MOLTEN-SALT REACTOR PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING APRIL 30, 1959
4 7 7 6 8 2 ng by the ion-exchange mothed at a rate of once per year and at a rate of 12 times per year were compared for the Interim Design Reactor fueled with U/sup233/. The nuclear performance of a one-region, graphite-moderated, unreflected, thoriumconversion molten-salt-fueled reactor was studied to evaluate the effects of processing the fuel at the rate of once per year and 12 times per year, the effects of decreasing the thickness of the core vessel and adding a blanket, and the effects of using a grapbite core vessel and blanket The initial nuclear characteristics of a one-region, heterogeneous, graphite-moderated and -reflected, thorium-conversion, molten-salt-fueled re actor were studied. It was found, that with 13 and 7 mole % ThF/sub 4/ in the fuel, maximum regeneration ratios of 0.846 and 0.798 at inventories of 1150 and 900 kg of U/sup 235/ were obtained. Calculations were also made of the performance of two-region graphite-moderated, molten-salt-fueled, breeder reactors. The effects of fast-neutron reactions in Be/sup 9/ on the reactivity of molten-salt-fueled reactors were studied. Component development and testing including journal bearings. seals, corrosion-testing loops, and MTR in-pile loops are described. The enthalpy, heat capacity, viscosity, and surface tension were experimentally determined for several additionsl fluoride salt mixtures containg BeF/sub 2/ with varyingamounts of UF/sub 4/ and/or ThFs/bu 4/. Corrosion studies were completed on four INOR-8 and four Inconel thermal-convection loops. The effects of penetration of graphite by molten-auoride-salt fuels are being investigated in order to evaluate the problems associated with the use of unclad graphite as a moderator. A series of tests were run to investigate the precipitation of U from fuel 130 (LiF-BeF/sub 2/-UF/sub 4) held in a graphite crucible at 1300 deg F. Several brazing alloys were tested for compatibillty with fuel 130. A specimen of INOR-8 was examined for evidence of carburization that had been exposed to fuel 130 for 4000 hr in a thermalconvection loop for leg at l3O0 deg F. The critical results obtained from rotating-beam tests at 1500 deg F indicate that INOR-8 has significantly better fatigue resistance than Inconel. Based on the results of tensile tests conducted on specimens of INOR-8 aged for 10,000 hr at 1000 to l40O deg F, it was concluded that INOR-8 does not exhibit embrittling tendencies that can be attributed to high-temperature instability. Procedures are being developed for fabricating INOR-8 material ranging in size from thin-walled tubing to heavy plate. A method was developed for brazing graphite to Inconel. Phase diagrams for the systems LiF--BeF/sub 2/-ThFs/bu 4/ and NaF-mat/ locations of primary phases in the system NaF-ThF/sub 4/- UF/sub 4/ determined. The SnF/sub 2/-NH/sub 4/HFs/bu 2/ system was investigated because of its potentialities as a strongly oxidizing, low-melting solvent for reprocessing fuels. Measurements were made of the solubility of PuF/sub 3/ in LiF-BeF/sub 2/- UF/sub 4/ (7O-10-20 mole %). The possibility of separating LiF from the mixture LiF-BeF/sub 2/ (63-37 mole %) by adding NaF and decreasing the temperature was investigated. Tests were initiated for determining the rate of exchange in a proposed method for decreasing the total rare earth content of molten fluoride salts. Two methods for separating U from fission products are being studied. Measurements were made of the solubility of neon in LiF-BeF/sub 2/,and CO/sub 2/ in NaFBeF/sub 2/. Melts from operating INOR-8 and Inconel forcedcirculation loops were analyzed for chromium. Test of the permeability of graphite by molten fluoride salts were continued. A device for testing a proposed fuel sampling and enriching mechanism was constructed and tested. The use of molten ammonium bifluoride as a
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- DOE Contract Number:
- W-7405-ENG-26
- NSA Number:
- NSA-13-016640
- OSTI ID:
- 4270021
- Report Number(s):
- ORNL-2723
- Resource Relation:
- Other Information: Orig. Receipt Date: 31-DEC-59
- Country of Publication:
- United States
- Language:
- English
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MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIODS ENDING JANUARY 31 AND APRIL 30, 1960
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Related Subjects
ALLOYS- BERYLLIUM FLUORIDES- BRAZING- FUELS- FUSED SALT FUEL- LITHIUM FLUORIDES- MEASURED VALUES- POWER PLANTS- REACTORS- URANIUM TETRAFLUORIDE
ALUMINUM ALLOYS- BEAMS- BRITTLENESS- CARBIDES- CHROMIUM ALLOYS- CONVECTION- FABRICATION- FATIGUE- FLUORIDES- FUELS- FUSED SALT FUEL- HIGH TEMPERATURE- INCONEL ALLOYS- INOR-8- MEASURED VALUES- MOLYBDENUM ALLOYS- NICKEL ALLOYS- NIOBIUM ALLOYS- PIPES- PLATES- POWER PLANTS- REACTORS- ROTATION- STABILITY- TENSILE PROPERTIES- TITANIUM ALLOYS- TUBES
ALUMINUM ALLOYS- BERYLLIUM FLUORIDES- BRAZING- CHROMIUM ALLOYS- CONFIGURATION- FUSED SALT FUEL- GRAPHITE- LITHIUM FLU