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Molten-Salt Reactor Program Semiannual Progress Report for Period Ending July 31, 1964

Description: Report issued by the Oak Ridge National Laboratory discussing semiannual progress made by the Molten-Salt Reactor Program. Descriptions of design, construction, and experimental progress is presented. This report includes tables, illustrations, and photographs.
Date: November 1964
Creator: Briggs, R. B.
Partner: UNT Libraries Government Documents Department

Molten-Salt Reactor Program Semiannual Progress Report, February 28, 1962

Description: Report issued by the Oak Ridge National Laboratory discussing semiannual progress made by the Molten-Salt Reactor Program. Descriptions of design, construction, and experimental progress are presented. This report includes tables, illustrations, and photographs.
Date: 1962
Creator: Oak Ridge National Laboratory
Partner: UNT Libraries Government Documents Department

The Preparation of Anhydrous White Salt in Green Salt Type Box Reactors

Description: Report discussing a method for producing anhydrous TO2F2 in a box-type reactor, which takes approximately seven hours. "The method consists of a heating period of the charge oxide under oxygen, treatment, with anhydrous hydrofluoric acid gas combined with oxygen, and flushing or removal of the excess hydrofluoric acid and formed water with nitrogen gas."
Date: January 29, 1946
Creator: Johnsson, Karl Otto & Clewett, G. H.
Partner: UNT Libraries Government Documents Department

Molten-Salt Reactor Program Semiannual Progress Report for Period Ending February 28, 1965

Description: Report issued by the Oak Ridge National Laboratory discussing semiannual progress made by the Molten-Salt Reactor Program. Progress of operations, construction, engineering analysis, and development is presented. This report includes tables, illustrations, and photographs.
Date: June 1965
Creator: Briggs, R. B.
Partner: UNT Libraries Government Documents Department

Measurements of the Viscosity of FLiNaK : (11.5 Mol % NAF, 42 Mo. % KF, and 46.5 Mol % LiF)

Description: Abstract: "To supplement measurements made with the falling-ball viscometer it was decided to obtain data taken by more rapid, though probably less exact devices. Three essentially different methods were employed to obtain this data. Measurements were made on Flinak which deviated from the average by about +/- 1 centipoise. The data obtained by the Physical Properties Laboratory is in substantial agreement with the measurements made by Francois Kertess and Frank Knox of the Chemistry Section. The results are presented graphically on p. 8. The viscosity varies from about 8 cp at 550 C to about 3 cp at 800 C. Work is progressing to improve the devices and to obtain data on other materials."
Date: 1952
Creator: Tobias, M.
Partner: UNT Libraries Government Documents Department

Removal of uranium from spent salt from the moltensalt oxidation process

Description: Molten salt oxidation (MSO) is a thermal process that has the capability of destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials. In this process, combustible waste and air are introduced into the molten sodium carbonate salt. The organic constituents of the waste materials are oxidized to carbon dioxide and water, while most of the inorganic constituents, including toxic metals, minerals, and radioisotopes, are retained in the molten salt bath. As these impurities accumulate in the salt, the process efficiency drops and the salt must be replaced. An efficient process is needed to separate these toxic metals, minerals, and radioisotopes from the spent carbonate to avoid generating a large volume of secondary waste. Toxic metals such as cadmium, chromium, lead, and zinc etc. are removed by a method described elsewhere. This paper describes a separation strategy developed for radioisotope removal from the mixed spent salt, as well as experimental results, as part of the spent salt cleanup. As the MSO system operates, inorganic products resulting from the reaction of halides, sulfides, phosphates, metals and radionuclides with carbonate accumulate in the salt bath. These must be removed to prevent complete conversion of the sodium carbonate, which would result in eventual losses of destruction efficiency and acid scrubbing capability. There are two operational modes for salt removal: (1) during reactor operation a slip-stream of molten salt is continuously withdrawn with continuous replacement by carbonate, or (2) the spent salt melt is discharged completely and the reactor then refilled with carbonate in batch mode. Because many of the metals and/or radionuclides captured in the salt are hazardous and/or radioactive, spent salt removed from the reactor would create a large secondary waste stream without further treatment. A spent salt clean up/recovery system is necessary to segregate these materials and minimize the amount ...
Date: March 1, 1997
Creator: Summers, L., Hsu, P.C., Holtz, E.V., Hipple, D., Wang, F., Adamson, M.
Partner: UNT Libraries Government Documents Department

Diffusion Welding of Alloys for Molten Salt Service - Status Report

Description: The present work is concerned with heat exchanger development for molten salt service, including the proposed molten salt reactor (MSR), a homogeneous reactor in which the fuel is dissolved in a circulating fluid of molten salt. It is an outgrowth of recent work done under the Next Generation Nuclear Plant (NGNP) program; what the two reactor systems have in common is an inherently safe nuclear plant with a high outlet temperature that is useful for process heat as well as more conventional generation The NGNP program was tasked with investigating the application of a new generation of nuclear power plants to a variety of energy needs. One baseline reactor design for this program is a high temperature, gas-cooled reactor (HTGR), which provides many options for energy use. These might include the conventional Rankine cycle (steam turbine) generation of electricity, but also other methods: for example, Brayton cycle (gas turbine) electrical generation, and the direct use of the high temperatures characteristic of HTGR output for process heat in the chemical industry. Such process heat is currently generated by burning fossil fuels, and is a major contributor to the carbon footprint of the chemical and petrochemical industries. The HTGR, based on graphite fuel elements, can produce very high output temperatures; ideally, temperatures of 900 °C or even greater, which has significant energy advantages. Such temperatures are, of course, at the frontiers of materials limitations, at the upper end of the performance envelope of the metallic materials for which robust construction codes exist, and within the realm of ceramic materials, the fabrication and joining of which, on the scale of large energy systems, are at an earlier stage of development. A considerable amount of work was done in the diffusion welding of materials of interest for HTGR service with alloys such as 617 ...
Date: September 1, 2012
Creator: Clark, Denis; Mizia, Ronald & Sabharwall, Piyush
Partner: UNT Libraries Government Documents Department