metal in a sealed tube of fused silica. The chloride melt was prepared by evacuation of the prepared mixture at room temperature followed by continued evacuation during slow heating until the sample fused. After heating at 600°C for 72 hr in contact with chromium the melt was still colorless, but there was a thin, dark-green coating on the pieces of chromium metal. The coating was probably Cr,0;, but no x-ray pattern could be obtained. Hydroxide ion resulting from hydrolysis during fusion of the chlorides could account for the formation of Cr203. The reaction 6LiOH + 2Cr°® — Cr203 + 3Li20 + 3H2 is strongly favored from the standpoint of free energy. The melt contained 300 ppm of chromium, that is, far in excess of the amount expected from the reaction Cro+ 2LiCl == 2Li°+ &Cl, . The equilibrium constant calculated from values of free energy of formation'® at 1000°K is 10-29, The chromium content in the melt may represent the solubility of Cr,O4 in LiCI-RbCl. Procedures for purifying LiCI-RbCl to remove hydroxides and oxides are being studied. An- hydrous HCI should be satisfactory, but Laitinen et al.?® report that the reaction is very slow in LiCI-KCl. If this is due to a very low solubility of HCl in the melt, the purification of large batches by this method may be very difficult. One proposed method is fusion with ZrCI4 under helium pressure followed by distillation to remove excess ZrCl, and filtration to remove ZrO,. The reactions ZrCl, + 2H,0 —> Zr0,, + 4HCl and ZrCl, + 2LiOH —> ZrO, + 2LiCl + 2HC should serve to remove all water and hydroxide ion. Experiments are being planned to determine the rate of mass transfer of chromium in Inconel in contact with LiCI-RbCl under a temperature gradient. PERMEABILITY OF GRAPHITE BY MOLTEN FLUORIDE SALTS G. J. Nessle J. E. Eorgan There is evidence that molten salts of the type now being considered for use in molten-salt reactors will permeate graphite. Since advanced designs now being considered include graphite moderators, some method must be devised to 19A, Glassner, A Survey of the Free Energies of Formation of the Fluorides, Chlorides, and Oxides of the Elements to 2500°F, ANL-5107 (Aug. 1953). 20H. A, Laitinen, W. S. Fergusen, and R. A. Oster- young, J. Electrochem. Soc. 104, 516 (1957). PERIOD ENDING JUNE 30, 1958 prevent such impregnation or to reduce it con- siderably, if the necessity of cladding the graphite or protecting it in some manner from contact with the molten-salt fuel is to be avoided. A possible method for preventing the graphite from absorbing the molten salt would be to saturate the graphite with an inert salt whose melting point is somewhat higher than the pro- posed operating temperatures. In order to test this method, samples of graphite were obtained and machined into 3-in, rods 1, ]/2, and ]/4 in. in diameter. A nickel rack was built to hold three rods of each size in a vertical position inside a nickel-lined flanged-topped receiver can. In a separate reactor vessel a 3-kg batch of LiF-MgF, (67.5-32.5 mole %) was purified by successive treatments with hydrogen and HF in the normal manner. The reactor vessel was then cooled to room temperature and connected to the receiver vessel containing the graphite rods by means of a 3/8-in. nickel tubing transfer line. The entire system was evacuated at room temperature for about 1 hr, The receiver vessel containing the graphite rods was then heated to about 1500°F under vacuum and maintained under these conditions for varying lengths of time, After a predetermined time for degassing the graphite, the reactor vessel con- taining the salt was heated to near 1700°F under When the salt had reached the desired it was transferred to the receiver vacuum. temperature vessel containing the graphite rods under vacuum. Upon completion of the salt transfer, with the graphite rods completely submerged, a pressure of 15 psi of helium was applied to the receiver vessel. After a specified length of time the pressure in the receiver vessel was relieved and the salt was transferred out. The graphite rods were then cooled, removed from the receiver and placed in a desiccator. Visual observation indicated that the rods were undamaged. The diameters of the rods showed no changes from the original mi- crometer measurements. vessel, examined visually, The rods were then weighed to determine whether they had gained in weight. No visible fluoride salt was adhering to the surface of any rod. The weight gains of the rods in each of three experiments are listed in Table 2.3.8. The treatment of the graphite in experiment 1 consisted of evacuation to a pressure of 200 g for 24 hr 99 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT Table 2,3.8. Weight Gains of Graphite in Impregnation Tests Experiment Red Rod Diameter Rod Weight (g) Net Weight Gain Average Net Weight No. No. (in.) Before Test After Test g % Gain (%) 1 1 1/4 3.8698 4.0700 0.2002 5.17 2 ‘/4 4.0173 4.2966 0.2793 6.95 5.97 3 ‘/4 3.8709 4.0952 0.2243 5.79 4 ‘/2 15.6341 16.6759 1.0418 6.66 5 ‘/2 15.7552 16.9083 1.1531 7.32 7.12 6 ‘/2 15.6054 16.7557 1.1503 7.37 7 1 62.9228 68.0075 50847 8.08 8 1 64.1091 69.1408 5.0317 7.85 7.87 9 1 63.9181 68.8318 4.9137 7.69 2 1 ‘41 3.9140 4.2294 0.3146 8.04 2 '/4 3.8092 4.0628 0.2536 6.66 7.30 3 ‘41 3.9463 4.2303 0.2840 7.20 4 ‘/2 16.1923 17.2518 1.0595 6.54 5 ‘/2 15.7918 17.0468 1.2550 7.95 7.11 6 ‘/2 16.1700 17.2746 1.1046 6.82 7 1 63.7184 69.6872 5.9688 9.37 8 1 64.0042 69.3868 5.3826 8.41 8.80 9 1 63.8327 69.3364 5.5038 8.62 3 1 1/4 3.8981 4.1530 0.2549 6.54 2 ’/4 3.7425 4.0287 0.2862 7.65 7.1 3 ‘/2 15.8908 17.0235 1.1327 7.13 4 ‘/2 15.8184 17.0048 1.1864 7.50 8.23 > % 15.0504 172852 1.2348 8.20 6 1 63.6333 69.3552 5.7219 8.99 7 1 64.2699 69.2838 5.0139 7.80 8.51 8 1 62.6176 68.0962 5.4786 8.74 100 followed by exposure to the salt mixture for 1 hr, in experiment 2, the graphite was evacuated at a pressure of 400 u for 54 hr and then exposed to the salt mixture for 48 hr. In experiment 3, the graphite was evacuated at a pressure of 100 x for 30 hr and exposed to the salt mixture for 48 hr, One l/z-in.-dic rod and one l-in.-dia rod from each experiment were used to determine the depth of salt penetration by machining several ]/32-in. cuts from each and submitting each cutting for analysis of the lithium and magnesium contents, One ]/4-in.-dia rod from each experiment was completely ground and analyzed for lithium and magnesium content, The analytical results obtained from the ]/4-in.-dia rods and the cuttings from the ]/2- and 1-in.-dia rods are listed in Table 2.3.9. Analysis of the salt bath before the test yielded the following results: Li, 16.5%; Mg, 19.5%; F, 64.7%. According to makeup of the salt composition the constituents should have been present in the following amounts: Li, 14.32%; Mg, 18.12%; F, 67.56%. For comparison with the analytical results obtained for the samples, the analytical values for the salt bath will be used. After the first experiment only six cuttings were made on the ]/2- and l-in.-dia rods, but twelve cuttings were made on the rods used graphite in experiments 2 and 3, and every other cutting was submitted for analysis. The data of Table 2.3.8 show that the repro- ducibility of the fluoride salt penetration of the graphite was fairly good. The weight gains were not as high as might have been expected, but the consistency of the results penetration under these conditions, The results presented in Table 2.3.9 show that the salt penetrated to the center of the rods with little ot no change in salt composition and in a uniform Samples of each size of rod that was in the LiF-MgF, mixture have been mounted on a rack in a receiver vessel and are now soaking in LiF-Bez-FZ-UF:4 (62-37-1 mole %) at 1200°F. The length of this test will be 1000 hr if no complications occur. At the end of the test the rods will be examined to determine the extent of penetration of the beryllium-containing fuel indicates maximum manner. soaked mixture, in particular, the uranium and beryllium compounds. PERIOD ENDING JUNE 30, 1958 PREPARATION OF PURIFIED MATERIALS J. P. Blakely F. F. Blankenship Preparation of CrF, B. J. Sturm The usual procedure for preparing CrF, is to reduce CrF, with H,, but, because of inherent difficulties with gas-solid reactions, it is difficult to completely reduce the CrF, without reducing a small amount of CrF, to Cr° Since there is a continuing need for pure CrF, uncontaminated with Cr°in experimental work related to corrosion problems method of preparation has been sought. The use of SnF, as both a fluorinating and reducing agent appears to be promising. Chromium fluoride has been prepared by heating anhydrous CrCl, and SnF, to 1100°C in a graphite container, in accordance with the reaction in fluoride melts, a more convenient 2CrCl, + 25nF, —> 2CrF, + SnCl, T + SnCI, T . A preliminary trial with stoichiometric proportions of the reactants indicated that some of the SnF, volatilized without reacting. A second attempt with 20% excess SnF, and a starting mixture compacted with a hydraulic press gave an im- proved product. No CrF, could be found by examination with a petrographic microscope; chemical analysis showed 2.5% Cl| and no de- tectable Sn. Production-Scale Operations Alteration of the production facility for use in the large-scale processing of beryllium-containing fluoride salt mixtures is in the final stages. The facility should be ready for use before July 1, 1958. Experimental-Scale Operations C. R. Croft The experimental facilities processed 30 batches of mixed salts totaling some 380 kg during the quarter. Twenty-one of these batches contained beryllium and nine did not. Of the batches con- taining beryllium, 16 were prepared for use in the molten-salt reactor program, three were used by the chemistry section, and two were prepared 101 MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT Table 2.3.9. Penetration Results Obtained from Graphite Impregnation Tests . a Material Balance Graphite Rod Salt Components Found Experiment Rod Diameter Cut (wt %) Lithiom {q) Magnesium (g) No. No. ) No. (in.) Li Mg Foundb Expected® Foundb Expected” 1 ] % 1.52 1.40 0.062 0.028 0.057 0.036 5 % 1 1.46 1.14 2 .48 1.26 3 1.55 1.22 0.267 0.143 0.198 0.209 4 1.51 1.16 5 1.55 1.15 6 1.95 1.10 7 1 1 1.27 1.31 2 1.26 1.17 3 1.38 1.23 4 1.34 1.24 1.020 0.728 0.836 0.921 5 1.31 1.19 6 1.43 1.21 2 1 ‘/4 1.47 1.34 0.062 0.045 0.057 0.057 4 % 1 1.04 0.91 3 1.17 1.03 5 1.38 1.05 7 1.68 0.99 0.254 0.152 0.169 0.192 9 1.59 0.95 " 1.97 0.95 8 1 1 1.27 1.15 3 1.40 1.27 5 1.56 1.06 1.034 0.771 0.805 0.975 7 1.59 1.20 9 1.59 1.14 1 1.50 1.14 3 2 % 1.1 1.23 0.045 0.041 0.050 0.051 4 A 1 1.22 0.90 3 1.41 1.55 5 1.52 1.32 0.218 0.170 0.189 0.214 7 1.55 1.34 9 1.99 1.55 7 1 1 1.03 1.18 3 1.40 1.57 5 1.45 1.47 0.901 0.718 1.005 0.909 7 1.19 1.48 9 .45 1.51 n 1.27 1.47 aAveroge results for the cuttings used. desed on percentage found analytically multiplied by gross weight of rod after experiment. “Based on percentage found analytically for original batch multiplied by net gain in weight of rod after experi- ment. 102 for physical properties study. Seven of the non- beryllium-containing batches were used in the molten-salt program, one is being used in fuel- reprocessing studies, and one is being used by the chemistry section. Transfer and Service Operations F. A. Doss Thirty-nine filling and draining operations were carried out during the quarter. These operations PERIOD ENDING JUNE 30, 1958 involved the transfer of about 95 kg of liquid metals and 200 kg of salt mixtures. Since larger and larger amounts of beryllium- containing salts are being handled by this facility, the area is being enclosed and cut off from general building traffic. It will, however, be directly connected with the newly enclosed production facility. The floors of this facility, together with those in Building 9928, have been sealed with a dust-proofing epoxy-resin compound. 103 WWMNMNNMNONMNPMNRDMNMNDNMNN - ad el e e et = SOV DBNFIOEON SO NGTHEON = 32-34. 35. 36. 37. 38. 39. 40. 41. 42. 43. 44, 45. 46. 47. 48. ) OSOPNOOA LD~ RCZTIMEOOEPPIOMELP E-MPOLICENEIONOETAMZOEOPMMCOME . Alexander Bettis Billington . Blakely . Blankenship . Blizard . Boch Borkowski . Boudreau . Boyd . Bredig Breeding . Briggs . Campbell . Carr Cathers . Center (K-25) . Charpie . Coobs . Culler . DeVYan . Emlet (K-25) . Ergen . Estabrook . Ferguson . Fraas . Franco-Ferreira . Frye, Jr. . Gall . Gresky . Gregg . Grimes . Guth . S. Harrill . W. Hoffman . Hollaender Householder . Jordan Keilholtz . Keim . Kelley . Kertesz . W. Kinyon . E. Lackey A. Lane . S. Livingston H4U=ET® INTERNAL DISTRIBUTION 49. 50. 51. 52. 53. 54. 55. 56. 57. 58. 59. 60. é1. 62. 63. 64. 65. 66. 67. 68. 69. 70. 71. 72. 73. 74. 75. 76. 77. 78. 79. 80. 81. 82. 83. 84. 85. 86. 87. 88. 89-92. 93-104. 105. 106-108. ORNL.-2551 Reactors~Power TID-4500 (13th ed., Rev.) February 15, 1958 . MacPherson . Manly . Mann . Mann . McDonald . McNally Metz . Milford . Miller . Morgan . Murray (Y-12) . Nelson . Osborn . Patriarca M. Perry . Phillips . Reyling . Roberts . Robinson . Savage . Savolainen . Scott . Seagren . Shipley Skinner . Snell . Storto A. Swartout . Taboada . Taylor . Thoma . Trauger . VonderLage . Watson . Weinberg . Whatley . Whitman . Williams . Winters J. Zasler ORNL = Y-12 Technical Library, Document Reference Section Laboratory Records Department Laboratory Records, ORNL R.C. Central Research Library TCOMM=E£~ 4% Ar OINOTLCADP>O00Q MO OEPOMUAMPE-MPEMICPIE-DIP VEE-AMOT-SErM=ET mOOUOmMmITTOmomI 105 EXTERNAL DISTRIBUTION 109. F. C. Moesel, AEC, Washington 110. Division of Research and Development, AEC, ORO 111-687. Given distribution as shown in TID-4500 (13th ed., Rev.) under Reactors—Power category (75 copies ~ OTS) 106