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Title: MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JULY 31, 1960

Abstract

Conceptual designs of the MSRE were made in which the core is constructed of vertical graphite stringers in which fuel passages are machined. A facility for thermally cycling freeze flanges between room temperature and 1400 deg F was designed, and fabrication is nearing completion. Two freeze valves, which have no moving parts, are being fabricated for test; one of these is heated by electrical resistance heaters, the second is heated by a high-frequency induction coil. Operation of forced-circulation corrosion loops continued. Nine INOR-8 loops and two Inconel loops are in operation. Resign and testing work on the MSRE primary and secondary pumps continued. The testing of in-salt bearings was continued. Graphite undergoes shrinkage at MSRE temperatures in a flux of neutrons with energies greater than 0.3 Mev. Calculations were made to determine the effects of this shrinkage in the MSRE core. An analysis was made of temperature effects in a graphitemoderated core with round and flat fuel channels. An analogcomputer analysis was made of loss-of-flow in the MSRE primary system. The last of three INOR-8 corrosion inserts was removed after 15000 hr from an INOR-8 forcedconvection loop. Two typs of solidified metal seals were developed for use with molten fluoridesmore » at elevated temperatures; one contained an alloy sump with a tongue-and-groove joint design, the other has an alloy-impregnated metalfiber compact. A method was devised and equipment was built for leak testing graphite-to-metal braze joints. Welding and back-brazing procedures are being developed for tubeto-tube sheet joints for the MSRE heat exchanger. Tensile tests were performed to determine the effect of low creep strains on the strength and ductility of INOR-8. Permeation tests were made with S-4 and AGOT graphite using LiFBeF/sub 2/--ThF/sub 4/--UF/sub 4/ at 1300 deg F at pressures of 25, 65, and 150 psig in 100-hr exposures. Additional tests were made in order to confirm data indicating that the thermal decomposition of NH/sub 4/F - HF removes oxygen contamination from graphite to such an extent that it could contain molten LiF-BeF/sub 2/--UF/sub 4/ at 1300 deg F without causing the usual UO/sub 2/ precipitation from the fuel. No carburization was detected on unstressed INOR-S specimens after exposure to a LiF-BeF/sub 2/--UF/sub 4/--graphite system for 12000 hr at 1300 deg F. The surface tensions of two NaF--BeF/sub 2/ mixtures were determined to fall between 200 and 150 dynes/cm over the temperature range 500 to 800 deg C. Heat transfer studies with LiF--BeF/sub 2/--UF/sub 4/--ThF/sub 4/ in Inconel and INOR-8 tubes are reported. Preliminary studies indicated that ThF/ sub 4/ in molten-salt reactor fuel may be decontaminated from rareearth fission products by dissolution of the rare-earth fluorides in SbF/sub 5/--HF. (For preceding period see ORNL2973.) (W.L.H.) 167l6 Research and development activities being carried out by du Pont, Nuclear Metals, Inc., Nuclear Development Corporation of America, Inc., and Sargent and Lundy are summarized. Design and construction progress on the Heavy Water Components Test Reactor is reported. The current status of tasks under the ECNG/FWCNG research and development program is outlined. The Carolinas-Viania Nuclear Power Associates work on the Power Demonstration Program is reviewed. (M.C.G.)« less

Authors:
Publication Date:
Research Org.:
Oak Ridge National Lab., Tenn.
OSTI Identifier:
4048751
Report Number(s):
ORNL-3014
NSA Number:
NSA-15-016715
DOE Contract Number:  
W-7405-ENG-26
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-61
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; ALLOYS; ALUMINUM ALLOYS; AMMONIUM COMPOUNDS; ANALOG SYSTEMS; BEARINGS; BERYLLIUM FLUORIDES; BRAZING; CHROMIUM ALLOYS; COMPUTERS; CORROSION; DEFORMATION; DETECTION; DIFFUSION; DUCTILITY; FISSION PRODUCTS; FLUORIDES; FUSED SALT FUEL; GRAPHITE; GRAPHITE MODERATOR; HEAT EXCHANGERS; HEAT TRANSFER; IMPURITIES; IN PILE LOOPS; INCONEL ALLOYS; INOR-8; IRRADIATION; JOINTS; LEAKS; LIQUID FLOW; LITHIUM FLUORIDES; METALS; MOCKUP; MOLYBDENUM ALLOYS; MSRE; NEUTRONS; NICKEL ALLOYS; NIOBIUM ALLOYS; PRESSURE; PUMPS; RARE EARTHS; REACTOR CORE; REACTORS; SEALS; SODIUM FLUORIDES; TEMPERATURE; TENSILE PROPERTIES; TESTING; THORIUM FLUORIDES; TITANIUM ALLOYS; TRANSI

Citation Formats

MacPherson, H. G. MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JULY 31, 1960. United States: N. p., 1960. Web. doi:10.2172/4048751.
MacPherson, H. G. MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JULY 31, 1960. United States. https://doi.org/10.2172/4048751
MacPherson, H. G. Thu . "MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JULY 31, 1960". United States. https://doi.org/10.2172/4048751. https://www.osti.gov/servlets/purl/4048751.
@article{osti_4048751,
title = {MOLTEN-SALT REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING JULY 31, 1960},
author = {MacPherson, H. G.},
abstractNote = {Conceptual designs of the MSRE were made in which the core is constructed of vertical graphite stringers in which fuel passages are machined. A facility for thermally cycling freeze flanges between room temperature and 1400 deg F was designed, and fabrication is nearing completion. Two freeze valves, which have no moving parts, are being fabricated for test; one of these is heated by electrical resistance heaters, the second is heated by a high-frequency induction coil. Operation of forced-circulation corrosion loops continued. Nine INOR-8 loops and two Inconel loops are in operation. Resign and testing work on the MSRE primary and secondary pumps continued. The testing of in-salt bearings was continued. Graphite undergoes shrinkage at MSRE temperatures in a flux of neutrons with energies greater than 0.3 Mev. Calculations were made to determine the effects of this shrinkage in the MSRE core. An analysis was made of temperature effects in a graphitemoderated core with round and flat fuel channels. An analogcomputer analysis was made of loss-of-flow in the MSRE primary system. The last of three INOR-8 corrosion inserts was removed after 15000 hr from an INOR-8 forcedconvection loop. Two typs of solidified metal seals were developed for use with molten fluorides at elevated temperatures; one contained an alloy sump with a tongue-and-groove joint design, the other has an alloy-impregnated metalfiber compact. A method was devised and equipment was built for leak testing graphite-to-metal braze joints. Welding and back-brazing procedures are being developed for tubeto-tube sheet joints for the MSRE heat exchanger. Tensile tests were performed to determine the effect of low creep strains on the strength and ductility of INOR-8. Permeation tests were made with S-4 and AGOT graphite using LiFBeF/sub 2/--ThF/sub 4/--UF/sub 4/ at 1300 deg F at pressures of 25, 65, and 150 psig in 100-hr exposures. Additional tests were made in order to confirm data indicating that the thermal decomposition of NH/sub 4/F - HF removes oxygen contamination from graphite to such an extent that it could contain molten LiF-BeF/sub 2/--UF/sub 4/ at 1300 deg F without causing the usual UO/sub 2/ precipitation from the fuel. No carburization was detected on unstressed INOR-S specimens after exposure to a LiF-BeF/sub 2/--UF/sub 4/--graphite system for 12000 hr at 1300 deg F. The surface tensions of two NaF--BeF/sub 2/ mixtures were determined to fall between 200 and 150 dynes/cm over the temperature range 500 to 800 deg C. Heat transfer studies with LiF--BeF/sub 2/--UF/sub 4/--ThF/sub 4/ in Inconel and INOR-8 tubes are reported. Preliminary studies indicated that ThF/ sub 4/ in molten-salt reactor fuel may be decontaminated from rareearth fission products by dissolution of the rare-earth fluorides in SbF/sub 5/--HF. (For preceding period see ORNL2973.) (W.L.H.) 167l6 Research and development activities being carried out by du Pont, Nuclear Metals, Inc., Nuclear Development Corporation of America, Inc., and Sargent and Lundy are summarized. Design and construction progress on the Heavy Water Components Test Reactor is reported. The current status of tasks under the ECNG/FWCNG research and development program is outlined. The Carolinas-Viania Nuclear Power Associates work on the Power Demonstration Program is reviewed. (M.C.G.)},
doi = {10.2172/4048751},
url = {https://www.osti.gov/biblio/4048751}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1960},
month = {12}
}