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Title: MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING OCTOBER 31, 1958

Abstract

5 = 6 5 6 L 7 7 6 nterim Design Reactor have resulted in an improved filland-drain system and alternate ways of transferring heat from reactor fuel to steam. A simplified fuel-tosteam heat transfer is being considered that elinfinates intermediate heat transfer fluids. Nuclear calculations relative to a 30-Mw experimental reactor are presented. Calculations of the initial states of U/sup 233/-fueled reactors were completed. Calculations were also made of the fissions in tbe blanket of the Interim Design Reactor. Gamma-ray energy absorption coefficients were computed as a function of energy for 24 elements. Development tests of salt-lubricated pump bearings were continued. A conventional Al bearing and an Inconel journal bearing that had been lubricated with Dowtherm "A" during 3200 hr of operation at 1200 deg F were examined. Materials required for a motor suitable for long term operation at 1250 deg F in a radiation field are being investigated. A small, submerged, centrimgal pump with a frozen-lead seal is being operated in evaluation tests. Freeze-flange and indented-seal-flange joints that had sealed successfully in high-temperature molten-salt lines were tested with Na. Heater sod insulation units designed for use on a 4-inch line were tested along with the large freeze fiange.more » Three commercially available expansion joints were tested and found to be satisfactory for use in moiten salt or Na lines. The enthalpies, heat capacities, and heats of fusion were determlned for the mixtures NaF--BeF/sub 2/- UF/sub 4/, LiF--BeF/sub 2/--UF/sub 4/, NaClCaCl/sub 2/, and LiCl-KCl. Graphite- mederated molten-salt reactors fueled with low-enrichment material were studied, and the results of preliminary calculations indicated that initial enrichments as low as 1.25% U/sup 235/ might be used in a circulatlng-fuel system. A simplified natural-convection moltensalt reactor for operation at 576 Mw (thermal) was studied to determine the approximate size of components and the fuel volume. Materials Studies. Two Inconel and four INOR-8 thermal-convection loops, which circulated various fused-fluoride-salt mixtures, were examined. Test results were also obtained for three Inconel forced-circulation loops operated with fluoride mixtures. INOR-8 samples removed from forced-circulation loop were found to have no evidence of surface attack. A sedium-graphite system was used to prepare Inconel and INOR-8 tensile specimens for a study of the effect of carburization on the mechanical properties. Corrosion tests of pure Ag and Ag- base brazing alloys in static fuel 130 showed Ag and its alloys to have very limited corrosion resistance. Phase equilibrium studies of fluoride-salt systems containing UF/sub 4/ and or ThF/sub 4/ were continued. Extensive gradientquenching experiments were completed on the LiF-fission-preduct ions on the solubility of Puf/sub 3/ in LiF-BeF/sub 2/ mixtures were studied. Solubility studies were made on several systems, including argon in LiF-BeF/sub 2/, HF in LiF--BeF/sub 2/, CeF/sub 3/ in NaF--LiF--BeF/sub 2/ , and CeF/sub 3/, LaF/sub 3/. and SmF/sub 3/ in LiF--BeF/sub 2/ -UF/sub 4/. Two methods of separating U from undesirable fission preducts in fluoride-salt fuel solvents are being studied. Tentative values were obtained for the activities and activity coefficients of Ni metal in Ni--Mo alloys. Apparatus was set up for measuring the surface tensions of BeF/sub 2/ mixtures. Further measurements of the activity coefficients of CrF/ sub 2/ dissolved in the molten mixture NaF-ZrF/sub 4/ were determined and equilibrium quotients were calculated. Studies of Cr diffusion in Cr-containing« less

Publication Date:
Research Org.:
Oak Ridge National Lab., Tenn.
OSTI Identifier:
4274257
Report Number(s):
ORNL-2626
NSA Number:
NSA-13-008291
DOE Contract Number:  
W-7405-ENG-26
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-59
Country of Publication:
United States
Language:
English
Subject:
REACTORS; ABSORPTION; ALUMINUM; ALUMINUM ALLOYS; ALUMINUM CHLORIDES; ARGON; BEARINGS; BERYLLIUM FLUORIDES; BLANKETS; BRAZING; CALCIUM CHLORIDES; CARBIDES; CERIUM FLUORIDES; CHEMICAL REACTIONS; CHROMIUM; CHROMIUM ALLOYS; CHROMIUM FLUORIDES; CONVECTION; COOLANT LOOPS; CORROSION; CORROSION PROTECTION; CRITICALITY; CROSS SECTIONS; DIFFUSION; ELEMENTS; ENERGY; ENERGY RANGE; ENRICHMENT; ENTHALPY; EQUATIONS; EXPANSION; FAILURES; FISSION; FISSION PRODUCTS; FUSED SALT FUEL; GAMMA RADIATION; GRAPHITE; GRAPHITE MODERATOR; HEAT EXCHANGERS; HEAT TRANSFER; HIGH TEMPERATURE; HYDROFLUORIC ACID; INCONEL ALLOYS; INOR-8; IONS; IRRADIATION; JOINTS; LANTHANUM FLUORIDES; LE

Citation Formats

. MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING OCTOBER 31, 1958. United States: N. p., 1959. Web. doi:10.2172/4274257.
. MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING OCTOBER 31, 1958. United States. https://doi.org/10.2172/4274257
. Sat . "MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING OCTOBER 31, 1958". United States. https://doi.org/10.2172/4274257. https://www.osti.gov/servlets/purl/4274257.
@article{osti_4274257,
title = {MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING OCTOBER 31, 1958},
author = {},
abstractNote = {5 = 6 5 6 L 7 7 6 nterim Design Reactor have resulted in an improved filland-drain system and alternate ways of transferring heat from reactor fuel to steam. A simplified fuel-tosteam heat transfer is being considered that elinfinates intermediate heat transfer fluids. Nuclear calculations relative to a 30-Mw experimental reactor are presented. Calculations of the initial states of U/sup 233/-fueled reactors were completed. Calculations were also made of the fissions in tbe blanket of the Interim Design Reactor. Gamma-ray energy absorption coefficients were computed as a function of energy for 24 elements. Development tests of salt-lubricated pump bearings were continued. A conventional Al bearing and an Inconel journal bearing that had been lubricated with Dowtherm "A" during 3200 hr of operation at 1200 deg F were examined. Materials required for a motor suitable for long term operation at 1250 deg F in a radiation field are being investigated. A small, submerged, centrimgal pump with a frozen-lead seal is being operated in evaluation tests. Freeze-flange and indented-seal-flange joints that had sealed successfully in high-temperature molten-salt lines were tested with Na. Heater sod insulation units designed for use on a 4-inch line were tested along with the large freeze fiange. Three commercially available expansion joints were tested and found to be satisfactory for use in moiten salt or Na lines. The enthalpies, heat capacities, and heats of fusion were determlned for the mixtures NaF--BeF/sub 2/- UF/sub 4/, LiF--BeF/sub 2/--UF/sub 4/, NaClCaCl/sub 2/, and LiCl-KCl. Graphite- mederated molten-salt reactors fueled with low-enrichment material were studied, and the results of preliminary calculations indicated that initial enrichments as low as 1.25% U/sup 235/ might be used in a circulatlng-fuel system. A simplified natural-convection moltensalt reactor for operation at 576 Mw (thermal) was studied to determine the approximate size of components and the fuel volume. Materials Studies. Two Inconel and four INOR-8 thermal-convection loops, which circulated various fused-fluoride-salt mixtures, were examined. Test results were also obtained for three Inconel forced-circulation loops operated with fluoride mixtures. INOR-8 samples removed from forced-circulation loop were found to have no evidence of surface attack. A sedium-graphite system was used to prepare Inconel and INOR-8 tensile specimens for a study of the effect of carburization on the mechanical properties. Corrosion tests of pure Ag and Ag- base brazing alloys in static fuel 130 showed Ag and its alloys to have very limited corrosion resistance. Phase equilibrium studies of fluoride-salt systems containing UF/sub 4/ and or ThF/sub 4/ were continued. Extensive gradientquenching experiments were completed on the LiF-fission-preduct ions on the solubility of Puf/sub 3/ in LiF-BeF/sub 2/ mixtures were studied. Solubility studies were made on several systems, including argon in LiF-BeF/sub 2/, HF in LiF--BeF/sub 2/, CeF/sub 3/ in NaF--LiF--BeF/sub 2/ , and CeF/sub 3/, LaF/sub 3/. and SmF/sub 3/ in LiF--BeF/sub 2/ -UF/sub 4/. Two methods of separating U from undesirable fission preducts in fluoride-salt fuel solvents are being studied. Tentative values were obtained for the activities and activity coefficients of Ni metal in Ni--Mo alloys. Apparatus was set up for measuring the surface tensions of BeF/sub 2/ mixtures. Further measurements of the activity coefficients of CrF/ sub 2/ dissolved in the molten mixture NaF-ZrF/sub 4/ were determined and equilibrium quotients were calculated. Studies of Cr diffusion in Cr-containing},
doi = {10.2172/4274257},
url = {https://www.osti.gov/biblio/4274257}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1959},
month = {10}
}