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Plans for the development of the IFR (Integral Fast Reactor) fuel cycle

Description: The Integral Fast Reactor (IFR) is a concept for a self-contained facility in which several sodium-cooled fast reactors of moderate size are located at the same site along with complete fuel-recycle and waste-treatment facilities. After the initial core loading with enriched uranium or plutonium, only natural or depleted uranium is shipped to the plant, and only wastes in final disposal forms are shipped out. The reactors have driver and blanket fuels of uranium-plutonium-zirconium alloys in stainless steel cladding. The use of metal alloy fuels is central to the IFR concept, contributing to the inherent safety of the reactor, the ease of reprocessing, and the relatively low capital and operating costs. Discharged fuels are recovered in a pyrochemical process that consists of two basic steps: an electrolytic process to separate fission products from actinides, and halide slagging to separate plutonium from uranium.
Date: January 1, 1986
Creator: Johnson, T.R.
Partner: UNT Libraries Government Documents Department

Immobilization of IFR salt wastes in mortar

Description: Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered.
Date: January 1, 1988
Creator: Fischer, D.F. & Johnson, T.R.
Partner: UNT Libraries Government Documents Department

Evaluation of available MHD seed-regeneration processes on the basis of energy considerations

Description: Of the several processes described in the literature that are capable of separating sulfur from alkali-metal sulfates, seven processes were selected as candidates for regenerating seed material for reuse in open-cycle MHD. After a brief assessment of each process, two were selected for a detailed analysis, namely, a process developed by the Pittsburgh Energy Research Center (PERC) and a modified version of the Tampella process. The processes were compared on the bases of energy requirements and the amount of research work needed to develop a seed-regeneration process for MHD systems. The energy requirements given should be considered as rough values, because factors such as heat losses and component efficiency were not included in the analysis. On the basis of energy consumption, the PERC process has a slight advantage over the Tampella process; on the basis of the present state of development of various components, the Tampella process has a clear advantage. Accordingly, it was recommended that developmental programs be carried out for both the PERC and Tampella processes.
Date: September 1, 1978
Creator: Sheth, A.C. & Johnson, T.R.
Partner: UNT Libraries Government Documents Department

Proposed methods for treating high-level pyrochemical process wastes. [Integral Fast Reactor]

Description: This survey illustrates the large variety and number of possible techniques available for treating pyrochemical wastes; there are undoubtedly other process types and many variations. The choice of a suitable process is complicated by the uncertainty as to what will be an acceptable waste form in the future for both TRU and non-TRU wastes.
Date: January 1, 1985
Creator: Johnson, T.R.; Miller, W.E. & Steunenberg, R.K.
Partner: UNT Libraries Government Documents Department

Choice of pyroprocess for Integral Fast Reactor fuel

Description: A design objective for the Integral Fast Reactor (IFR) is fuel self sufficiency. This can be achieved only by employing chemical reprocessing as part of the fuel cycle. Because the fuel is a metal alloy (U-Pu-Zr), direct production of metal is highly advantageous. This makes a pyrometallurgical process attractive. (JDB)
Date: January 1, 1985
Creator: Miller, W.E.; Johnson, T.R. & Tomczuk, Z.
Partner: UNT Libraries Government Documents Department

Condensation and deposition of seed in the MHD bottoming plant

Description: The computer models of slag vapor nucleation and particle deposition have been extended to predict the growth and deposition of seed particles in the steam and air heater sections of the MHD bottoming plant. The model represents a hot combustion gas stream, which contains vaporized seed and entrained slag particles of a selected initial size distribution, flowing through a bank of cooled tubes. The energy balance includes convective and radiant heat transfer to the cool surfaces. The material balance for the condensible species considers convective mass transport of seed vapor to cool surfaces, and the deposition of particles on cooled surfaces by thermophoresis. The analyses provide the bases for design trade-off studies of steam tube size and spacing, gas velocity, and system configuration to optimize the effectiveness and cost of the steam plant. In the absence of entrained slag particles, sample calculations indicated that, as the gas is cooled in passing through a tube bank, the bulk of the seed vapor condenses in the gas stream to form particles with diameters in the range of 0.02 to 0.2 ..mu..m. In the presence of the submicron slag particles formed upstream in the MHD diffuser, the largest fraction of the seed vapor condenses on the existing entrained particles, causing them to grow to a size in the range of approximately one micron. In both cases, these particles are deposited on heat exchange surfaces throughout the heat recovery system and a large fraction is present in the cool combustion gas entering the exhaust gas clean-up system.
Date: January 1, 1979
Creator: Im, K.H.; Patten, J.; Johnson, T.R. & Tempelmeyer, K.
Partner: UNT Libraries Government Documents Department

Process to remove rare earth from IFR electrolyte

Description: The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.
Date: January 1, 1992
Creator: Ackerman, J. P. & Johnson, T. R.
Partner: UNT Libraries Government Documents Department

Vitrification of low-level and mixed wastes

Description: The US Department of Energy (DOE) and nuclear utilities have large quantities of low-level and mixed wastes that must be treated to meet repository performance requirements, which are likely to become even more stringent. The DOE is developing cost-effective vitrification methods for producing durable waste forms. However, vitrification processes for high-level wastes are not applicable to commercial low-level wastes containing large quantities of metals and small amounts of fluxes. New vitrified waste formulations are needed that are durable when buried in surface repositories.
Date: December 31, 1994
Creator: Johnson, T.R.; Bates, J.K. & Feng, Xiangdong
Partner: UNT Libraries Government Documents Department

Testing of pyrochemical centrifugal contactors

Description: A centrifugal contactor that performs oxidation and reduction exchange reactions between molten metals and salts at 500 degrees Centigrade has been tested successfully at Argonne National Laboratory (ANL). The design is based on contactors for aqueous- organic systems operation near room temperature. In tests to demonstrate the performance of the pyrocontactor, cadmium and LICl- KCl eutectic salt were the immiscible solvent phases, and rare earths were the distributing solutes. The tests showed that the pyrocontactor mixed and separated the phases well, with stage efficiencies approaching 99% at rotor speeds near 2700 rpm. The contactor ran smoothly and reliably over the entire range of speeds that was tested.
Date: August 1996
Creator: Chow, L. S.; Carls, E. L.; Basco, J. K. & Johnson, T. R.
Partner: UNT Libraries Government Documents Department

Performance of high plutonium-containing glasses for the immobilization of surplus fissile materials

Description: Plutonium from dismantled weapons is being evaluated for geological disposal. While a final waste form has not been chosen, borosilicate glass will be one of the waste forms to be evaluated. The reactivity of the reference blend glass containing the standard amount of Pu ({approximately}0.01 wt %) to be produced by the Defense Waste Processing Facility (DWPF) is compared to that of glasses made from the same nominal frit composition but doped with 2 and 7 wt % Pu, and also equal mole percentages of Gd{sub 2}O{sub 3}. The Gd is added to act as a neutron poison to address criticality concerns. The four different glasses have been reacted using the PCT-B method with a SA/V of 20,000 m{sup {minus}1} and the Argonne Vapor Hydration Test (VHT) method. Both test methods accelerate the reaction of the glass. PCT-B is used to determine the reactivity of the glass by analyzing the solution and reacted test components, while the VHT is used to evaluate the long-term reactivity of the glass and the distribution of Pu to secondary phases that will control the long-term reaction of the glass. The results of the tests with high levels of Pu are compared to those with the nominal levels to be produced in the standard DWPF glass.
Date: July 1, 1995
Creator: Bates, J.K.; Emery, J.W.; Hoh, J.C. & Johnson, T.R.
Partner: UNT Libraries Government Documents Department

Treatment of high-level wastes from the IFR fuel cycle

Description: The Integral Fast Reactor (IFR) is being developed as a future commercial power source that promises to have important advantages over present reactors, including improved resource conservation and waste management. The spent metal alloy fuels from an IFR will be processed in an electrochemical cell operating at 500{degree}C with a molten chloride salt electrolyte and cadmium metal anode. After the actinides have been recovered from several batches of core and blanket fuels, the salt cadmium in this electrorefiner will be treated to separate fission products from residual transuranic elements. This treatment produces a waste salt that contains the alkali metal, alkaline earth, and halide fission products; some of the rare earths; and less than 100 nCi/g of alpha activity. The treated metal wastes contain the rest of the fission products (except T, Kr, and Xe) small amounts of uranium, and only trace amounts of transuranic elements. The current concept for the salt waste form is an aluminosilicate matrix, and the concept for the metal waste form is a corrosion-resistant metal alloy. The processes and equipment being developed to treat and immobilize the salt and metal wastes are described.
Date: January 1, 1992
Creator: Johnson, T.R.; Lewis, M.A.; Newman, A.E. & Laidler, J.J.
Partner: UNT Libraries Government Documents Department

Effective method for MHD retrofit of power plants

Description: Retrofitting existing power plants with an open-cycle MHD system has been re-examined in light of recent developments in the heat and seed recovery technology area. A new retrofit cycle configuration has been developed which provides for a direct gas-gas coupling; also, the MHD topping cycle can be decoupled from the existing plant for either separate or joint operation. As an example, the MHD retrofit concept has been applied to Illinois Power Company's Vermilion Station No. 1, a coal-fired power plant presently in operation. Substantial increases in efficiency have been demonstrated and the economic validity of the MHD retrofit approach has been established.
Date: October 1, 1981
Creator: Berry, G.F.; Dennis, C.B.; Johnson, T.R. & Minkov, V.
Partner: UNT Libraries Government Documents Department

Actinide consumption: Nuclear resource conservation without breeding

Description: A new approach to the nuclear power issue based on a metallic fast reactor fuel and pyrometallurgical processing of spent fuel is showing great potential and is approaching a critical demonstration phase. If successful, this approach will complement and validate the LWR reactor systems and the attendant infrastructure (including repository development) and will alleviate the dominant concerns over the acceptability of nuclear power. The Integral Fast Reactor (IFR) concept is a metal-fueled, sodium-cooled pool-type fast reactor supported by a pyrometallurgical reprocessing system. The concept of a sodium cooled fast reactor is broadly demonstrated by the EBR-II and FFTF in the US; DFR and PFR in the UK; Phenix and SuperPhenix in France; BOR-60, BN-350, BN-600 in the USSR; and JOYO in Japan. The metallic fuel is an evolution from early EBR-II fuels. This fuel, a ternary U-Pu-Zr alloy, has been demonstrated to be highly reliable and fault tolerant even at very high burnup (160-180,000 MWd/MT). The fuel, coupled with the pool type reactor configuration, has been shown to have outstanding safety characteristics: even with all active safety systems disabled, such a reactor can survive a loss of coolant flow, a loss of heat sink, or other major accidents. Design studies based on a small modular approach show not only its impressive safety characteristics, but are projected to be economically competitive. The program to explore the feasibility of actinide recovery from spent LWR fuel is in its initial phase, but it is expected that technical feasibility could be demonstrated by about 1995; DOE has not yet committed funds to achieve this objective. 27 refs.
Date: January 1, 1991
Creator: Hannum, W.H.; Battles, J.E.; Johnson, T.R. & McPheeters, C.C.
Partner: UNT Libraries Government Documents Department

Waste removal in pyrochemical fuel processing for the Integral Fast Reactor

Description: Electrorefining in a molten salt electrolyte is used in the Integral Fast Reactor fuel cycle to recover actinides from spent fuel. Processes that are being developed for removing the waste constituents from the electrorefiner and incorporating them into the waste forms are described in this paper. During processing, halogen, chalcogen, alkali, alkaline earth, and rare earth fission products build up in the molten salt as metal halides and anions, and fuel cladding hulls and noble metal fission products remain as metals of various particle sizes. Essentially all transuranic actinides are collected as metals on cathodes, and are converted to new metal fuel. After processing, fission products and other waste are removed to a metal and a mineral waste form. The metal waste form contains the cladding hulls, noble metal fission products, and (optionally) most rare earths in a copper or stainless steel matrix. The mineral waste form contains fission products that have been removed from the salt into a zeolite or zeolite-derived matrix.
Date: January 1, 1994
Creator: Ackerman, J. P.; Johnson, T. R. & Laidler, J. J.
Partner: UNT Libraries Government Documents Department

Measurements of Transverse Beam Diffusion Rates in the Fermilab Tevatron Collider

Description: The transverse beam diffusion rate vs. particle oscillation amplitude was measured in the Tevatron using collimator scans. All collimator jaws except one were retracted. As the jaw of interest was moved in small steps, the local shower rates were recorded as a function of time. By using a diffusion model, the time evolution of losses could be related to the diffusion rate at the collimator position. Preliminary results of these measurements are presented.
Date: August 1, 2011
Creator: Stancari, G.; Annala, G.; Johnson, T.R.; Still, D.A.; Valishev, A. & /Fermilab
Partner: UNT Libraries Government Documents Department

Magnesium transport extraction of transuranium elements from LWR fuel

Description: This report discusses a process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl{sub 2} and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800{degrees}C to about 850{degrees}C to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl{sub 2} having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO{sub 2}. The Ca metal and CaCl{sub 2} is recycled to reduce additional oxide fuel. The U-Fe alloy having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with Mg metal which takes up the actinide and rare earth fission product metals. The U-Fe alloy retains the noble metal fission products and is stored while the Mg is distilled and recycled leaving the transuranium actinide and rare earth fission products isolated.
Date: December 31, 1991
Creator: Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Miller, W.E. & Pierce, R.D.
Partner: UNT Libraries Government Documents Department

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

Salt transport extraction of transuranium elements from LWR fuel

Description: This report discusses a process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl{sub 2} and a Cu-Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750{degrees}C to about 850{degrees}C to precipitate uranium metal and some of the noble metal fission products leaving the Cu-Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl{sub 2} having Cao and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO{sub 2}. The Ca metal and CaCl{sub 2} is recycled to reduce additional oxide fuel. The Cu-Mg alloy having transuranium metals and rare earth fission product metals and the noble metal fission products dissolved therein is contacted with a transport salt including Mg C1{sub 2} to transfer Mg values from the transport salt to the Cu-Mg alloy .hile transuranium actinide and rare earth fission product metals transfer from the Cu-Mg alloy to the transport salt. Then the transport salt is mixed with a Mg-Zn alloy to transfer Mg values from the alloy to the transport salt while the transuranium actinide and rare earth fission product values dissolved in the salt are reduced and transferred to the Mg-Zn alloy.
Date: December 31, 1991
Creator: Pierce, R.D.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R. & Miller, W.E.
Partner: UNT Libraries Government Documents Department

IFR fuel cycle--pyroprocess development

Description: The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs.
Date: January 1, 1992
Creator: Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P. & Battles, J.E.
Partner: UNT Libraries Government Documents Department