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Uranium chloride extraction of transuranium elements from LWR fuel

Description: A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800{degrees}C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.
Date: December 31, 1991
Creator: Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R. & Pierce, R.D.
Partner: UNT Libraries Government Documents Department

NEW LABORATORY DEVELOPMENTS IN THE ZIRCEX PROCESS

Description: A new Zircex flowsheet is proposed in which the nonvolatile products from hydrochlorination of uranium-zirconium alloys are chlorinated with carbon tetrachloride, thereby avoiding the loss of 1 to 6% of the uranium observed in engineering development studies of the older flowsheet for STR fuel in which the hydrochlorination residue was dissolved in nitric acid. Other potential advantages of the new flowsheet include decreased corrosion and elimination of possible explosions between uranium--zirconium alloys and nitric acid. The uranium may be recovered by aqueous dissolution and solvent extraction or by gas- phase fluorination at 200 to 400 deg C of uranium chlorides. (auth)
Date: June 1, 1961
Creator: Gens, T.A. & Jolley, R.L.
Partner: UNT Libraries Government Documents Department

The Preparation of Uranium

Description: The method used for the preparation of uranium metal in a fused state was reduction of uranium chloride with calcium in a refractory-lined bomb. The reaction was started by externally heating the bomb with a gas flame. The metal was obtained in a solid chunk which was covered with a layer of fused calcium chloride. The metal obtained by this process had a density of 17.6 which on remelting in a vacuum induction furnace rose to 18.8. The melting temperature of the metal was estimated to be no greater than 1400 C. The metal was malleable, and had a silvery surface when freshly cut which rapidly tarnished, becoming black in the course of a few days.
Date: August 26, 1948
Creator: Rodden, Clement J.
Partner: UNT Libraries Government Documents Department

ZIRCEX FLUORIDE VOLATILITY COMBINATION

Description: In using the Zircex Process head-end to recover Zr from fuel elements before recovery of U by fluoride volatility, loss of U as UCl/sub 4/ is a problem. The suitability of using a Ni wire filter for trapping entrained UCl/sub 4/, and recovery of the trapped U by direct fluorination of filter and residue is investigated. It is recommended from these studies that water vapor and oxygen should be kept out of the reactor, and that the filter and filter material should be further investigated. (T.R.H.)
Date: March 20, 1958
Creator: Ammann, P.R.; Madden, D.A. & Swift, D.L.
Partner: UNT Libraries Government Documents Department

HIGH TEMPERATURE CLADDING ALLOYS

Description: For high-temperature applications, fuel element cladding materials are required to protect the base metal structure and to contain the fuel so as to minimize fission product release and fuel loss. Since the early work on clad Mo fuel element concepts, many metallic cladding systems have been investigated for direct cycle reactors. The developments in stainless steel and Ni alloy claddings are reviewed and the more recent applications of new oxidation- resistant ferrous-base and chromium-base alloys as claddings for refractory metals are presented. (auth)
Date: August 1, 1960
Creator: Collins, J.F. & McGurty, J.A.
Partner: UNT Libraries Government Documents Department

Spectrophotometric investigation of U(VI) chloride complexation in the NaCl/NaClO{sub 4} system

Description: The option of a nuclear waste disposal in deep salt formations such as Gorleben in Germany, and the WIPP (Waste Isolation Pilot Plant) in southeastern New Mexico, US has generated, over the last ten years, interests in thermodynamic data of radioactive trace elements in concentrated electrolyte solutions. Furthermore, post closure radioactive release scenarios from geologic salt formation, such as the WIPP include hydrologic transport of radionuclides through a chloride saturated aquifer. Consequently, the understanding of actinide solution chemistry in brines is essential for modeling requiring accurate knowledge of the interaction between AnO{sub 2}{sup 2+} and chloride ions. The goal is to develop models capable of predicting their behavior in natural multicomponent brines of high concentration. Complexation constants of two U(VI) chloride species, UO{sub 2}Cl{sup +} and UO{sub 2}Cl{sub 2}{sup 0}, have been intensively studied for about 40 years using different methods. However, large uncertainties reflect the general difficulty in determining accurate stability constants of weak complexes. In order to model the behavior of U(VI) in brines, the authors studied the formation of its chloride complexes by UV-Vis spectroscopy as a function of the NaCl concentration at 25 C.
Date: February 1, 1999
Creator: Paviet-Hartmann, P.; Lin, M.R. & Runde, W.H.
Partner: UNT Libraries Government Documents Department

Microwave combustion and sintering without isostatic pressure

Description: In recent years interest has grown rapidly in the application of microwave energy to the processing of ceramics, composites, polymers, and other materials. Advances in the understanding of microwave/materials interactions will facilitate the production of new ceramic materials with superior mechanical properties. One application of particular interest is the use of microwave energy for the mobilization of uranium for subsequent redeposition. Phase III (FY98) will focus on the microwave assisted chemical vapor infiltration tests for mobilization and redeposition of radioactive species in the mixed sludge waste. Uranium hexachloride and uranium (IV) borohydride are volatile compounds for which the chemical vapor infiltration procedure might be developed for the separation of uranium. Microwave heating characterized by an inverse temperature profile within a preformed ceramic matrix will be utilized for CVI using a carrier gas. Matrix deposition is expected to commence from the inside of the sample where the highest temperature is present. The preform matrix materials, which include aluminosilicate based ceramics and silicon carbide based ceramics, are all amenable to extreme volume reduction, densification, and vitrification. Important parameters of microwave sintering such as frequency, power requirement, soaking temperature, and holding time will be investigated to optimize process conditions for the volatilization of uranyl species using a reactive carrier gas in a microwave chamber.
Date: January 1, 1998
Creator: Ebadian, M.A.
Partner: UNT Libraries Government Documents Department

Method for Making a Uranium Chloride Salt Product

Description: The subject apparatus provides a means to produce UCl3, in large quantities without incurring corrosion of the containment vessel or associated apparatus. Gaseous Cl is injected into a lower layer of Cd where CdCl2 is formed. Due to is lower density, the CdCl2 rises through the Cd layer into a layer of molten LiCl-KCL salt where a rotatable basket containing uranium ingots is suspended. The CdCl2 reacts with the uranium to form UCl, and Cd. Due to density differences, the Cd sinks down to the liquid Cd layer and is reused. The UCl3 combines with the molten salt. During production the temperature is maintained at about 600 degrees C. while after the uranium has been depleted the salt temperature is lowered, the molten salt is pressure siphoned from the vessel, and the salt product LiCl-KCl-30 mol% UCl3 is solidified.
Date: October 5, 2004
Creator: Miller, William F. & Tomczuk, Zygmunt
Partner: UNT Libraries Government Documents Department

CHEMICAL PROBLEMS OF NON-AQUEOUS FLUID-FUEL REACTORS

Description: The three main chemical problems of non-aqueous fluidfuel reactors are selection of a fuel system which meets nuclear and thermal requirements, control of corrosion of structural materials, and development of an efficient and economical separation process. The problems of the fastneutron reactor with a core of fused NaCl, PbCl/sub 2/, UCl/sub 4/ mixture and a blanket of fused UCl/ sub 4/ are discussed, except for the separation process. Brief treatment is given the Bi- U - Pu fuel system for thermal reactors, including tentative flowsheets for the separation process. A detailed discussion of the experiments of Bareis at Brookhaven and related experiments on the distribution of various metals between liquid Bi and fused-salt solutions is given, the experimental results correlated, and application made to reactor problems. A general discussion is given of nonaqueous high-temperature separation processes (L.M.T.)
Date: October 15, 1952
Creator: Scatchard, G.; Clark, H.M.; Golden, S.; Boltax, A. & Schuhmann, R. Jr.
Partner: UNT Libraries Government Documents Department

Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

Description: This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day operations as well as obtaining historical information. Over 12,000 driver ...
Date: May 1, 2007
Creator: Pope, Chad
Partner: UNT Libraries Government Documents Department

LABORATORY DEVELOPMENT OF A COMBINED CHLORIDE VOLATILITY-AQUEOUS PROCESSING METHOD FOR URANIUM-ZIRCONIUM NUCLEAR FUELS

Description: The operations in a process proposed for recovering uranium from spent uranium-- zirconium alloy fuels, including collecting the volatilized chlorination products (mainiy zirconium tetrachioride and uranium pentachloride) in boiling water, concentrating the resulting solution, lowering the freezing point by removing chloride with hydrogen peroxide, and recovering uranium from the 5 M Zr product solution by solvent extraction with tributyl phosphate in Amsco diluent, were investigated in the laboratory and appeared to be reducible to large-scale practice. The high temperature chlorination equipment would also be adaptable for burning graphite matrix fuels and when combined with Darex equipment for processing fuels containing stainless steel, molybdenum, or aluminum may provide the basis for a feasible universal fuel processing system. (auth)
Date: October 15, 1963
Creator: Gens, T.A.
Partner: UNT Libraries Government Documents Department

DYNAMIC CORROSION SCREENING TESTS ON INCONEL AND NICKEL IN NaCl-MgCl$sub 2$- UCl$sub 3$Bath

Description: Nickel is more susceptible to mass transfer ina 100hr nonisothermal dynamic corrosion system than is Inconel when exposed to a NaCl-MgCl/sub 2/-UCl/ sub 3/ (50.01800 F. No nickel mass transfer was observed in a 500hr test at 1350 F, but Inconel showed some attack under s transfer was observed in both tests, being more severe at the higher temperature. On the bases of these preliminary tests, it appears that nickel is a more satisfactory container than Inconel for the chloride bath at temperatures in the region of 1350 F. Chromium is more susceptible to selective leaching from inconel at 1800 F during a short 100-hr test (0.26%Cr in bath) than in a 500-hr test (<0.001% Cr in bath) at a lower temperature (1350 ). (auth)
Date: June 19, 1957
Creator: Jansen, D.H.
Partner: UNT Libraries Government Documents Department

EVALUATION OF GAMMA-RAY ATTENUATION TECHNIQUES FOR MEASURING THE DENSITY AND HOMOGENEITY OF THORIUM OXIDE SLURRIES CIRCULATING AT 300 C

Description: The feasftbility of using gamma -ray attenuation techniques was demonstrated in determining the density and homogeneity of slurries circulating at 300 deg C. The method appears to have general utility in the engineering studies of slurry because the equipment is mounted externally and does not interfere with the system under investigation. Two types of detector systems were evaluated for routine service. The decimal counter-recorder system was found more adaptable for rugged operations, whereas the single channel analyzer was fouW to be more accurate and flexible. (auth)
Date: October 10, 1957
Creator: Richardson, M. & Kitzes, A.S.
Partner: UNT Libraries Government Documents Department