National Library of Energy BETA

Sample records for nuclear fuel reprocessing

  1. Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    Michael F. Simpson; Jack D. Law

    2010-02-01

    This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

  2. Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore. Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor

  3. Spent nuclear fuel reprocessing modeling

    SciTech Connect (OSTI)

    Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V.; Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I.

    2013-07-01

    The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

  4. Summary of nuclear fuel reprocessing activities around the world

    SciTech Connect (OSTI)

    Mellinger, P.J.; Harmon, K.M.; Lakey, L.T.

    1984-11-01

    This review of international practices for nuclear fuel reprocessing was prepared to provide a nontechnical summary of the current status of nuclear fuel reprocessing activities around the world. The sources of information are widely varied.

  5. Classic Nuclear Fuel Reprocessing Flowsheet

    SciTech Connect (OSTI)

    Fallgren, Andrew James

    2015-02-13

    This is a flowsheet as well as a series of subsheets to be used for discussion on the standard design of a reprocessing plant. This flowsheet consists of four main sections: offgas handling, separations, solvent wash, and acid recycle. As well as having the main flowsheet, subsections have been broken off into their own sheets to provide for larger font and ease of printing.

  6. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    Jubin, R. T.; Strachan, D. M.; Ilas, G.; Spencer, B. B.; Soelberg, N. R.

    2014-09-01

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is

  7. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    SciTech Connect (OSTI)

    Soelberg, Nicolas R.; Garn, Troy; Greenhalgh, Mitchell; Law, Jack; Jubin, Robert T.; Strachan, Denis M.; Thallapally, Praveen K.

    2013-07-22

    Nuclear fission results in the production of fission products and activation products, some of which tend to be volatile during used fuel reprocessing. These can evolve in volatile species in the reprocessing facility off-gas streams, depending on the separations and reprocessing technologies that are used. Radionuclides that have been identified as “volatile radionuclides” are noble gases (most notably isotopes of Kr and Xe); 3H; 14C; and 129I. Radionuclides that tend to form volatile species that evolve into reprocessing facility off-gas systems are more challenging to efficiently control compared to radionuclides that tend to stay in solid or liquid phases. Future used fuel reprocessing facilities in the United States can require efficient capture of some volatile radionuclides in their off-gas streams to meet regulatory emission requirements. In aqueous reprocessing, these radionuclides are most commonly expected to evolve into off-gas streams in tritiated water [3H2O (T2O) and 3HHO (THO)], radioactive CO2, noble gases, and gaseous HI, I2, or volatile organic iodides. The fate and speciation of these radionuclides from a non-aqueous fuel reprocessing facility is less well known at this time, but active investigations are in progress. An Off-Gas Sigma Team was formed in late FY 2009 to integrate and coordinate the Fuel Cycle Research and Development (FCR&D) activities directed towards the capture and sequestration of the these volatile radionuclides (Jubin 2012a). The Sigma Team concept was envisioned to bring together multidisciplinary teams from across the DOE complex that would work collaboratively to solve the technical challenges and to develop the scientific basis for the capture and immobilization technologies such that the sum of the efforts was greater than the individual parts. The Laboratories currently participating in this effort are Argonne National Laboratory (ANL), Idaho National Laboratory (INL), Oak Ridge National Laboratory (ORNL), Pacific

  8. Method for reprocessing and separating spent nuclear fuels

    DOE Patents [OSTI]

    Krikorian, Oscar H.; Grens, John Z.; Parrish, Sr., William H.

    1983-01-01

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  9. Method for reprocessing and separating spent nuclear fuels. [Patent application

    DOE Patents [OSTI]

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and nonvolatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

  10. Radioactive Iodine and Krypton Control for Nuclear Fuel Reprocessing Facilities

    SciTech Connect (OSTI)

    N. R. Soelberg; J. D. Law; T. G. Garn; M. Greenhalgh; R. T. Jubin; P. Thallapally; D. M. Strachan

    2013-08-01

    The removal of volatile radionuclides generated during used nuclear fuel reprocessing in the US is almost certain to be necessary for the licensing of a reprocessing facility in the US. Various control technologies have been developed, tested, or used over the past 50 years for control of volatile radionuclide emissions from used fuel reprocessing plants. The US DOE has sponsored, since 2009, an Off-gas Sigma Team to perform research and development focused on the most pressing volatile radionuclide control and immobilization problems. In this paper, we focus on the control requirements and methodologies for 85Kr and 129I. Numerous candidate technologies have been studied and developed at laboratory and pilot-plant scales in an effort to meet the need for high iodine control efficiency and to advance alternatives to cryogenic separations for krypton control. Several of these show promising results. Iodine decontamination factors as high as 105, iodine loading capacities, and other adsorption parameters including adsorption rates have been demonstrated under some conditions for both silver zeolite (AgZ) and Ag-functionalized aerogel. Sorbents, including an engineered form of AgZ and selected metal organic framework materials (MOFs), have been successfully demonstrated to capture Kr and Xe without the need for separations at cryogenic temperatures.

  11. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOE Patents [OSTI]

    Campbell, D.O.; Buxton, S.R.

    1980-06-16

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  12. Process for recovery of palladium from nuclear fuel reprocessing wastes

    DOE Patents [OSTI]

    Campbell, David O.; Buxton, Samuel R.

    1981-01-01

    Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M, (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound, (c) heating the solution at reflux temperature until precipitation is complete, and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

  13. Molten tin reprocessing of spent nuclear fuel elements

    DOE Patents [OSTI]

    Heckman, Richard A.

    1983-01-01

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

  14. MICROBIAL TRANSFORMATIONS OF RADIONUCLIDES RELEASED FROM NUCLEAR FUEL REPROCESSING PLANTS.

    SciTech Connect (OSTI)

    FRANCIS,A.J.

    2006-10-18

    Microorganisms can affect the stability and mobility of the actinides U, Pu, Cm, Am, Np, and the fission products Tc, I, Cs, Sr, released from nuclear fuel reprocessing plants. Under appropriate conditions, microorganisms can alter the chemical speciation, solubility and sorption properties and thus could increase or decrease the concentrations of radionuclides in solution and the bioavailability. Dissolution or immobilization of radionuclides is brought about by direct enzymatic action or indirect non-enzymatic action of microorganisms. Although the physical, chemical, and geochemical processes affecting dissolution, precipitation, and mobilization of radionuclides have been investigated, we have only limited information on the effects of microbial processes. The mechanisms of microbial transformations of the major and minor actinides and the fission products under aerobic and anaerobic conditions in the presence of electron donors and acceptors are reviewed.

  15. Idaho Site Completes Demolition of Cold War-era Nuclear Fuel Reprocessing

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Facility | Department of Energy Completes Demolition of Cold War-era Nuclear Fuel Reprocessing Facility Idaho Site Completes Demolition of Cold War-era Nuclear Fuel Reprocessing Facility December 22, 2011 - 11:12am Addthis Media Contact Erik Simpson (208) 360-0426 A gravel mound, larger than half a city block and several feet thick, is the only visible feature that remains at the site of a Cold War-era spent nuclear fuel reprocessing facility at the U.S. Department of Energy's Idaho site.

  16. Overview of reductants utilized in nuclear fuel reprocessing/recycling

    SciTech Connect (OSTI)

    Patricia Paviet-Hartmann; Catherine Riddle; Keri Campbell; Edward Mausolf

    2013-10-01

    promises as a replacement for AHA. FHA undergoes hydrolysis to formic acid which is volatile, thus allowing the recycling of nitric acid. Unfortunately, FHA powder was not stable in the experiments we ran in our laboratory. In addition, AHA and FHA also decompose to hydroxylamine which may undergo an autocatalytic reaction. Other reductants are available and could be extremely useful for actinides separation. The review presents the current plutonium reductants used in used nuclear fuel reprocessing and will introduce innovative and novel reductants that could become reducers for future research on UNF separation.

  17. Composition of high fission product wastes resulting from future reprocessing of commercial nuclear fuels

    SciTech Connect (OSTI)

    Swanson, J.L

    1986-07-01

    Pacific Northwest Laboratory studies, aimed at defining appropriate glass compositions for future disposal of high-level wastes, have developed composition ranges for the waste that will likely result during reprocessing of Light Water Reactor (LWR) and Liquid Metal Reactor (LMR) fuels. The purpose of these studies was to provide baseline waste characterizations for possible future commercial high-level waste so that waste immobilization technologies (e.g., vitrification) can be studied. Ranges in waste composition are emphasized because the waste will vary with time as different fuels are reprocesses, because choice of process chemicals is nuclear, and because fuel burnups will vary. Consequently, composition ranges are based on trends in fuel reprocessing procedures and on achievable burnups in operating reactors. In addition to the fission product and actinide elements, which are the primary hazardous materials in the waste, likely composition ranges are given for inert elements that may be present in the waste. These other elements may be present because of being present in the fuel, because of being added as process chemical during reprocessing, because of being added during equipment decontamination, or because of corrosion of plant equipment and/or fuel element cladding. This report includes a discussion of the chemicals added in variation of the PUREX process, which is likely to remain the favored reprocessing technique for commercial nuclear fuels. Consideration is also given to a pyrochemical process proposed for the reprocessing of some LMR fuels.

  18. The Economics of Reprocessing Versus Direct Disposal of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Bunn, Matthew; Holdren, John P.; Fetter, Steve; Zwaan, Bob van der

    2005-06-15

    We assess the economics of reprocessing versus direct disposal of spent fuel. The uranium price at which reprocessing spent fuel from light water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is estimated for a range of reprocessing prices and other fuel cycle costs. The contribution of both fuel cycle options to the cost of electricity is also estimated. A similar analysis is performed to compare fast neutron reactors (FRs) with LWRs. We review available information about various fuel cycle costs, as well as the quantities of uranium likely to be recoverable at a range of future prices. We conclude that the once-through LWR fuel cycle is likely to remain significantly cheaper than recycling in either LWRs or FRs for at least the next 50 yr, even with substantial growth in nuclear power.

  19. Idaho site completes demolition of Cold War-era nuclear fuel reprocessing

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    facility doe logo CH2M-WG logo Joint News Release For Immediate Release Date: December 22, 2011 Media Contact: Erik Simpson, (208) 360-0426 Idaho site completes demolition of Cold War-era nuclear fuel reprocessing facility A gravel mound, larger than half a city block and several feet thick, is the only visible feature that remains at the site of a Cold War-era spent nuclear fuel reprocessing facility at the U.S. Department of Energy�s Idaho site. About $44 million in American Recovery

  20. Method for cleaning solution used in nuclear fuel reprocessing

    DOE Patents [OSTI]

    Tallent, O.K.; Crouse, D.J.; Mailen, J.C.

    1980-12-17

    Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

  1. Method for cleaning solution used in nuclear fuel reprocessing

    DOE Patents [OSTI]

    Tallent, Othar K. (Oak Ridge, TN); Crouse, David J. (Oak Ridge, TN); Mailen, James C. (Oak Ridge, TN)

    1982-01-01

    Nuclear fuel processing solution consisting of tri-n-butyl phosphate and dodecane, with a complex of uranium, plutonium, or zirconium and with a solvent degradation product such as di-n-butyl phosphate therein, is contacted with an aqueous solution of a salt formed from hydrazine and either a dicarboxylic acid or a hydroxycarboxylic acid, thereby removing the aforesaid complex from the processing solution.

  2. New method of uranium and plutonium extraction in reprocessing of the spent nuclear fuel

    SciTech Connect (OSTI)

    Volk, V.; Dvoeglazov, K.; Veslov, S.; Rubisov, V.; Alekseenko, V.; Krivitsky, Y.; Alekseenko, S.; Bondin, V.

    2013-07-01

    It is shown that a two-stage process of uranium and plutonium extraction during the reprocessing of spent nuclear fuel solves the problem of obtaining a high-concentrated extract without increasing the loss risk with raffinate and avoids the accumulation of plutonium in the unit. A possible further optimization of the process would be the creation of steps inside the stages.

  3. Method for cleaning solution used in nuclear fuel reprocessing

    DOE Patents [OSTI]

    Tallent, Othar K. (Oak Ridge, TN); Dodson, Karen E. (Knoxville, TN); Mailen, James C. (Oak Ridge, TN)

    1983-01-01

    A nuclear fuel processing solution containing (1) hydrocarbon diluent, (2) tri-n-butyl phosphate or tri-2-ethylhexyl phosphate, and (3) monobutyl phosphate, dibutyl phosphate, mono-2-ethylhexyl phosphate, di-2-ethylhexyl phosphate, or a complex formed by plutonium, uranium, or a fission product thereof with monobutyl phosphate, dibutyl phosphate, mono-2-ethylhexyl phosphate, or di-2-ethylhexyl phosphate is contacted with silica gel having alkali ions absorbed thereon to remove any one of the degradation products named in section (3) above from said solution.

  4. An option making for nuclear fuel reprocessing by using supercritical carbon dioxide

    SciTech Connect (OSTI)

    Enokida, Youichi; Sawada, Kayo; Shimada, Takashi; Yamamoto, Ichiro

    2007-07-01

    A four-year-research has been completed as a collaborative work by Nagoya University Mitsubishi Heavy Industries Corporation and Japan Atomic Energy Agency (JAEA) in order to develop a super critical carbon dioxide (SF-CO{sub 2}) based technology, 'SUPER-DIREX process', for nuclear fuel reprocessing. As a result obtained in Phase II of the Japan's feasibility Studies on Commercialized Fast Reactor Cycle Systems, this technology was evaluated as one of the alternatives for the advanced Purex process for he future FBR fuel cycle. Although further investigation is required for a scaled-up demonstration of processing spent fuels by SUPER-DIREX process, we could conclude that an option has been made for nuclear fuel reprocessing by using supercritical carbon dioxide. (authors)

  5. THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL

    SciTech Connect (OSTI)

    Matthew Bunn; Steve Fetter; John P. Holdren; Bob van der Zwaan

    2003-07-01

    This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.

  6. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

    2013-11-05

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  7. Advanced dry head-end reprocessing of light water reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

    2014-06-10

    A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

  8. Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

    DOE Patents [OSTI]

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.

    1982-01-19

    Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in the reaction region of a separation vessel which includes a reflux region positioned above the molten tin solvent. The reflux region minimizes loss of evaporated solvent during the separation of the actinide fuels from the volatile fission products. Additionally, inclusion of the reflux region permits the separation of the more volatile fission products (noncondensable) from the less volatile ones (condensable).

  9. Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions

    DOE Patents [OSTI]

    Horwitz, E. Philip; Delphin, Walter H.

    1979-07-24

    A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

  10. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1992-01-01

    A single canister process container for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining their integrity at temperature necessary to oxide the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container.

  11. Molten tin reprocessing of spent nuclear fuel elements. [Patent application; continuous process

    DOE Patents [OSTI]

    Heckman, R.A.

    1980-12-19

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support te liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

  12. Apparatus and method for reprocessing and separating spent nuclear fuels. [Patent application

    DOE Patents [OSTI]

    Krikorian, O.H.; Grens, J.Z.; Parrish, W.H. Sr.; Coops, M.S.

    1982-01-19

    A method and apparatus for separating and reprocessing spent nuclear fuels includes a separation vessel housing a molten metal solvent in a reaction region, a reflux region positioned above and adjacent to the reaction region, and a porous filter member defining the bottom of the separation vessel in a supporting relationship with the metal solvent. Spent fuels are added to the metal solvent. A nonoxidizing nitrogen-containing gas is introduced into the separation vessel, forming solid actinide nitrides in the metal solvent from actinide fuels, while leaving other fission products in solution. A pressure of about 1.1 to 1.2 atm is applied in the reflux region, forcing the molten metal solvent and soluble fission products out of the vessel, while leaving the solid actinide nitrides in the separation vessel.

  13. Reprocessing RERTR silicide fuels

    SciTech Connect (OSTI)

    Rodrigues, G.C.; Gouge, A.P.

    1983-05-01

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High-density, low-enrichment, aluminum-clad uranium silicide fuels may be substituted for the highly enriched aluminum-clad alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR silicide fuels at Savannah River Plant. Results of dissolution and feed preparation tests and solvent extraction processing demonstrations with both unirradiated and irradiated uranium silicide fuels are presented.

  14. Container for reprocessing and permanent storage of spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Forsberg, C.W.

    1992-03-24

    A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

  15. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    SciTech Connect (OSTI)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D.

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  16. Development of On-Line Spectroscopic pH Monitoring for Nuclear Fuel Reprocessing Plants: Weak Acid Schemes

    SciTech Connect (OSTI)

    Casella, Amanda J.; Hylden, Laura R.; Campbell, Emily L.; Levitskaia, Tatiana G.; Peterson, James M.; Smith, Frances N.; Bryan, Samuel A.

    2015-05-19

    Knowledge of real-time solution properties and composition is a necessity for any spent nuclear fuel reprocessing method. Metal-ligand speciation in aqueous solutions derived from the dissolved commercial spent fuel is highly dependent upon the acid concentration/pH, which influences extraction efficiency and the resulting speciation in the organic phase. Spectroscopic process monitoring capabilities, incorporated in a counter current centrifugal contactor bank, provide a pathway for on-line real-time measurement of solution pH. The spectroscopic techniques are process-friendly and can be easily configured for on-line applications, while classic potentiometric pH measurements require frequent calibration/maintenance and have poor long-term stability in aggressive chemical and radiation environments. Our research is focused on developing a general method for on-line determination of pH of aqueous solutions through chemometric analysis of Raman spectra. Interpretive quantitative models have been developed and validated under the range of chemical composition and pH using a lactic acid/lactate buffer system. The developed model was applied to spectra obtained on-line during solvent extractions performed in a centrifugal contactor bank. The model predicted the pH within 11% for pH > 2, thus demonstrating that this technique could provide the capability of monitoring pH on-line in applications such as nuclear fuel reprocessing.

  17. The used nuclear fuel problem - can reprocessing and consolidated storage be complementary?

    SciTech Connect (OSTI)

    Phillips, C.; Thomas, I.

    2013-07-01

    This paper describes our CISF (Consolidated Interim Storage Facilities) and Reprocessing Facility concepts and show how they can be combined with a geologic repository to provide a comprehensive system for dealing with spent fuels in the USA. The performance of the CISF was logistically analyzed under six operational scenarios. A 3-stage plan has been developed to establish the CISF. Stage 1: the construction at the CISF site of only a rail receipt interface and storage pad large enough for the number of casks that will be received. The construction of the CISF Canister Handling Facility, the Storage Cask Fabrication Facility, the Cask Maintenance Facility and supporting infrastructure are performed during stage 2. The construction and placement into operation of a water-filled pool repackaging facility is completed for Stage 3. By using this staged approach, the capital cost of the CISF is spread over a number of years. It also allows more time for a final decision on the geologic repository to be made. A recycling facility will be built, this facility will used the NUEX recycling process that is based on the aqueous-based PUREX solvent extraction process, using a solvent of tri-N-butyl phosphate in a kerosene diluent. It is capable of processing spent fuels at a rate of 5 MT per day, at burn-ups up to 50 GWD per ton of spent fuels and a minimum of 5 years out-of-reactor cooling.

  18. Novel sorbent development and evaluation for the capture of krypton and xenon from nuclear fuel reprocessing off-gas stream

    SciTech Connect (OSTI)

    Garn, T.G.; Greenhalgh, M.R.; Law, J.D.

    2013-07-01

    The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, Idaho National Laboratory sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up. (authors)

  19. Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

    SciTech Connect (OSTI)

    Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

    2013-10-01

    The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.

  20. Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

    SciTech Connect (OSTI)

    Troy G. Garn; Mitchell R. Greenhalgh; Jack D. Law

    2013-09-01

    The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.

  1. Corrosion property of 9Cr-ODS steel in nitric acid solution for spent nuclear fuel reprocessing

    SciTech Connect (OSTI)

    Takeuchi, M.; Koizumi, T.; Inoue, M.; Koyama, S.I.

    2013-07-01

    Corrosion tests of oxide dispersion strengthened with 9% Cr (9Cr-ODS) steel, which is one of the desirable materials for cladding tube of sodium-cooled fast reactors, in pure nitric acid solution, spent FBR fuel solution, and its simulated solution were performed to understand the corrosion behavior in a spent nuclear fuel reprocessing. In this study, the 9Cr-ODS steel with lower effective chromium content was evaluated to understand the corrosion behavior conservatively. As results, the tube-type specimens of the 9Cr-ODS steels suffered severe weight loss owing to active dissolution at the beginning of the immersion test in pure nitric acid solution in the range from 1 to 3.5 M. In contrast, the weight loss was decreased and they showed a stable corrosion in the higher nitric acid concentration, the dissolved FBR fuel solution, and its simulated solution by passivation. The corrosion rates of the 9Cr-ODS steel in the dissolved FBR fuel solution and its simulated solution were 1-2 mm/y and showed good agreement with each other. The passivation was caused by the shift of corrosion potential to noble side owing to increase in nitric acid concentration or oxidative ions in the dissolved FBR fuel solution and the simulated spent fuel solution. (authors)

  2. Economic Study of Spent Nuclear Fuel Storage and Reprocessing Practices in Russia

    SciTech Connect (OSTI)

    C. E. Singer; G. H. Miley

    1997-10-01

    This report describes a study of nuclear power economics in Russia. It addresses political and institutional background factors which constrain Russia's energy choices in the short and intermediate run. In the approach developed here, political and institutional factors might dominate short-term decisions, but the comparative costs of Russia's fuel-cycle options are likely to constrain her long-term energy strategy. To this end, the authors have also formulated a set of policy questions which should be addressed using a quantitative decision modeling which analyzes economic costs for all major components of different fuel cycle options, including the evolution of uranium prices.

  3. Equipment specifications for an electrochemical fuel reprocessing plant

    SciTech Connect (OSTI)

    Hemphill, Kevin P

    2010-01-01

    Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

  4. DEVELOPMENT OF A HYDROGEN MORDENITE SORBENT FOR THE CAPTURE OF KRYPTON FROM USED NUCLEAR FUEL REPROCESSING OFF-GAS STREAMS

    SciTech Connect (OSTI)

    Mitchell Greenhalgh; Troy G. Garn; Jack D. Law

    2014-04-01

    A novel new sorbent for the separation of krypton from off-gas streams resulting from the reprocessing of used nuclear fuel has been developed and evaluated. A hydrogen mordenite powder was successfully incorporated into a macroporous polymer binder and formed into spherical beads. The engineered form sorbent retained the characteristic surface area and microporosity indicative of mordenite powder. The sorbent was evaluated for krypton adsorption capacities utilizing thermal swing operations achieving capacities of 100 mmol of krypton per kilogram of sorbent at a temperature of 191 K. A krypton adsorption isotherm was also obtained at 191 K with varying krypton feed gas concentrations. Adsorption/desorption cycling effects were also evaluated with results indicating that the sorbent experienced no decrease in krypton capacity throughout testing.

  5. Uranium transport to solid electrodes in pyrochemical reprocessing of nuclear fuel

    SciTech Connect (OSTI)

    Tomczuk, Z.; Ackerman, J.P.; Wolson, R.D.; Miller, W.E. . Chemical Technology Div.)

    1992-12-01

    A unique pyrochemical process developed for the separation of metallic nuclear fuel from fission products by electrotransport through molten LiCl-KCl eutectic salt to solid and liquid metal cathodes. The process allow for recovery and reuse of essentially all of the actinides in spent fuel from the integral fast reactor (IFR) and disposal of wastes in satisfactory forms. Electrotransport is used to minimize reagent consumption and, consequently, waste volume. In particular, electrotransport to solid cathodes is used for recovery of an essentially pure uranium product in the presence of other actinides; removal of pure uranium is used to adjust the electrolyte composition in preparation for recovery of a plutonium-rich mixture with uranium in liquid cadmium cathodes. This paper presents experiments that delineate the behavior of key actinide and rare-earth elements during electrotransport to a solid electrode over a useful range of PuCl[sub 3]/UCl[sub 3] ratios in the electrolyte, a thermodynamic basis for that behavior, and a comparison of the observed behavior with that calculated from a thermodynamic model. This work clearly established that recovery of nearly pure uranium can be a key step in the overall pyrochemical-fuel-processing strategy for the IFR.

  6. On the possibility of reprocessing spent nuclear fuel and radioactive waste by plasma methods

    SciTech Connect (OSTI)

    Vorona, N. A.; Gavrikov, A. V. Samokhin, A. A.; Smirnov, V. P.; Khomyakov, Yu. S.

    2015-12-15

    The concept of plasma separation of spent nuclear fuel and radioactive waste is presented. An approach that is based on using an accelerating potential to overcome the energy and angular spread of plasma ions at the separation region inlet and utilizing a potential well to separate spatially the ions of different masses is proposed. It is demonstrated that such separation may be performed at distances of about 1 m with electrical potentials of about 1 kV and a magnetic field of about 1 kG. The estimates of energy consumption and performance of the plasma separation method are presented. These estimates illustrate its potential for technological application. The results of development and construction of an experimental setup for testing the method of plasma separation are presented.

  7. A REVIEW OF THORIUM FUEL REPROCESSING EX

    Office of Scientific and Technical Information (OSTI)

    A REVIEW OF THORIUM FUEL REPROCESSING EXPERIENCE R. E. Brooksbank W. T. McDuffee R. H. ... with the current dependence on foreign sources to meet a large fraction of these ...

  8. Metal-Organic Frameworks for Removal of Xe and Kr from Nuclear Fuel Reprocessing Plants

    SciTech Connect (OSTI)

    Liu, Jian; Thallapally, Praveen K.; Strachan, Denis M.

    2012-08-07

    Removal of Xenon (Xe) and Krypton (Kr) from in parts per million (ppm) levels were demonstrated for the first time using two well known metal-organic frameworks (MOFs), HKUST-1 and Ni/DOBDC. Results of an activated carbon were also included for comparison. Ni/DOBDC has higher Xe/Kr selectivities than those of the activated carbon. Moreover, results show that the Ni/DOBDC and HKUST-1 can selectively adsorb Xe and Kr from air even at 1000 ppm concentration. This shows a promising future for MOFs in a radioactive nuclides separation from spent fuel.

  9. Commercial Nuclear Reprocessing in the United States

    SciTech Connect (OSTI)

    Sherrill, Charles Leland; Balatsky, Galya Ivanovna

    2015-09-09

    The short presentation outline: Reprocessing Overview; Events leading up to Carter’s Policy; Results of the decision; Policy since Nuclear Nonproliferation Act. Conclusions reached: Reprocessing ban has become an easy and visible fix to the public concern about proliferation, but has not completely stopped proliferation; and, Reprocessing needs to become detached from political considerations, so technical research can continue, regardless of the policy decisions we decide to take.

  10. Mesoscale to plant-scale models of nuclear waste reprocessing.

    SciTech Connect (OSTI)

    Noble, David Frederick; O'Hern, Timothy John; Moffat, Harry K.; Nemer, Martin B.; Domino, Stefan Paul; Rao, Rekha Ranjana; Cipiti, Benjamin B.; Brotherton, Christopher M.; Jove-Colon, Carlos F.; Pawlowski, Roger Patrick

    2010-09-01

    Imported oil exacerabates our trade deficit and funds anti-American regimes. Nuclear Energy (NE) is a demonstrated technology with high efficiency. NE's two biggest political detriments are possible accidents and nuclear waste disposal. For NE policy, proliferation is the biggest obstacle. Nuclear waste can be reduced through reprocessing, where fuel rods are separated into various streams, some of which can be reused in reactors. Current process developed in the 1950s is dirty and expensive, U/Pu separation is the most critical. Fuel rods are sheared and dissolved in acid to extract fissile material in a centrifugal contactor. Plants have many contacts in series with other separations. We have taken a science and simulation-based approach to develop a modern reprocessing plant. Models of reprocessing plants are needed to support nuclear materials accountancy, nonproliferation, plant design, and plant scale-up.

  11. DEVELOPMENT OF GLASS COMPOSITIONS TO IMMOBILIZE ALKALI, ALKALINE EARTH, LANTHANIDE AND TRANSITION METAL FISSION PRODUCTS FROM NUCLEAR FUEL REPROCESSING

    SciTech Connect (OSTI)

    Marra, J.; Billings, A.

    2009-06-24

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.

  12. DEVELOPMENT OF GLASS COMPOSITIONS TO IMMOBILIZE ALKALI, ALKALINE EARTH, LANTHANIDE AND TRANSITION METAL FISSION PRODUCTS FROM NUCLEAR FUEL REPROCESSING

    SciTech Connect (OSTI)

    Marra, James C.; Billings, Amanda Y.; Crum, Jarrod V.; Ryan, Joseph V.; Vienna, John D.

    2010-02-26

    The Advanced Fuel Cycle Initiative (AFCI) waste management strategy revolves around specific treatment of individual or groups of separated waste streams. A goal for the separations processes is to efficiently manage the waste to be dispositioned as high level radioactive waste. The Advanced Fuel Cycle Initiative (AFCI) baseline technology for immobilization of the lanthanide (Ln) and transition metal fission product (TM) wastes is vitrification into a borosilicate glass. A current interest is to evaluate the feasibility of vitrifying combined waste streams to most cost effectively immobilize the wastes resulting from aqueous fuel reprocessing. Studies showed that high waste loadings are achievable for the Ln only (Option 1) stream. Waste loadings in excess of 60 wt % (on a calcined oxide basis) were demonstrated via a lanthanide borosilicate (LaBS) glass. The resulting glasses had excellent relative durability as determined by the Product Consistency Test (PCT). For a combined Ln and TM waste stream glass (Option 2), noble metal solubility was found to limit waste loading. However, the measured PCT normalized elemental releases for this glass were at least an order of magnitude below that of Environmental Assessment (EA) glass. Current efforts to evaluate the feasibility of vitrifying combined Ln, TM, alkali (Cs is the primary radionuclide of concern) and alkaline earth (Sr is the primary radionuclide of concern) wastes (Option 3) have shown that these approaches are feasible. However, waste loading limitations with respect to heat load (Cs/Sr loading), molybdenum solubility and/or noble metal solubility will likely be realized and must be considered in determining the cost effectiveness of these approaches.

  13. MONITORING SPENT NUCLEAR FUEL REPROCESSING CONDITIONS NON-DESTRUCTIVELY AND IN NEAR-REAL-TIME USING THE MULTI-ISOTOPE PROCESS (MIP) MONITOR

    SciTech Connect (OSTI)

    Orton, Christopher R.; Fraga, Carlos G.; Douglas, Matthew; Christensen, Richard; Schwantes, Jon M.

    2010-05-07

    Researchers from Pacific Northwest National Laboratory and The Ohio State University are working to develop a system for monitoring spent nuclear fuel reprocessing facilities on-line, nondestructively, and in near-real-time. This method, known as the Multi-Isotope Process (MIP) Monitor, is based upon the measurement of distribution patterns of a suite of indicator (radioactive) isotopes present within product and waste streams of a nuclear reprocessing facility. Signatures from these indicator isotopes are monitored on-line by gamma spectrometry and compared, in near-real-time, to patterns representing "normal" process conditions using multivariate pattern recognition software. By targeting gamma-emitting indicator isotopes, the MIP Monitor approach is compatible with the use of small, portable, high-resolution gamma detectors that may be easily deployed throughout an existing facility. In addition, utilization of a suite of radio-elements, including ones with multiple oxidation states, increases the likelihood that attempts to divert material via process manipulation would be detected. Proof-of-principle modeling exercises simulating changes in acid strength have been completed and the results are promising. Laboratory testing is currently under way and significant results are available. Recent experimental results, along with an overview of the method are presented.

  14. THE MULTI-ISOTOPE PROCESS (MIP) MONITOR: A NEAR-REAL-TIME, NON-DESTRUCTIVE, INDICATOR OF SPENT NUCLEAR FUEL REPROCESSING CONDITIONS

    SciTech Connect (OSTI)

    Schwantes, Jon M.; Orton, Christopher R.; Fraga, Carlos G.; Douglas, Matthew; Christensen, Richard

    2010-05-07

    Researchers from Pacific Northwest National Laboratory and The Ohio State University are working to develop a system for monitoring spent nuclear fuel reprocessing facilities on-line, non-destructively, and in near-real-time. This method, known as the Multi-Isotope Process (MIP) Monitor, is based upon the measurement of distribution patterns of a suite of indicator (radioactive) isotopes present within product and waste streams of a nuclear reprocessing facility. Signatures from these indicator isotopes are monitored on-line by gamma spectrometry and compared, in near-real-time, to patterns representing "normal" process conditions using multivariate pattern recognition software. By targeting gamma-emitting indicator isotopes, the MIP Monitor approach is compatible with the use of small, portable, high-resolution gamma detectors that may be easily deployed throughout an existing facility. In addition, utilization of a suite of radio-elements, including ones with multiple oxidation states, increases the likelihood that attempts to divert material via process manipulation would be detected. Proof-of-principle modeling exercises simulating changes in acid strength have been completed and the results are promising. Laboratory validation is currently under way and significant results are available. The latest experimental results, along with an overview of the method will be presented.

  15. Proof of Concept Simulations of the Multi-Isotope Process Monitor: An Online, Nondestructive, Near-Real-Time Safeguards Monitor for Nuclear Fuel Reprocessing Facilities

    SciTech Connect (OSTI)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard; Schwantes, Jon M.

    2011-02-11

    The International Atomic Energy Agency (IAEA) will require the development of advanced technologies to effectively safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of nondestructive, near-real-time, autonomous process monitoring. This paper describes recent results from model simulations designed to test the Multi-Isotope Process (MIP) monitor, a novel approach to safeguarding reprocessing plants. The MIP monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in near-real-time. Three computer models including ORIGEN-ARP, AMUSE, and SYNTH were used in series to predict spent nuclear fuel composition, estimate element partitioning during separation, and simulate spectra from product and raffinate streams using a variety of gamma detectors, respectively. Simulations were generated for fuel with various irradiation histories and under a variety of plant operating conditions. Principal component analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup, and cooling time. Hierarchical cluster analysis (HCA) and partial least squares (PLS) were also used in the analysis. The MIP monitor was found to be sensitive to induced variations of several operating parameters including distinguishing ±2.5% variation from normal process acid concentrations. The ability of PLS to predict burnup levels from simulated spectra was also demonstrated to be within 3.5% of measured values.

  16. Proof of Concept Experiments of the Multi-Isotope Process Monitor: An Online, Nondestructive, Near Real-Time Monitor for Spent Nuclear Fuel Reprocessing Facilities

    SciTech Connect (OSTI)

    Orton, Christopher R.; Fraga, Carlos G.; Christensen, Richard; Schwantes, Jon M.

    2012-04-21

    Operators, national regulatory agencies and the IAEA will require the development of advanced technologies to efficiently control and safeguard nuclear material at increasingly large-scale nuclear recycling facilities. Ideally, the envisioned technologies would be capable of non-destructive, near-real-time (NRT), autonomous process monitoring. This paper describes results from proof-of-principle experiments designed to test the Multi-Isotope Process (MIP) Monitor, a novel approach to safeguarding reprocessing facilities. The MIP Monitor combines the detection of intrinsic gamma ray signatures emitted from process solutions with multivariate analysis to detect off-normal conditions in process streams nondestructively and in NRT. Commercial spent nuclear fuel of various irradiation histories was dissolved and separated using a PUREX-based batch solvent extraction. Extractions were performed at various nitric acid concentrations to mimic both normal and off-normal industrial plant operating conditions. Principal Component Analysis (PCA) was applied to the simulated gamma spectra to investigate pattern variations as a function of acid concentration, burnup and cooling time. Partial Least Squares (PLS) regression was applied to attempt to quantify both the acid concentration and burnup of the dissolved spent fuel during the initial separation stage of recycle. The MIP Monitor demonstrated sensitivity to induced variations of acid concentration, including the distinction of {+-} 1.3 M variation from normal process conditions by way of PCA. Acid concentration was predicted using measurements from the organic extract and PLS resulting in predictions with <0.7 M relative error. Quantification of burnup levels from dissolved fuel spectra using PLS was demonstrated to be within 2.5% of previously measured values.

  17. FY 2007 LDRD Director's R&D Progress SummaryProposal Title: Developing a Science Base for Fuel Reprocessing Separations in the Global Nuclear Energy Program

    SciTech Connect (OSTI)

    de Almeida, Valmor F; Tsouris, Costas; Birdwell Jr, Joseph F; D'Azevedo, Ed F; Jubin, Robert Thomas; DePaoli, David W; Moyer, Bruce A

    2011-01-01

    This work is aimed at developing an experimentally validated computational capability for understanding the complex processes governing the performance of solvent extraction devices used for separations in nuclear fuel reprocessing. These applications pose a grand challenge due to the combination of complicating factors in a three-dimensional, turbulent, reactive, multicomponent, multiphase/interface fluid flow system. The currently limited process simulation and scale-up capabilities provides uncertainty in the ability to select and design the separations technology for the demonstration plan of the Global Nuclear Energy Partnership (GNEP) program. We anticipate the development of science-based models for technology development and design. This project will position ORNL to address the emerging opportunity by creating an expandable process model validated experimentally. This project has three major thrusts, namely, a prototype experimental station, a continuum modeling and simulation effort, and molecular modeling and kinetics support. Excellent progress has been made in corresponding activities in this first year in: (1) defining, assembling, and operating a relevant prototype system for model validation; (2) establishing a mathematical model for fluid flow and transport; (3) deploying sub-scale molecular modeling.

  18. RAIL ROUTING - CURRENT PRACTICES FOR SPENT NUCLEAR FUEL AND HIGH...

    Office of Environmental Management (EM)

    Two types of highly radioactive materials are spent nuclear fuel and high-level radioactive waste that resulted from reprocessing spent fuel. Transportation of these radioactive ...

  19. W.T.; Rainey, R.H. 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS;...

    Office of Scientific and Technical Information (OSTI)

    thorium fuel reprocessing experience Brooksbank, R.E.; McDuffee, W.T.; Rainey, R.H. 11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; NUCLEAR MATERIALS DIVERSION; SAFEGUARDS; SPENT FUELS;...

  20. Advanced dry head-end reprocessing of light water reactor spent...

    Office of Scientific and Technical Information (OSTI)

    reprocessing of light water reactor spent nuclear fuel Citation Details In-Document Search Title: Advanced dry head-end reprocessing of light water reactor spent nuclear fuel ...

  1. Reprocessing of research reactor fuel the Dounreay option

    SciTech Connect (OSTI)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  2. Head-end process for the reprocessing of HTGR spent fuel

    SciTech Connect (OSTI)

    Chen, J.; Wen, M.

    2013-07-01

    The reprocessing of HTGR spent fuels is in favor of the sustainable development of nuclear energy to realize the maximal use of nuclear resource and the minimum disposal of nuclear waste. The head-end of HTGR spent fuels reprocessing is different from that of the LWR spent fuels reprocessing because of the difference of spent fuel structure. The dismantling of the graphite spent fuel element and the highly effective dissolution of fuel kernel is the most difficult process in the head end of the reprocessing. Recently, some work on the head-end has been done in China. First, the electrochemical method with nitrate salt as electrolyte was studied to disintegrate the graphite matrix from HTGR fuel elements and release the coated fuel particles, to provide an option for the head-end technology of reprocessing. The results show that the graphite matrix can be effectively separated from the coated particle without any damage to the SiC layer. Secondly, the microwave-assisted heating was applied to dissolve the UO{sub 2} kernel from the crashed coated fuel particles. The ceramic UO{sub 2} as the solute has a good ability to absorb the microwave energy. The results of UO{sub 2} kernel dissolution from crushed coated particles by microwave heating show that the total dissolution percentage of UO{sub 2} is more than 99.99% after 3 times cross-flow dissolution with the following parameters: 8 mol/L HNO{sub 3}, temperature 100 Celsius degrees, initial ratio of solid to liquid 1.2 g/ml. (authors)

  3. Analysis of nuclear proliferation resistance reprocessing and recycling technologies

    SciTech Connect (OSTI)

    Patricia Paviet-Hartmann; Gary Cerefice; Marcela Stacey; Steven Bakhtiar

    2011-05-01

    The PUREX process has been progressively and continuously improved during the past three decades, and these improvements account for successful commercialization of reprocessing in a few countries. The renewed interest in nuclear energy and the international growth of nuclear electricity generation do not equate – and should not be equated -with increasing proliferation risks. Indeed, the nuclear renaissance presents a unique opportunity to enhance the culture of non-proliferation. With the recent revival of interest in nuclear technology, technical methods for prevention of nuclear proliferation are being revisited. Robust strategies to develop new advanced separation technologies are emerging worldwide for sustainability and advancement of nuclear energy with enhanced proliferation resistance. On the other hand, at this moment, there are no proliferation resistance advanced technologies. . Until now proliferation resistance as it applies to reprocessing has been focused on not separating a pure stream of weapons-usable plutonium. France, as an example, has proposed a variant of the PUREX process, the COEX TM process, which does not result on a pure plutonium product stream. A further step is to implement a process based on group extraction of actinides and fission products associated with a homogeneous recycling strategy (UNEX process in the US, GANEX process in France). Such scheme will most likely not be deployable on an industrial scale before 2030 or so because it requires intensive R&D and robust flowsheets. Finally, future generation recycling schemes will handle the used nuclear fuel in fast neutron reactors. This means that the plutonium throughput of the recycling process may increase. The need is obvious for advanced aqueous recycling technologies that are intrinsically more proliferation resistant than the commercial PUREX process. In this paper, we review the actual PUREX process along with the advanced recycling technologies that will enhance

  4. A REVIEW OF THORIUM FUEL REPROCESSING EX

    Office of Scientific and Technical Information (OSTI)

    ... Other reactor programs involving thorium fuel include the Peach Bottom reactor (which was ... these reactors, even though the Peach Bottom spent fuel was sent to Italy for a ...

  5. Models Help Pinpoint Material for Better Nuclear Fuel Recycling

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Researchers are investigating a new material that might help in nuclear fuel recycling and waste reduction by capturing certain gases released during reprocessing. Conventional ...

  6. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    SciTech Connect (OSTI)

    Mcwilliams, A. J.

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  7. Historic American Engineering Record, Idaho National Laboratory, Idaho Chemical Processing Plant, Fuel Reprocessing Complex

    SciTech Connect (OSTI)

    Susan Stacy; Julie Braun

    2006-12-01

    Just as automobiles need fuel to operate, so do nuclear reactors. When fossil fuels such as gasoline are burned to power an automobile, they are consumed immediately and nearly completely in the process. When the fuel is gone, energy production stops. Nuclear reactors are incapable of achieving this near complete burn-up because as the fuel (uranium) that powers them is burned through the process of nuclear fission, a variety of other elements are also created and become intimately associated with the uranium. Because they absorb neutrons, which energize the fission process, these accumulating fission products eventually poison the fuel by stopping the production of energy from it. The fission products may also damage the structural integrity of the fuel elements. Even though the uranium fuel is still present, sometimes in significant quantities, it is unburnable and will not power a reactor unless it is separated from the neutron-absorbing fission products by a method called fuel reprocessing. Construction of the Fuel Reprocessing Complex at the Chem Plant started in 1950 with the Bechtel Corporation serving as construction contractor and American Cyanamid Company as operating contractor. Although the Foster Wheeler Corporation assumed responsibility for the detailed working design of the overall plant, scientists at Oak Ridge designed all of the equipment that would be employed in the uranium separations process. After three years of construction activity and extensive testing, the plant was ready to handle its first load of irradiated fuel.

  8. Spent Nuclear Fuel

    U.S. Energy Information Administration (EIA) Indexed Site

    Nuclear & Uranium Glossary FAQS Overview Data Status of U.S. nuclear outages (interactive) Nuclear power plants Uranium & nuclear fuel Spent nuclear fuel All nuclear data ...

  9. Method for photochemical reduction of uranyl nitrate by tri-N-butyl phosphate and application of this method to nuclear fuel reprocessing

    DOE Patents [OSTI]

    De Poorter, Gerald L.; Rofer-De Poorter, Cheryl K.

    1978-01-01

    Uranyl ion in solution in tri-n-butyl phosphate is readily photochemically reduced to U(IV). The product U(IV) may effectively be used in the Purex process for treating spent nuclear fuels to reduce Pu(IV) to Pu(III). The Pu(III) is readily separated from uranium in solution in the tri-n-butyl phosphate by an aqueous strip.

  10. Electrolysis cell for reprocessing plutonium reactor fuel

    DOE Patents [OSTI]

    Miller, W.E.; Steindler, M.J.; Burris, L.

    1985-01-04

    An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals is claimed. The cell includes a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket. The anode basket is extendable into the lower pool to dissolve at least some metallic contaminants; the anode basket contains the spent fuel acting as a second anode when in the electrolyte.

  11. Electrolysis cell for reprocessing plutonium reactor fuel

    DOE Patents [OSTI]

    Miller, William E.; Steindler, Martin J.; Burris, Leslie

    1986-01-01

    An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals, the cell including a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket and the anode basket being extendable into the lower pool to dissolve at least some metallic contaminants, the anode basket containing the spent fuel acting as a second anode when in the electrolyte.

  12. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect (OSTI)

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  13. Spent Fuel Reprocessing: More Value for Money Spent in a Geological Repository?

    SciTech Connect (OSTI)

    Kaplan, P.; Vinoche, R.; Devezeaux, J-G.; Bailly, F.

    2003-02-25

    Today, each utility or country operating nuclear power plants can select between two long-term spent fuel management policies: either, spent fuel is considered as waste to dispose of through direct disposal or, spent fuel is considered a resource of valuable material through reprocessing-recycling. Reading and listening to what is said in the nuclear community, we understand that most people consider that the choice of policy is, actually, a choice among two technical paths to handle spent fuel: direct disposal versus reprocessing. This very simple situation has been recently challenged by analysis coming from countries where both policies are on survey. For example, ONDRAF of Belgium published an interesting study showing that, economically speaking for final disposal, it is worth treating spent fuel rather than dispose of it as a whole, even if there is no possibility to recycle the valuable part of it. So, the question is raised: is there such a one-to-one link between long term spent fuel management political option and industrial option? The purpose of the presentation is to discuss the potential advantages and drawbacks of spent fuel treatment as an implementation of the policy that considers spent fuel as waste to dispose of. Based on technical considerations and industrial experience, we will study qualitatively, and quantitatively when possible, the different answers proposed by treatment to the main concerns of spent-fuel-as-a-whole geological disposal.

  14. An Assessment of Spent Fuel Reprocessing for Actinide Destruction and Resource Sustainability.

    SciTech Connect (OSTI)

    Cipiti, Benjamin B.; Smith, James D.

    2008-09-01

    The reprocessing and recycling of spent nuclear fuel can benefit the nuclear fuel cycle by destroying actinides or extending fissionable resources if uranium supplies become limited. The purpose of this study was to assess reprocessing and recycling in both fast and thermal reactors to determine the effectiveness for actinide destruction and resource utilization. Fast reactor recycling will reduce both the mass and heat load of actinides by a factor of 2, but only after 3 recycles and many decades. Thermal reactor recycling is similarly effective for reducing actinide mass, but the heat load will increase by a factor of 2. Economically recoverable reserves of uranium are estimated to sustain the current global fleet for the next 100 years, and undiscovered reserves and lower quality ores are estimated to contain twice the amount of economically recoverable reserves--which delays the concern of resource utilization for many decades. Economic analysis reveals that reprocessed plutonium will become competitive only when uranium prices rise to about %24360 per kg. Alternative uranium sources are estimated to be competitive well below that price. Decisions regarding the development of a near term commercial-scale reprocessing fuel cycle must partially take into account the effectiveness of reactors for actnides destruction and the time scale for when uranium supplies may become limited. Long-term research and development is recommended in order to make more dramatic improvements in actinide destruction and cost reductions for advanced fuel cycle technologies.The original scope of this work was to optimize an advanced fuel cycle using a tool that couples a reprocessing plant simulation model with a depletion analysis code. Due to funding and time constraints of the late start LDRD process and a lack of support for follow-on work, the project focused instead on a comparison of different reprocessing and recycling options. This optimization study led to new insight into

  15. A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

    SciTech Connect (OSTI)

    Parker, Frank L.

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long

  16. Consolidated Fuel Reprocessing Program progress report, 1 October-31 December 1979. [HEF

    SciTech Connect (OSTI)

    Unger, W.E.

    1980-05-01

    Progress is reported in four areas: process research and development, engineering research, engineering systems, technical support, and HTGR fuel reprocessing. (DLC)

  17. Plasma Mass Filters For Nuclear Waste Reprocessing

    SciTech Connect (OSTI)

    Abraham J. Fetterman and Nathaniel J. Fisch

    2011-05-26

    Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

  18. Plasma Mass Filters For Nuclear Waste Reprocessing

    SciTech Connect (OSTI)

    Abraham J. Fetterman and Nathaniel J. Fisch

    2011-05-25

    Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

  19. Nuclear Fuels: Promise and Limitations

    SciTech Connect (OSTI)

    Harold F. McFarlane

    2012-03-01

    From 1950 through 1980, scientists, engineers and national leaders confidently predicted an early twenty-first century where fast breeder reactors and commercial nuclear fuel reprocessing were commonplace. Such a scenario seemed necessary for a world with the more than 1000 GWe of nuclear energy needed to meet such an ever-increasing thirst for energy. Thirty years later uranium reserves are increasing on pace with consumption, the growth of nuclear power has been slowed, commercial breeder reactors have yet to enter the marketplace, and less than a handful of commercial reprocessing plants operate. As Nobel Laureate Niels Bohr famously said, Prediction is very difficult, especially if its about the future. The programme for IChemEs 2012 conference on the nuclear fuel cycle features a graphic of an idealized nuclear fuel cycle that symbolizes the quest for a closed nuclear fuel cycle featuring careful husbanding of precious resources while minimizing the waste footprint. Progress toward achieving this ideal has been disrupted by technology innovations in the mining and petrochemical industries, as well as within the nuclear industry.

  20. Spectroscopic Online Monitoring for Process Control and Safeguarding of Radiochemical Fuel Reprocessing Streams - 13553

    SciTech Connect (OSTI)

    Bryan, S.A.; Levitskaia, T.G.; Casella, Amanda; Peterson, James

    2013-07-01

    There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the used nuclear fuel. Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. This paper discusses application of absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. Both techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Static spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical safeguards and process control. (authors)

  1. Spectroscopic Online Monitoring for Process Control and Safeguarding of Radiochemical Fuel Reprocessing Streams

    SciTech Connect (OSTI)

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Casella, Amanda J.; Peterson, James M.

    2013-02-24

    There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle. The long term successful use of nuclear power is critically dependent upon adequate and safe processing and disposition of the spent nuclear fuel. Liquid-liquid extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments such as high radiation fields and aggressive chemical matrices. In addition, the ability for continuous online monitoring allows for numerous benefits. This paper reviews application of the absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed the potential of Raman and spectrophotometric techniques for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in solutions relevant to spent fuel reprocessing. Both techniques demonstrated robust performance in the repetitive batch measurements of each analyte in a wide concentration range using simulant and commercial dissolved spent fuel solutions. Static spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated using on-line centrifugal contactor extraction tests. Satisfactory prediction of the analytes concentrations in these preliminary experiments warrants further development of the spectroscopy-based methods for radiochemical safeguards and process control.

  2. Analysis of the Reuse of Uranium Recovered from the Reprocessing of Commercial LWR Spent Fuel

    SciTech Connect (OSTI)

    DelCul, Guillermo Daniel; Trowbridge, Lee D; Renier, John-Paul; Ellis, Ronald James; Williams, Kent Alan; Spencer, Barry B; Collins, Emory D

    2009-02-01

    This report provides an analysis of the factors involved in the reuse of uranium recovered from commercial light-water-reactor (LWR) spent fuels (1) by reenrichment and recycling as fuel to LWRs and/or (2) by recycling directly as fuel to heavy-water-reactors (HWRs), such as the CANDU (registered trade name for the Canadian Deuterium Uranium Reactor). Reuse is an attractive alternative to the current Advanced Fuel Cycle Initiative (AFCI) Global Nuclear Energy Partnership (GNEP) baseline plan, which stores the reprocessed uranium (RU) for an uncertain future or attempts to dispose of it as 'greater-than-Class C' waste. Considering that the open fuel cycle currently deployed in the United States already creates a huge excess quantity of depleted uranium, the closed fuel cycle should enable the recycle of the major components of spent fuel, such as the uranium and the hazardous, long-lived transuranic (TRU) actinides, as well as the managed disposal of fission product wastes. Compared with the GNEP baseline scenario, the reuse of RU in the uranium fuel cycle has a number of potential advantages: (1) avoidance of purchase costs of 11-20% of the natural uranium feed; (2) avoidance of disposal costs for a large majority of the volume of spent fuel that is reprocessed; (3) avoidance of disposal costs for a portion of the depleted uranium from the enrichment step; (4) depending on the {sup 235}U assay of the RU, possible avoidance of separative work costs; and (5) a significant increase in the production of {sup 238}Pu due to the presence of {sup 236}U, which benefits somewhat the transmutation value of the plutonium and also provides some proliferation resistance.

  3. New Material has Potential to Cut Costs and Make Nuclear Fuel...

    Broader source: Energy.gov (indexed) [DOE]

    RICHLAND, Wash. - Researchers are investigating a new material that might help in nuclear fuel recycling and waste reduction by capturing certain gases released during reprocessing...

  4. Advanced nuclear fuel

    SciTech Connect (OSTI)

    Terrani, Kurt

    2014-07-14

    Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

  5. Advanced nuclear fuel

    ScienceCinema (OSTI)

    Terrani, Kurt

    2014-07-15

    Kurt Terrani uses his expertise in materials science to develop safer fuel for nuclear power plants.

  6. Next-generation purex flowsheets with acetohydroxamic acid as complexant for FBR and thermal-fuel reprocessing

    SciTech Connect (OSTI)

    Kumar, Shekhar; Koganti, S.B.

    2008-07-01

    Acetohydroxamic acid (AHA) is a novel complexant for recycle of nuclear-fuel materials. It can be used in ordinary centrifugal extractors, eliminating the need for electro-redox equipment or complex maintenance requirements in a remotely maintained hot cell. In this work, the effect of AHA on Pu(IV) distribution ratios in 30% TBP system was quantified, modeled, and integrated in SIMPSEX code. Two sets of batch experiments involving macro Pu concentrations (conducted at IGCAR) and one high-Pu flowsheet (literature) were simulated for AHA based U-Pu separation. Based on the simulation and validation results, AHA based next-generation reprocessing flowsheets are proposed for co-processing based FBR and thermal-fuel reprocessing as well as evaporator-less macro-level Pu concentration process required for MOX fuel fabrication. Utilization of AHA results in significant simplification in plant design and simpler technology implementations with significant cost savings. (authors)

  7. Advantages of co-located spent fuel reprocessing, repository and underground reactor facilities

    SciTech Connect (OSTI)

    Mahar, James M.; Kunze, Jay F.; Wes Myers, Carl; Loveland, Ryan

    2007-07-01

    The purpose of this work is to extend the discussion of potential advantages of the underground nuclear park (UNP) concept by making specific concept design and cost estimate comparisons for both present Generation III types of reactors and for some of the modular Gen IV or the GNEP modular concept. For the present Gen III types, we propose co-locating reprocessing and (re)fabrication facilities along with disposal facilities in the underground park. The goal is to determine the site costs and facility construction costs of such a complex which incorporates the advantages of a closed fuel cycle, nuclear waste repository, and ultimate decommissioning activities all within the UNP. Modular power generation units are also well-suited for placement underground and have the added advantage of construction using current and future tunnel boring machine technology. (authors)

  8. Consolidated fuel reprocessing program. Progress report, July 1-September 30, 1981

    SciTech Connect (OSTI)

    None

    1981-12-01

    Technical progress is reported in overview fashion in the following areas: process development, laboratory R and D, engineering research, engineering systems, integrated equipment test facility (IET) operations, and HTGR fuel reprocessing. (DLC)

  9. Spent Nuclear Fuel

    Gasoline and Diesel Fuel Update (EIA)

    Spent Nuclear Fuel Release date: December 7, 2015 Next release date: Late 2018 Spent nuclear fuel data are collected by the U.S. Energy Information Administration (EIA) for the Department of Energy's Office of Standard Contract Management (Office of the General Counsel) on the Form GC-859, "Nuclear Fuel Data Survey." The data include detailed characteristics of spent nuclear fuel discharged from commercial U.S. nuclear power plants and currently stored at commercial sites in the United

  10. Contaminants of the bismuth phosphate process as signifiers of nuclear reprocessing history.

    SciTech Connect (OSTI)

    Schwantes, Jon M.; Sweet, Lucas E.

    2012-10-01

    Reagents used in spent nuclear fuel recycling impart unique contaminant patterns into the product stream of the process. Efforts are underway at Pacific Northwest National Laboratory to characterize and understand the relationship between these patterns and the process that created them. A main challenge to this effort, recycling processes that were employed at the Hanford site from 1944-1989 have been retired for decades. This precludes direct measurements of the contaminant patterns that propagate within product streams of these facilities. In the absence of any operating recycling facilities at Hanford, we have taken a multipronged approach to cataloging contaminants of U.S. reprocessing activities using: (1) historical records summarizing contaminants within the final Pu metal button product of these facilities; (2) samples of opportunity that represent intermediate products of these processes; and (3) lab-scale experiments and model simulations designed to replicate contaminant patterns at each stage of nuclear fuel reprocessing. This report provides a summary of the progress and results from Fiscal Year (April 1, 2010-September 30) 2011.

  11. Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

    SciTech Connect (OSTI)

    Trtilek, Radek; Podlaha, Josef [UJV Rez, a. s., Hlavni 130, 25068 Husinec-Rez (Czech Republic)] [UJV Rez, a. s., Hlavni 130, 25068 Husinec-Rez (Czech Republic)

    2013-07-01

    After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the second shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)

  12. Process Description and Operating History for the CPP-601/-640/-627 Fuel Reprocessing Complex at the Idaho National Engineering and Environmental Laboratory

    SciTech Connect (OSTI)

    E. P. Wagner

    1999-06-01

    The Fuel Reprocessing Complex (FRC) at the Idaho Nuclear Technology and Engineering Center at the Idaho National Engineering and Environmental Laboratory was used for reprocessing spent nuclear fuel from the early 1950's until 1992. The reprocessing facilities are now scheduled to be deactivated. As part of the deactivation process, three Resource Conservation and Recovery Act (RCRA) interim status units located in the complex must be closed. This document gathers the historical information necessary to provide a rational basis for the preparation of a comprehensive closure plan. Included are descriptions of process operations and the operating history of the FRC. A set of detailed tables record the service history and present status of the process vessels and transfer lines.

  13. Vented nuclear fuel element

    DOE Patents [OSTI]

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

    A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

  14. Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

    SciTech Connect (OSTI)

    Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.; Baryshnikov, M.V.; Kryukov, O.V.; Khaperskaya, A.V.

    2013-07-01

    REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

  15. Materials management in an internationally safeguarded fuels reprocessing plant

    SciTech Connect (OSTI)

    Hakkila, E.A.; Baker, A.L.; Cobb, D.D.

    1980-04-01

    The following appendices are included: aqueous reprocessing and conversion technology, reference facilities, process design and operating features relevant to materials accounting, operator's safeguards system structure, design principles of dynamic materials accounting systems, modeling and simulation approach, optimization of measurement control, aspects of international verification problem, security and reliability of materials measurement and accounting system, estimation of in-process inventory in solvent-extraction contactors, conventional measurement techniques, near-real-time measurement techniques, isotopic correlation techniques, instrumentation available to IAEA inspectors, and integration of materials accounting and containment and surveillance. (DLC)

  16. Nuclear fuel composition

    DOE Patents [OSTI]

    Feild, Jr., Alexander L.

    1980-02-19

    1. A high temperature graphite-uranium base nuclear fuel composition containing from about 1 to about 5 five weight percent rhenium metal.

  17. The Nuclear Energy Advanced Modeling and Simulation Safeguards and Separations Reprocessing Plant Toolkit

    SciTech Connect (OSTI)

    McCaskey, Alex; Billings, Jay Jay; de Almeida, Valmor F

    2011-08-01

    This report details the progress made in the development of the Reprocessing Plant Toolkit (RPTk) for the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. RPTk is an ongoing development effort intended to provide users with an extensible, integrated, and scalable software framework for the modeling and simulation of spent nuclear fuel reprocessing plants by enabling the insertion and coupling of user-developed physicochemical modules of variable fidelity. The NEAMS Safeguards and Separations IPSC (SafeSeps) and the Enabling Computational Technologies (ECT) supporting program element have partnered to release an initial version of the RPTk with a focus on software usability and utility. RPTk implements a data flow architecture that is the source of the system's extensibility and scalability. Data flows through physicochemical modules sequentially, with each module importing data, evolving it, and exporting the updated data to the next downstream module. This is accomplished through various architectural abstractions designed to give RPTk true plug-and-play capabilities. A simple application of this architecture, as well as RPTk data flow and evolution, is demonstrated in Section 6 with an application consisting of two coupled physicochemical modules. The remaining sections describe this ongoing work in full, from system vision and design inception to full implementation. Section 3 describes the relevant software development processes used by the RPTk development team. These processes allow the team to manage system complexity and ensure stakeholder satisfaction. This section also details the work done on the RPTk ``black box'' and ``white box'' models, with a special focus on the separation of concerns between the RPTk user interface and application runtime. Section 4 and 5 discuss that application runtime component in more detail, and describe the dependencies, behavior, and rigorous testing of its constituent components.

  18. Monitoring, Controlling and Safeguarding Radiochemical Streams at Spent Fuel Reprocessing Facilities, Part 1: Optical Spectroscopic Methods

    SciTech Connect (OSTI)

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Schwantes, Jon M.; Orton, Christopher R.; Peterson, James M.; Casella, Amanda J.

    2012-02-07

    Abstract: The International Atomic Energy Agency (IAEA) has established international safeguards standards for fissionable material at spent fuel reprocessing plants to ensure that significant quantities of weapons-useable nuclear material are not diverted from these facilities. For large throughput nuclear facilities, it is difficult to satisfy the IAEA safeguards accountancy goal for detection of abrupt diversion. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resource-intensive destructive assay (DA). Leveraging new on-line non-destructive assay (NDA) process monitoring techniques in conjunction with the traditional and highly precise DA methods may provide an additional measure to nuclear material accountancy which would potentially result in a more timely, cost-effective and resource efficient means for safeguards verification at such facilities. By monitoring process control measurements (e.g. flowrates, temperatures, or concentrations of reagents, products or wastes), abnormal plant operations can be detected. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies based upon gamma-ray and optical spectroscopic measurements to potentially reduce the time and resource burden associated with current techniques. The Multi-Isotope Process (MIP) Monitor uses gamma spectroscopy and multivariate analysis to identify off-normal conditions in process streams. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major stable flowsheet reagents using UV-Vis, Near IR and Raman spectroscopy. Multi-variate analysis is also applied to the optical measurements in order to quantify concentrations of analytes of interest within a complex array of radiochemical streams. This paper will provide an overview of these methods and reports on-going efforts

  19. Advanced dry head-end reprocessing of light water reactor spent...

    Office of Scientific and Technical Information (OSTI)

    Patent: Advanced dry head-end reprocessing of light water reactor spent nuclear fuel Citation Details In-Document Search Title: Advanced dry head-end reprocessing of light water ...

  20. Financing Strategies for Nuclear Fuel Cycle Facility

    SciTech Connect (OSTI)

    David Shropshire; Sharon Chandler

    2005-12-01

    To help meet our nation’s energy needs, reprocessing of spent nuclear fuel is being considered more and more as a necessary step in a future nuclear fuel cycle, but incorporating this step into the fuel cycle will require considerable investment. This report presents an evaluation of financing scenarios for reprocessing facilities integrated into the nuclear fuel cycle. A range of options, from fully government owned to fully private owned, was evaluated using a DPL (Dynamic Programming Language) 6.0 model, which can systematically optimize outcomes based on user-defined criteria (e.g., lowest life-cycle cost, lowest unit cost). Though all business decisions follow similar logic with regard to financing, reprocessing facilities are an exception due to the range of financing options available. The evaluation concludes that lowest unit costs and lifetime costs follow a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. Other financing arrangements, however, including regulated utility ownership and a hybrid ownership scheme, led to acceptable costs, below the Nuclear Energy Agency published estimates. Overwhelmingly, uncertainty in annual capacity led to the greatest fluctuations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; the annual operating costs dominate the government case. It is concluded that to finance the construction and operation of such a facility without government ownership could be feasible with measures taken to mitigate risk, and that factors besides unit costs should be considered (e.g., legal issues, social effects, proliferation concerns) before making a decision on financing strategy.

  1. TEPP- Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    This scenario provides the planning instructions, guidance, and evaluation forms necessary to conduct an exercise involving a highway shipment of spent nuclear fuel.  This exercise manual is one in...

  2. Nuclear fuel element

    DOE Patents [OSTI]

    Zocher, Roy W.

    1991-01-01

    A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

  3. Closing nuclear fuel cycle with fast reactors: problems and prospects

    SciTech Connect (OSTI)

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V.

    2013-07-01

    The closed nuclear fuel cycle (CNFC) with fast reactors (FR) is the most promising way of nuclear energetics development because it prevents spent nuclear fuel (SNF) accumulation and minimizes radwaste volume due to minor actinides (MA) transmutation. CNFC with FR requires the elaboration of safety, environmentally acceptable and economically effective methods of treatment of SNF with high burn-up and low cooling time. The up-to-date industrially implemented SNF reprocessing technologies based on hydrometallurgical methods are not suitable for the reprocessing of SNF with high burn-up and low cooling time. The alternative dry methods (such as electrorefining in molten salts or fluoride technologies) applicable for such SNF reprocessing have not found implementation at industrial scale. So the cost of SNF reprocessing by means of dry technologies can hardly be estimated. Another problem of dry technologies is the recovery of fissionable materials pure enough for dense fuel fabrication. A combination of technical solutions performed with hydrometallurgical and dry technologies (pyro-technology) is proposed and it appears to be a promising way for the elaboration of economically, ecologically and socially accepted technology of FR SNF management. This paper deals with discussion of main principle of dry and aqueous operations combination that probably would provide safety and economic efficiency of the FR SNF reprocessing. (authors)

  4. Nuclear fuel element

    DOE Patents [OSTI]

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  5. The Prospective Role of JAEA Nuclear Fuel Cycle Engineering Laboratories

    SciTech Connect (OSTI)

    Ojima, Hisao; Dojiri, Shigeru; Tanaka, Kazuhiko; Takeda, Seiichiro; Nomura, Shigeo

    2007-07-01

    JAEA Nuclear Fuel Cycle Engineering Laboratories was established in 2005 to take over the activities of the JNC Tokai Works. Many kinds of development activities have been carried out since 1959. Among these, the results on the centrifuge for U enrichment, LWR spent fuel reprocessing and MOX fuel fabrication have already provided the foundation of the fuel cycle industry in Japan. R and D on the treatment and disposal of high-level waste and FBR fuel reprocessing has also been carried out. Through such activities, radioactive material release to the environment has been appropriately controlled and all nuclear materials have been placed under IAEA safeguards. The Laboratories has sufficient experience and ability to establish the next generation closed cycle and strives to become a world-class Center Of Excellence (COE). (authors)

  6. Nuclear fuel pin scanner

    DOE Patents [OSTI]

    Bramblett, Richard L.; Preskitt, Charles A.

    1987-03-03

    Systems and methods for inspection of nuclear fuel pins to determine fiss loading and uniformity. The system includes infeed mechanisms which stockpile, identify and install nuclear fuel pins into an irradiator. The irradiator provides extended activation times using an approximately cylindrical arrangement of numerous fuel pins. The fuel pins can be arranged in a magazine which is rotated about a longitudinal axis of rotation. A source of activating radiation is positioned equidistant from the fuel pins along the longitudinal axis of rotation. The source of activating radiation is preferably oscillated along the axis to uniformly activate the fuel pins. A detector is provided downstream of the irradiator. The detector uses a plurality of detector elements arranged in an axial array. Each detector element inspects a segment of the fuel pin. The activated fuel pin being inspected in the detector is oscillated repeatedly over a distance equal to the spacing between adjacent detector elements, thereby multiplying the effective time available for detecting radiation emissions from the activated fuel pin.

  7. MOX Reprocessing at Tokai Reprocessing Plant

    SciTech Connect (OSTI)

    Taguchi, Katsuya; Nagaoka, Shinichi; Yamanaka, Atsushi; Nakamura, Yoshinobu; Omori, Eiichi; SATO, Takehiko; MIURA, Nobuyuki

    2007-07-01

    In March 2007, the first reprocessing of the 'Type B' MOX spent fuels of the Prototype Advanced Thermal Reactor FUGEN was initiated at Tokai Reprocessing Plant as a plant-scale demonstration of MOX fuel reprocessing. The operation was advanced satisfactorily and it has been confirmed that the MOX fuels as well as UO{sub 2} fuels can be reprocessed safely. Some characteristics of MOX fuels on reprocessing, such as properties of undissolved residue affecting the clarification process, are becoming visible. Reprocessing of the 'Type B' MOX fuels will be continued for several more years from now on, further investigations on solubility of fuels, characteristics of undissolved residues, progress of solvent degradation and so on will be continued. (authors)

  8. Nuclear Fuel Cycle Options Catalog

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Hydrogen Production Market Transformation Fuel Cells Predictive Simulation of Engines ... Twitter Google + Vimeo Newsletter Signup SlideShare Nuclear Fuel Cycle Options Catalog ...

  9. Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Ruz, J.; Descalle, M. A.; Alameda, J. B.; Chichester, David L.; Decker, T. A.; Fernandez-Perea, M.; Hill, R. M.; Kisner, Roger A.; Melin, Alexander M.; Patton, Bruce W.; et al

    2016-05-24

    The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. As a result, the experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in futuremore » measurement campaigns.« less

  10. Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    Dale, Deborah J.

    2014-10-28

    These slides will be presented at the training course “International Training Course on Implementing State Systems of Accounting for and Control (SSAC) of Nuclear Material for States with Small Quantity Protocols (SQP),” on November 3-7, 2014 in Santa Fe, New Mexico. The slides provide a basic overview of the Nuclear Fuel Cycle. This is a joint training course provided by NNSA and IAEA.

  11. Characterization of Nuclear Fuel using Multivariate Statistical Analysis

    SciTech Connect (OSTI)

    Robel, M; Robel, M; Robel, M; Kristo, M J; Kristo, M J

    2007-11-27

    Various combinations of reactor type and fuel composition have been characterized using principle components analysis (PCA) of the concentrations of 9 U and Pu isotopes in the 10 fuel as a function of burnup. The use of PCA allows the reduction of the 9-dimensional data (isotopic concentrations) into a 3-dimensional approximation, giving a visual representation of the changes in nuclear fuel composition with burnup. Real-world variation in the concentrations of {sup 234}U and {sup 236}U in the fresh (unirradiated) fuel was accounted for. The effects of reprocessing were also simulated. The results suggest that, 15 even after reprocessing, Pu isotopes can be used to determine both the type of reactor and the initial fuel composition with good discrimination. Finally, partial least squares discriminant analysis (PSLDA) was investigated as a substitute for PCA. Our results suggest that PLSDA is a better tool for this application where separation between known classes is most important.

  12. Nuclear fuel element

    DOE Patents [OSTI]

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1980-04-29

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

  13. Nuclear fuel element

    DOE Patents [OSTI]

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

  14. Study of the potential uses of the Barnwell Nuclear Fuel Plant (BNFP). Final report

    SciTech Connect (OSTI)

    Not Available

    1980-03-25

    The purpose of this study is to provide an evaluation of possible international and domestic uses for the Barnwell Nuclear Fuel Plant, located in South Carolina, at the conclusion of the International Nuclear Fuel Cycle Evaluation. Four generic categories of use options for the Barnwell plant have been considered: storage of spent LWR fuel; reprocessing of LWR spent fuel; safeguards development and training; and non-use. Chapters are devoted to institutional options and integrated institutional-use options.

  15. Nuclear reactor fuel element

    DOE Patents [OSTI]

    Johnson, Carl E.; Crouthamel, Carl E.

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  16. PRELIMINARY STUDY OF CERAMICS FOR IMMOBILIZATION OF ADVANCED FUEL CYCLE REPROCESSING WASTES

    SciTech Connect (OSTI)

    Fox, K.; Billings, A.; Brinkman, K.; Marra, J.

    2010-09-22

    The Savannah River National Laboratory (SRNL) developed a series of ceramic waste forms for the immobilization of Cesium/Lanthanide (CS/LN) and Cesium/Lanthanide/Transition Metal (CS/LN/TM) waste streams anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3}, CaO, and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores, zirconolite, and other minor metal titanate phases. Identification of excess Al{sub 2}O{sub 3} via X-ray Diffraction (XRD) and Scanning Electron Microscopy with Energy Dispersive Spectroscopy (SEM/EDS) in the first series of compositions led to a Phase II study, with significantly reduced Al{sub 2}O{sub 3} concentrations and increased waste loadings. Three fabrication methodologies were used, including melting and crystallizing, pressing and sintering, and Spark Plasma Sintering (SPS), with the intent of studying phase evolution under various sintering conditions. XRD and SEM/EDS results showed that the partitioning of the waste elements in the sintered materials was very similar, despite varying stoichiometry of the phases formed. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD, and had phase assemblages that were closer to the initial targets. Chemical composition measurements showed no significant issues with meeting the target compositions. However, volatilization of Cs and Mo was identified, particularly during melting, since sintering of the pressed pellets and SPS were performed at lower temperatures. Partitioning of some of the waste components was difficult to determine via XRD. SEM/EDS mapping showed that those elements, which were generally present in small concentrations, were well distributed throughout the waste forms. Initial studies of radiation damage tolerance using ion beam irradiation at Los

  17. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOE Patents [OSTI]

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

  18. Nuclear fuel electrorefiner

    DOE Patents [OSTI]

    Ahluwalia, Rajesh K.; Hua, Thanh Q.

    2004-02-10

    The present invention relates to a nuclear fuel electrorefiner having a vessel containing a molten electrolyte pool floating on top of a cadmium pool. An anodic fuel dissolution basket and a high-efficiency cathode are suspended in the molten electrolyte pool. A shroud surrounds the fuel dissolution basket and the shroud is positioned so as to separate the electrolyte pool into an isolated electrolyte pool within the shroud and a bulk electrolyte pool outside the shroud. In operation, unwanted noble-metal fission products migrate downward into the cadmium pool and form precipitates where they are removed by a filter and separator assembly. Uranium values are transported by the cadmium pool from the isolated electrolyte pool to the bulk electrolyte pool, and then pass to the high-efficiency cathode where they are electrolytically deposited thereto.

  19. Sustainability Considerations in Spent Light-water Nuclear Fuel Retrievability

    SciTech Connect (OSTI)

    Wood, Thomas W.; Rothwell, Geoffrey

    2012-01-10

    This paper examines long-term cost differences between two competing Light Water Reactor (LWR) fuels: Uranium Oxide (UOX) and Mixed Uranium Oxide-Plutonium Oxide (MOX). Since these costs are calculated on a life-cycle basis, expected savings from lower future MOX fuel prices can be used to value the option of substituting MOX for UOX, including the value of maintaining access to the used UOX fuel that could be reprocessed to make MOX. The two most influential cost drivers are the price of natural uranium and the cost of reprocessing. Significant and sustained reductions in reprocessing costs and/or sustained increases in uranium prices are required to give positive value to the retrievability of Spent Nuclear Fuel. While this option has positive economic value, it might not be exercised for 50 to 200 years. Therefore, there are many years for a program during which reprocessing technology can be researched, developed, demonstrated, and deployed. Further research is required to determine whether the cost of such a program would yield positive net present value and/or increases the sustainability of LWR energy systems.

  20. A European perspective on the US nuclear fuel cycle

    SciTech Connect (OSTI)

    Wilkinson, W.L.

    1989-01-01

    Many Europeans believe that the main problems which have impeded progress in solving the back end of the nuclear fuel cycle in the United States have been a series of ideological and political hang-ups and these, coupled with excessive bureaucracy, have made logical decision making on the back-end problems impossible. This situation has been caused by a succession of political nondecisions. Public confidence in nuclear generation was thereby undermined and, because of plentiful supplies of other energy sources, there was no urgent need to expand the nuclear program in the United States. Since uranium was cheap and fast reactors not commercially attractive, there was no economic incentive to reprocess fuel from existing reactors in the United States. The problem facing the United States is that of managing the large stocks of spent fuel which have arisen over many years. A logical way forward for the United States would appear to be as follows: build more storage for spent fuel; consider overseas reprocessing to provide plutonium; develop reprocessing technology; and develop direct disposal technology.

  1. A review of nuclear fuel cycle options for developing nations

    SciTech Connect (OSTI)

    Harrison, R.K.; Scopatz, A.M.; Ernesti, M.

    2007-07-01

    A study of several nuclear reactor and fuel cycle options for developing nations was performed. All reactor choices were considered under a GNEP framework. Two advanced alternative reactor types, a nuclear battery-type reactor and a fuel reprocessing fast reactor were examined and compared with a conventional Generation III+ LWR reactor. The burn of nuclear fuel was simulated using ORIGEN 2.2 for each reactor type and the resulting information was used to compare the options in terms of waste produced, waste quality and repository impact. The ORIGEN data was also used to evaluate the economics of the fuel cycles using unit costs, discount rates and present value functions with the material balances. The comparison of the fuel cycles and reactors developed in this work provides a basis for the evaluation of subsidy programs and cost-benefit comparisons for various reactor parameters such as repository impact and proliferation risk versus economic considerations. (authors)

  2. Advanced instrumentation for reprocessing.

    SciTech Connect (OSTI)

    Cipiti, Benjamin B.

    2005-10-01

    Recent interest in reprocessing nuclear fuel in the U.S. has led to advanced separations processes that employ continuous processing and multiple extraction steps. These advanced plants will need to be designed with state-of-the-art instrumentation for materials accountancy and control. This research examines the current and upcoming instrumentation for nuclear materials accountancy for those most suited to the reprocessing environment. Though this topic has received attention time and again in the past, new technologies and changing world conditions require a renewed look and this subject. The needs for the advanced UREX+ separations concept are first identified, and then a literature review of current and upcoming measuring techniques is presented. The report concludes with a preliminary list of recommended instruments and measurement locations.

  3. A Combined Nonfertile and UO{sub 2} PWR Fuel Assembly for Actinide...

    Office of Scientific and Technical Information (OSTI)

    the CONFU assembly exhibits negative reactivity feedback coefficients comparable in ... NUCLEAR FUELS; PWR TYPE REACTORS; REACTIVITY COEFFICIENTS; REPROCESSING; SIMULATION; ...

  4. THORP Thermal Oxide Reprocessing Plant | Open Energy Information

    Open Energy Info (EERE)

    Plant) Place: Cumbria, England, United Kingdom Zip: CA20 1PG Product: England-based nuclear fuel reprocessing plant. Coordinates: 54.63044, -2.89984 Show Map Loading map......

  5. Benefits and concerns of a closed nuclear fuel cycle

    SciTech Connect (OSTI)

    Widder, Sarah H.

    2010-11-17

    Nuclear power can play an important role in our energy future, contributing to increasing electricity demand while at the same time decreasing carbon dioxide emissions. However, the nuclear fuel cycle in the United States today is unsustainable. As stated in the 1982 Nuclear Waste Policy Act, the U.S. Department of Energy is responsible for disposing of spent nuclear fuel generated by commercial nuclear power plants operating in a “once-through” fuel cycle in the deep geologic repository located at Yucca Mountain. However, unyielding political opposition to the site has hindered the commissioning process to the extant that the current administration has recently declared the unsuitability of the Yucca Mountain site. In light of this the DOE is exploring other options, including closing the fuel cycle through recycling and reprocessing of spent nuclear fuel. The possibility of closing the fuel cycle is receiving special attention because of its ability to minimize the final high level waste (HLW) package as well as recover additional energy value from the original fuel. The technology is, however, still very controversial because of the increased cost and proliferation risk it can present. To lend perspective on the closed fuel cycle alternative, this presents the arguments for and against closing the fuel cycle with respect to sustainability, proliferation risk, commercial viability, waste management, and energy security.

  6. Swelling-resistant nuclear fuel

    DOE Patents [OSTI]

    Arsenlis, Athanasios; Satcher, Jr., Joe; Kucheyev, Sergei O.

    2011-12-27

    A nuclear fuel according to one embodiment includes an assembly of nuclear fuel particles; and continuous open channels defined between at least some of the nuclear fuel particles, wherein the channels are characterized as allowing fission gasses produced in an interior of the assembly to escape from the interior of the assembly to an exterior thereof without causing significant swelling of the assembly. Additional embodiments, including methods, are also presented.

  7. Nuclear Fuels | Department of Energy

    Energy Savers [EERE]

    A nuclear fuel assembly consists of hundreds of thousands of uranium pellets, stacked and ... scales from micro-structural level to individual pellets to entire rods and bundles. ...

  8. Nuclear Regulatory Commission's Integrated Strategy for Spent...

    Office of Environmental Management (EM)

    Growing Spent Fuel Inventory Cumulative Used Nuclear Fuel Scenarios 50,000 100,000 150,000 ... of a regulatory framework for reprocessing - Reviewing YM application ...

  9. AN ANALYTICAL FRAMEWORK FOR ASSESSING RELIABLE NUCLEAR FUEL SERVICE APPROACHES: ECONOMIC AND NON-PROLIFERATION MERITS OF NUCLEAR FUEL LEASING

    SciTech Connect (OSTI)

    Kreyling, Sean J.; Brothers, Alan J.; Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.

    2010-08-11

    The goal of international nuclear policy since the dawn of nuclear power has been the peaceful expansion of nuclear energy while controlling the spread of enrichment and reprocessing technology. Numerous initiatives undertaken in the intervening decades to develop international agreements on providing nuclear fuel supply assurances, or reliable nuclear fuel services (RNFS) attempted to control the spread of sensitive nuclear materials and technology. In order to inform the international debate and the development of government policy, PNNL has been developing an analytical framework to holistically evaluate the economics and non-proliferation merits of alternative approaches to managing the nuclear fuel cycle (i.e., cradle-to-grave). This paper provides an overview of the analytical framework and discusses preliminary results of an economic assessment of one RNFS approach: full-service nuclear fuel leasing. The specific focus of this paper is the metrics under development to systematically evaluate the non-proliferation merits of fuel-cycle management alternatives. Also discussed is the utility of an integrated assessment of the economics and non-proliferation merits of nuclear fuel leasing.

  10. Characterization plan for Hanford spent nuclear fuel

    SciTech Connect (OSTI)

    Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

    1994-12-01

    Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

  11. Utilization of Used Nuclear Fuel in a Potential Future US Fuel Cycle Scenario - 13499

    SciTech Connect (OSTI)

    Worrall, Andrew

    2013-07-01

    To date, the US reactor fleet has generated approximately 68,000 MTHM of used nuclear fuel (UNF) and even with no new nuclear build in the US, this stockpile will continue to grow at approximately 2,000 MTHM per year for several more decades. In the absence of reprocessing and recycle, this UNF is a liability and needs to be dealt with accordingly. However, with the development of future fuel cycle and reactor technologies in the decades ahead, there is potential for UNF to be used effectively and efficiently within a future US nuclear reactor fleet. Based on the detailed expected operating lifetimes, the future UNF discharges from the existing reactor fleet have been calculated on a yearly basis. Assuming a given electricity demand growth in the US and a corresponding growth demand for nuclear energy via new nuclear build, the future discharges of UNF have also been calculated on a yearly basis. Using realistic assumptions about reprocessing technologies and timescales and which future fuels are likely to be reprocessed, the amount of plutonium that could be separated and stored for future reactor technologies has been determined. With fast reactors (FRs) unlikely to be commercially available until 2050, any new nuclear build prior to then is assumed to be a light water reactor (LWR). If the decision is made for the US to proceed with reprocessing by 2030, the analysis shows that the UNF from future fuels discharged from 2025 onwards from the new and existing fleet of LWRs is sufficient to fuel a realistic future demand from FRs. The UNF arising from the existing LWR fleet prior to 2025 can be disposed of directly with no adverse effect on the potential to deploy a FR fleet from 2050 onwards. Furthermore, only a proportion of the UNF is required to be reprocessed from the existing fleet after 2025. All of the analyses and conclusions are based on realistic deployment timescales for reprocessing and reactor deployment. The impact of the delay in recycling the UNF

  12. Monitoring, Controlling and Safeguarding Radiochemical Streams at Spent Fuel Reprocessing Facilities, Part 2: Gamma-Ray Spectroscopic Methods

    SciTech Connect (OSTI)

    Schwantes, Jon M.; Bryan, Samuel A.; Orton, Christopher R.; Levitskaia, Tatiana G.; Fraga, Carlos G.

    2012-02-10

    The International Atomic Energy Agency (IAEA) has established international safeguards standards for fissionable material at spent fuel reprocessing plants to ensure that significant quantities of weapons-useable nuclear material are not diverted from these facilities. For large throughput nuclear facilities, it is difficult to satisfy the IAEA safeguards accountancy goal for detection of abrupt diversion. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resource-intensive destructive assay (DA). Leveraging new on-line non-destructive assay (NDA) process monitoring techniques in conjunction with the traditional and highly precise DA methods may provide an additional measure to nuclear material accountancy which would potentially result in a more timely, cost-effective and resource efficient means for safeguards verification at such facilities. By monitoring process control measurements (e.g. flowrates, temperatures, or concentrations of reagents, products or wastes), abnormal plant operations can be detected. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies based upon gamma-ray and optical spectroscopic measurements to potentially reduce the time and resource burden associated with current techniques. The Multi-Isotope Process (MIP) Monitor uses gamma spectroscopy and multivariate analysis to identify off-normal conditions in process streams. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major stable flowsheet reagents using UV-Vis, Near IR and Raman spectroscopy. Multi-variate analysis is also applied to the optical measurements in order to quantify concentrations of analytes of interest within a complex array of radiochemical streams. This paper will provide an overview of these methods and reports on-going efforts to develop

  13. Monitoring, Controlling and Safeguarding Radiochemical Streams at Spent Fuel Reprocessing Facilities with Optical and Gamma-Ray Spectroscopic Methods

    SciTech Connect (OSTI)

    Schwantes, Jon M.; Bryan, Samuel A.; Orton, Christopher R.; Levitskaia, Tatiana G.; Fraga, Carlos G.

    2012-11-06

    The International Atomic Energy Agency (IAEA) has established international safeguards standards for fissionable material at spent fuel reprocessing plants to ensure that significant quantities of weapons-useable nuclear material are not diverted from these facilities. For large throughput nuclear facilities, it is difficult to satisfy the IAEA safeguards accountancy goal for detection of abrupt diversion. Currently, methods to verify material control and accountancy (MC&A) at these facilities require time-consuming and resourceintensive destructive assay (DA). Leveraging new on-line non-destructive assay (NDA) process monitoring techniques in conjunction with the traditional and highly precise DA methods may provide an additional measure to nuclear material accountancy which would potentially result in a more timely, cost-effective and resource efficient means for safeguards verification at such facilities. By monitoring process control measurements (e.g. flowrates, temperatures, or concentrations of reagents, products or wastes), abnormal plant operations can be detected. Pacific Northwest National Laboratory (PNNL) is developing on-line NDA process monitoring technologies based upon gamma-ray and optical spectroscopic measurements to potentially reduce the time and resource burden associated with current techniques. The Multi-Isotope Process (MIP) Monitor uses gamma spectroscopy and multivariate analysis to identify offnormal conditions in process streams. The spectroscopic monitor continuously measures chemical compositions of the process streams including actinide metal ions (U, Pu, Np), selected fission products, and major stable flowsheet reagents using UV-Vis, Near IR and Raman spectroscopy. Multi-variate analysis is also applied to the optical measurements in order to quantify concentrations of analytes of interest within a complex array of radiochemical streams. This paper will provide an overview of these methods and reports on-going efforts to develop

  14. Advanced Nuclear Fuel Cycle Options

    SciTech Connect (OSTI)

    Roald Wigeland; Temitope Taiwo; Michael Todosow; William Halsey; Jess Gehin

    2010-06-01

    A systematic evaluation has been conducted of the potential for advanced nuclear fuel cycle strategies and options to address the issues ascribed to the use of nuclear power. Issues included nuclear waste management, proliferation risk, safety, security, economics and affordability, and sustainability. The two basic strategies, once-through and recycle, and the range of possibilities within each strategy, are considered for all aspects of the fuel cycle including options for nuclear material irradiation, separations if needed, and disposal. Options range from incremental changes to today’s implementation to revolutionary concepts that would require the development of advanced nuclear technologies.

  15. APEX nuclear fuel cycle for production of LWR fuel and elimination of radioactive waste

    SciTech Connect (OSTI)

    Steinberg, M.; Powell, J.R.

    1981-08-01

    The development of a nuclear fission fuel cycle is proposed which eliminates all the radioactive fission product waste effluent and the need for geological-age high level waste storage and provides a long term supply of fissile fuel for an LWR power reactor economy. The fuel cycle consists of reprocessing LWR spent fuel (1 to 2 years old) to remove the stable nonradioactive (NRFP, e.g. lanthanides, etc.) and short-lived fission products (SLFP e.g. half-lives of (1 to 2 years) and returning, in dilute form, the long-lived fission products, ((LLFPs, e.g. 30 y half-life Cs, Sr, and 10 y Kr, and 16 x 10/sup 6/ y I) and the transuranics (TUs, e.g. Pu, Am, Cm, and Np) to be refabricated into fresh fuel elements. Makeup fertile and fissile fuel are to be supplied through the use of a Spallator (linear accelerator spallation-target fuel-producer). The reprocessing of LWR fuel elements is to be performed by means of the Chelox process which consists of Airox treatment (air oxidation and hydrogen reduction) followed by chelation with an organic reagent (..beta..-diketonate) and vapor distillation of the organometallic compounds for separation and partitioning of the fission products.

  16. Apex nuclear fuel cycle for production of light water reactor fuel and elimination of radioactive waste

    SciTech Connect (OSTI)

    Steinberg, M.; Hiroshi, T.; Powell, J.R.

    1982-09-01

    The development of a nuclear fission fuel cycle is proposed that eliminates all the radioactive fission product (FP) waste effluent and the need for geological age high-level waste storage and provides a longterm supply of fissile fuel for a light water reactor (LWR) economy. The fuel cycle consists of reprocessing LWR spent fuel (1 to 2 yr old) to remove the stable nonradioactive FPs (NRFPs) e.g., lanthanides, etc.) and short-lived FPs (SLFP) (e.g., half-lives of less than or equal to 1 to 2 yr) and returning, in dilute form, the long-lived FPs (LLFPs) (e.g., 30-yr half-life cesium and strontium, 10-yr krypton, and 16 X 10/sup 6/-yr iodine) and the transuranics (TUs) (e.g., plutonium, americium, curium, and neptunium) to be refabricated into fresh fuel elements. Makeup fertile and fissile fuel (FF) are to be supplied through the use of the spallator (linear accelerator spallation-target fuel producer). The reprocessing of LWR fuel elements is to be performed by means of the chelox process, which consists of chopping and leaching with an organic chelating reagent (..beta..-diketonate) and distillation of the organometallic compounds formed for purposes of separating and partitioning the FPs. The stable NRFPs and SLFPs are allowed to decay to background in 10 to 20 yr for final disposal to the environment.

  17. Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

    DOE Patents [OSTI]

    Lloyd, Milton H.

    1983-01-01

    Method for direct coprocessing of nuclear fuels derived from a product stream of a fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

  18. Coprocessed nuclear fuels containing (U, Pu) values as oxides, carbides or carbonitrides

    DOE Patents [OSTI]

    Lloyd, M.H.

    1981-01-09

    Method for direct coprocessing of nuclear fuels derived from a product stream of fuels reprocessing facility containing uranium, plutonium, and fission product values comprising nitrate stabilization of said stream vacuum concentration to remove water and nitrates, neutralization to form an acid deficient feed solution for the internal gelation mode of sol-gel technology, green spherule formation, recovery and treatment for loading into a fuel element by vibra packed or pellet formation technologies.

  19. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    SciTech Connect (OSTI)

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-02-25

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R&D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  20. Modeling the Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    Jacob J. Jacobson; A. M. Yacout; G. E. Matthern; S. J. Piet; A. Moisseytsev

    2005-07-01

    The Advanced Fuel Cycle Initiative is developing a system dynamics model as part of their broad systems analysis of future nuclear energy in the United States. The model will be used to analyze and compare various proposed technology deployment scenarios. The model will also give a better understanding of the linkages between the various components of the nuclear fuel cycle that includes uranium resources, reactor number and mix, nuclear fuel type and waste management. Each of these components is tightly connected to the nuclear fuel cycle but usually analyzed in isolation of the other parts. This model will attempt to bridge these components into a single model for analysis. This work is part of a multi-national laboratory effort between Argonne National Laboratory, Idaho National Laboratory and United States Department of Energy. This paper summarizes the basics of the system dynamics model and looks at some results from the model.

  1. Spent nuclear fuel removal program at the West Valley Demonstration Project: Topical report

    SciTech Connect (OSTI)

    Connors, B. J.; Golden, M. P.; Valenti, P. J.; Winkel, J. J.

    1987-03-01

    The spent nuclear fuel removal program at the West Valley Demonstration Project (WVDP) consisted of removing the spent nuclear fuel (SNF) assemblies from the storage pool in the plant, loading them in shielded casks, and preparing the casks for transportation. So far, four fuel removal campaigns have been completed with the return of 625 spent nuclear fuel assemblies to their four utility owners. A fifth campaign, which is not yet completed, will transfer the remaining 125 fuel assemblies to a government site in Idaho. A spent fuel rod consolidation demonstration has been completed, and the storage canisters and their racks are being removed from the fuel receiving and storage pool to make way for installation of the size reduction equipment. A brief history of the West Valley reprocessing plant and the events leading to the storage and ownership of the spent nuclear fuel assemblies and their subsequent removal from West Valley are also recorded as background information. 3 refs., 16 figs., 9 tabs.

  2. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Bassett, C.H.

    1961-05-16

    A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

  3. Spent fuel accumulations and arisings in 'fuel user' states: implications for a world nuclear partnership

    SciTech Connect (OSTI)

    Lineberry, M.J.

    2007-07-01

    Consider the question: In a GNEP world, what are the implications of current spent fuel inventories and possible future accumulations in 'fuel user' states? The inventory today ({approx}95,000 MTHM) and the growth rate ({approx}4,000 MTHM/yr) in likely fuel user states is roughly twice that in the U.S., and thus the problem is substantial. The magnitude of the issue increases if, as is expected, fuel user states grow their nuclear capacity. For purposes of illustration, 2% annual growth assumed in this study. There is an implied imperative to remove spent fuel inventories from fuel user states in some reasonable time (if the spent fuel backlog cannot be cleared by 2100, what motivation is there for GNEP today?). Clearing the backlog requires massive reprocessing capacity, but that makes no economic or strategic sense until the advanced burner reactors are developed and deployed. Thus, burner reactor development is the pacing item with any practical GNEP schedule. With regard to economics, cost estimates for reprocessing (which drive key GNEP costs) are much too disparate today. The disparities must be lessened in order to permit rational decision-making. (author)

  4. Spent Nuclear Fuel project, project management plan

    SciTech Connect (OSTI)

    Fuquay, B.J.

    1995-10-25

    The Hanford Spent Nuclear Fuel Project has been established to safely store spent nuclear fuel at the Hanford Site. This Project Management Plan sets forth the management basis for the Spent Nuclear Fuel Project. The plan applies to all fabrication and construction projects, operation of the Spent Nuclear Fuel Project facilities, and necessary engineering and management functions within the scope of the project

  5. Nuclear reactors and the nuclear fuel cycle

    SciTech Connect (OSTI)

    Pearlman, H

    1989-11-01

    According to the author, the first sustained nuclear fission chain reaction was not at the University of Chicago, but at the Oklo site in the African country of Gabon. Proof of this phenomenon is provided by mass spectrometric and analytical chemical measurements by French scientists. The U.S. experience in developing power-producing reactors and their related fuel and fuel cycles is discussed.

  6. NUCLEAR REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

    Anderson, W.F.; Tellefson, D.R.; Shimazaki, T.T.

    1962-04-10

    A plate type fuel element which is particularly useful for organic cooled reactors is described. Generally, the fuel element comprises a plurality of fissionable fuel bearing plates held in spaced relationship by a frame in which the plates are slidably mounted in grooves. Clearance is provided in the grooves to allow the plates to expand laterally. The plates may be rigidly interconnected but are floatingly supported at their ends within the frame to allow for longi-tudinal expansion. Thus, this fuel element is able to withstand large temperature differentials without great structural stresses. (AEC)

  7. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  8. NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS...

    Office of Scientific and Technical Information (OSTI)

    Title list of documents made publicly available, January 1-31, 1998 NONE 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS; BIBLIOGRAPHIES; NUCLEAR POWER PLANTS;...

  9. Spent Nuclear Fuel Project Safety Management Plan

    SciTech Connect (OSTI)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities.

  10. NUCLEAR REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

    Currier, E.L. Jr.; Nicklas, J.H.

    1963-06-11

    A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

  11. Methodology for Determining the Radiological Status of a Process: Application to Decommissioning of a Fuel Reprocessing Plant

    SciTech Connect (OSTI)

    Girones, Ph.; Ducros, C.; Legoaller, C.; Lamadie, F.; Fulconis, J.M.; Thiebaut, V.; Mahe, C.

    2006-07-01

    Decommissioning a nuclear facility is subject to various constraints including regulatory safety requirements, but also the obligation to limit the waste volume and toxicity. To meet these requirements the activity level in each component must be known at each stage of decommissioning, from the preliminary studies to the final release of the premises. This document describes a set of methods used to determine the radiological state of a spent fuel reprocessing plant. This approach begins with a bibliographical survey covering the nature of the chemical processes, the operational phases, and the radiological assessments during the plant operating period. In this phase it is also very important to analyze incidents and waste management practices. All available media should be examined, including photos and videos which can provide valuable data and must not be disregarded. At the end of this phase, any items requiring verification or additional data are reviewed to define further investigations. Although it is not unusual at this point to carry out an additional bibliographical survey, the essential task is to carry out in situ measurements. The second phase thus consists in performing in situ measurement campaigns involving essentially components containing significant activity levels. The most routinely used methods combine the results of elementary measurements such as the dose rate or more sophisticated measurements such as gamma spectrometry using CdZnTe detectors and gamma imaging to estimate and localize the radioactivity. Each instrument provides part of the answer (location of a contamination hot spot, standard spectrum, activity). The results are combined and verified through the use of calculation codes: Mercure, Visiplan and Microshield. (authors)

  12. FUEL ELEMENTS FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Blainey, A.; Lloyd, H.

    1961-07-11

    A method of sheathing a tubular fuel element for a nuclear reactor is described. A low melting metal core member is centered in a die, a layer of a powdered sheathing substance is placed on the bottom of the die, the tubular fuel element is inserted in the die, the space between the tubular fuel element and the die walls and core member is filled with the same powdered sheathing substance, a layer of the same substance is placed over the fissile material, and the charge within the die is subjected to pressure in the direction of the axis of the fuel element at the sintering temperature of the protective substance.

  13. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Bassett, C.H.

    1961-05-01

    A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

  14. Used Fuel Disposition Used Nuclear Fuel Storage and Transportation

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Used Nuclear Fuel Storage and Transportation Overview Steve Marschman Field Demonstration Lead Idaho National Laboratory NEET ASI Review Meeting September 17, 2014 Used Fuel ...

  15. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect (OSTI)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and

  16. Nuclear Fuel Leasing, Recycling and proliferation: Modeling a Global View

    SciTech Connect (OSTI)

    Crozat, M P; Choi, J; Reis, V H; Hill, R

    2004-03-10

    On February 11, 2004, U.S. President George W. Bush, in a speech to the National Defense University stated: ''The world must create a safe, orderly system to field civilian nuclear plants without adding to the danger of weapons proliferation. The world's leading nuclear exporters should ensure that states have reliable access at reasonable cost to fuel for civilian reactors, so long as those states renounce enrichment and reprocessing. Enrichment and reprocessing are not necessary for nations seeking to harness nuclear energy for peaceful purposes.'' This concept would require nations to choose one of two paths for civilian nuclear development: those that only have reactors and those that contain one or more elements of the nuclear fuel cycle, including recycling. ''Fuel cycle'' states would enrich uranium, manufacture and lease fuel to ''reactor'' states and receive the reactor states' spent fuel. All parties would accede to stringent security and safeguard standards, embedded within a newly invigorated international regime. Reactor states would be relieved of the financial, environmental (and political) burden of enriching and manufacturing fuel and dealing with spent fuel. Fuel cycle states would potentially earn money on leasing the fuel and perhaps on sales of reactors to the reactor states. Such a leasing concept is especially interesting in scenarios which envision growth in nuclear power, and an important consideration for such a nuclear growth regime is the role of recycling of civilian spent fuel. Recycling holds promise for improved management of spent fuel and efficient utilization of resources, but continues to raise the specter of a world with uncontrolled nuclear weapons proliferation. If done effectively, a fuel-leasing concept could help create a political and economic foundation for significant growth of clean, carbon-free nuclear power while providing a mechanism for significant international cooperation to reduce proliferation concern. This

  17. Nuclear reactor composite fuel assembly

    DOE Patents [OSTI]

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  18. Sensitivity analysis and optimization of the nuclear fuel cycle

    SciTech Connect (OSTI)

    Passerini, S.; Kazimi, M. S.; Shwageraus, E.

    2012-07-01

    A sensitivity study has been conducted to assess the robustness of the conclusions presented in the MIT Fuel Cycle Study. The Once Through Cycle (OTC) is considered as the base-line case, while advanced technologies with fuel recycling characterize the alternative fuel cycles. The options include limited recycling in LWRs and full recycling in fast reactors and in high conversion LWRs. Fast reactor technologies studied include both oxide and metal fueled reactors. The analysis allowed optimization of the fast reactor conversion ratio with respect to desired fuel cycle performance characteristics. The following parameters were found to significantly affect the performance of recycling technologies and their penetration over time: Capacity Factors of the fuel cycle facilities, Spent Fuel Cooling Time, Thermal Reprocessing Introduction Date, and in core and Out-of-core TRU Inventory Requirements for recycling technology. An optimization scheme of the nuclear fuel cycle is proposed. Optimization criteria and metrics of interest for different stakeholders in the fuel cycle (economics, waste management, environmental impact, etc.) are utilized for two different optimization techniques (linear and stochastic). Preliminary results covering single and multi-variable and single and multi-objective optimization demonstrate the viability of the optimization scheme. (authors)

  19. Spallator and APEX nuclear fuel cycle: a new option for nuclear power

    SciTech Connect (OSTI)

    Steinberg, M.

    1982-01-01

    A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high-energy (1 to 2 GeV) protons on a heavy-metal target. The neutrons are absorbed in a surrounding natural-uranium or thorium blanket in which fissile Pu-239 to U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high-beam-current continuous-wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a self-sufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of short-lived fission products external to the fuel cycle eliminates the need for long-term geological age shortage of fission-product waste.

  20. The Spallator and APEX nuclear fuel cycle: a new option for nuclear power

    SciTech Connect (OSTI)

    Steinberg, M.

    1983-02-01

    A new nuclear fuel cycle is described which provides a long term supply of nuclear fuel for the thermal LWR nuclear power reactors and eliminates the need for long-term storage of radioactive waste. Fissile fuel is produced by the Spallator which depends on the production of spallation neutrons by the interaction of high energy (1 to 2 GeV) protons on a heavy metal target. The neutrons are absorbed in a surrounding natural uranium or thorium blanket in which fissile Pu-239 or U-233 is produced. Advances in linear accelerator technology makes it possible to design and construct a high beam current continuous wave proton linac for production purposes. The target is similar to a sub-critical reactor and produces heat which is converted to electricity for supplying the linac. The Spallator is a selfsufficient fuel producer, which can compete with the fast breeder. The APEX fuel cycle depends on recycling the transuranics and long-lived fission products while extracting the stable and short-lived fission products when reprocessing the fuel. Transmutation and decay within the fuel cycle and decay of the short-lived fission products external to the fuel cycle eliminates the need for long-term geological age storage of fission product waste.

  1. Protected Nuclear Fuel Element

    DOE Patents [OSTI]

    Kittel, J. H.; Schumar, J. F.

    1962-12-01

    A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

  2. Fire resistant nuclear fuel cask

    DOE Patents [OSTI]

    Heckman, Richard C.; Moss, Marvin

    1979-01-01

    The disclosure is directed to a fire resistant nuclear fuel cask employing reversibly thermally expansible bands between adjacent cooling fins such that normal outward flow of heat is not interfered with, but abnormal inward flow of heat is impeded or blocked.

  3. World nuclear fuel cycle requirements 1991

    SciTech Connect (OSTI)

    Not Available

    1991-10-10

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

  4. Fully ceramic nuclear fuel and related methods

    DOE Patents [OSTI]

    Venneri, Francesco; Katoh, Yutai; Snead, Lance Lewis

    2016-03-29

    Various embodiments of a nuclear fuel for use in various types of nuclear reactors and/or waste disposal systems are disclosed. One exemplary embodiment of a nuclear fuel may include a fuel element having a plurality of tristructural-isotropic fuel particles embedded in a silicon carbide matrix. An exemplary method of manufacturing a nuclear fuel is also disclosed. The method may include providing a plurality of tristructural-isotropic fuel particles, mixing the plurality of tristructural-isotropic fuel particles with silicon carbide powder to form a precursor mixture, and compacting the precursor mixture at a predetermined pressure and temperature.

  5. Pilot Application to Nuclear Fuel Cycle Options

    Office of Energy Efficiency and Renewable Energy (EERE)

    A Screening Method for Guiding R&D Decisions: Pilot Application to Screen Nuclear Fuel Cycle Options

  6. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOE Patents [OSTI]

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  7. Compositions and methods for treating nuclear fuel

    DOE Patents [OSTI]

    Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

    2013-08-13

    Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

  8. Compositions and methods for treating nuclear fuel

    DOE Patents [OSTI]

    Soderquist, Chuck Z; Johnsen, Amanda M; McNamara, Bruce K; Hanson, Brady D; Smith, Steven C; Peper, Shane M

    2014-01-28

    Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

  9. Issues Associated with IAEA Involvement in Assured Nuclear Fuel Supply Arrangements

    SciTech Connect (OSTI)

    Kessler, Carol E.; Mathews, Carrie E.

    2008-02-08

    Assured nuclear fuel supply has been discussed at various times as a mechanism to help limit expansion of enrichment and reprocessing (E&R) capability beyond current technology holders. Given the events in the last few years in North Korea and Iran, concern over weapons capabilities gained from acquisition of E&R capabilities has heightened and brought assured nuclear fuel supply (AFS) again to the international agenda. Successful AFS programs can be valuable contributions to strengthening the nonproliferation regime and helping to build public support for expanding nuclear energy.

  10. NUCLEAR MATERIAL ATTRACTIVENESS: AN ASSESSMENT OF MATERIAL FROM PHWR'S IN A CLOSED THORIUM FUEL CYCLE

    SciTech Connect (OSTI)

    Sleaford, B W; Collins, B A; Ebbinghaus, B B; Bathke, C G; Prichard, A W; Wallace, R K; Smith, B W; Hase, K R; Bradley, K S; Robel, M; Jarvinen, G D; Ireland, J R; Johnson, M W

    2010-04-26

    This paper examines the attractiveness of material mixtures containing special nuclear materials (SNM) associated with reprocessing and the thorium-based LWR fuel cycle. This paper expands upon the results from earlier studies that examined the attractiveness of SNM associated with the reprocessing of spent light water reactor (LWR) fuel by various reprocessing schemes and the recycle of plutonium as a mixed oxide (MOX) fuel in LWR. This study shows that {sup 233}U that is produced in thorium-based fuel cycles is very attractive for weapons use. Consistent with other studies, these results also show that all fuel cycles examined to date need to be rigorously safeguarded and provided moderate to high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE), and are based on the calculation of 'attractiveness levels' that has been couched in terms chosen for consistency with those normally used for nuclear materials in DOE nuclear facilities. The methodology and key findings will be presented.

  11. Used nuclear fuel separations process simulation and testing

    SciTech Connect (OSTI)

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D.

    2013-07-01

    Recent efforts in separations process simulation at Argonne have expanded from the traditional focus on solvent extraction flowsheet design in order to capture process dynamics and to simulate other components, processing and systems of a used nuclear fuel reprocessing plant. For example, the Argonne Model for Universal Solvent Extraction (AMUSE) code has been enhanced to make it both more portable and more readily extensible. Moving away from a spreadsheet environment makes the addition of new species and processes simpler for the expert user, which should enable more rapid implementation of chemical models that simulate evolving processes. The dyAMUSE (dynamic AMUSE) version allows the simulation of transient behavior across an extractor. Electrochemical separations have now been modeled using spreadsheet codes that simulate the electrochemical recycle of fast reactor fuel. The user can follow the evolution of the salt, products, and waste compositions in the electro-refiner, cathode processors, and drawdown as a function of fuel batches treated. To further expand capabilities in integrating multiple unit operations, a platform for linking mathematical models representing the different operations that comprise a reprocessing facility was adapted to enable systems-level analysis and optimization of facility functions. (authors)

  12. Nuclear fuel microsphere gamma analyzer

    DOE Patents [OSTI]

    Valentine, Kenneth H.; Long, Jr., Ernest L.; Willey, Melvin G.

    1977-01-01

    A gamma analyzer system is provided for the analysis of nuclear fuel microspheres and other radioactive particles. The system consists of an analysis turntable with means for loading, in sequence, a plurality of stations within the turntable; a gamma ray detector for determining the spectrum of a sample in one section; means for analyzing the spectrum; and a receiver turntable to collect the analyzed material in stations according to the spectrum analysis. Accordingly, particles may be sorted according to their quality; e.g., fuel particles with fractured coatings may be separated from those that are not fractured, or according to other properties.

  13. Study on Nuclear Fuel Cycle System using Coated-particle Fuel and Hybrid Micro-capsule Separation Method

    SciTech Connect (OSTI)

    Wakabayashi, Toshio; Mimura, Hitoshi

    2007-07-01

    A new concept of nuclear fuel cycle system for the fast reactors was proposed. The concept is mainly composed of a reprocessing and separation system based on hybrid micro-capsules containing extractant, a remote fuel fabrication system for carbide coated particle fuels and a gas-cooled fast reactor. The gas-cooled fast reactor system features a fast-neutron-spectrum helium-cooled reactor and closed fuel cycle to be able to have sustainable energy resources in future. The feasibility study of the concept has been performed from the viewpoints core performance, radioactive waste reduction, separation technology, fabrication technology, etc. It was shown that the new concept of nuclear fuel cycle system including the gas-cooled fast reactor is very attractive to the future energy supply system in this study. (authors)

  14. Method for locating defective nuclear fuel elements

    SciTech Connect (OSTI)

    Lawrie, W.E.; White, N.W.; Womack, R.E.

    1982-02-02

    Defects in nuclear fuel elements are ascertained and located within an assembled fuel assembly by ultrasonic means. In a typical embodiment of the invention, an ultrasonic search unit is positioned within the fuel assembly opposite the lower plenum of the fuel element to be tested. An ultrasonic pulse is radially projected into the element. Defective fuel elements are ascertained by ultrasonic reflection measurements.

  15. Method for producing nuclear fuel

    DOE Patents [OSTI]

    Haas, Paul A.

    1983-01-01

    Nuclear fuel is made by contacting an aqueous solution containing an actinide salt with an aqueous solution containing ammonium hydroxide, ammonium oxalate, or oxalic acid in an amount that will react with a fraction of the actinide salt to form a precipitate consisting of the hydroxide or oxalate of the actinide. A slurry consisting of the precipitate and solution containing the unreacted actinide salt is formed into drops which are gelled, calcined, and pressed to form pellets.

  16. Method for producing nuclear fuel

    SciTech Connect (OSTI)

    Haas, P.A.

    1981-04-24

    Nuclear fuel is made by contacting an aqueous solution containing an actinide salt with an aqueous solution containing ammonium hydroxide, ammonium oxalate, or oxalic acid in an amount that will react with a fraction of the actinide salt to form a precipitate consisting of the hydroxide or oxalate of the actinide. A slurry consisting of the precipitate and solution containing the unreacted actinide salt is formed into drops which are gelled, calcined, and pressed to form pellets.

  17. Curium concentration in spent nuclear fuel

    SciTech Connect (OSTI)

    Beddingfield, D. H.; Swinhoe, M. T.

    2004-01-01

    Neutron measurements are frequently used to characterize spent nuclear fuel. Curium is the primary neutron source from most spent nuclear fuel materials. Recent developments in nuclear safeguards measurements of spent nuclear fuel have increased the reliance upon curium assay for materials accounting on the back end of the fuel cycle. The curium assay is used to determine the fuel composition by the curium-ratio technique. The interpretation of these measurements is based upon the results of calculations using depletion codes. In this paper we will examine depletion code results to determine if there is reason for concern for the reliability of curium concentration from calculation.

  18. Lead Slowing Down Spectrometry Analysis of Data from Measurements on Nuclear Fuel

    SciTech Connect (OSTI)

    Warren, Glen A.; Anderson, Kevin K.; Kulisek, Jonathan A.; Danon, Yaron; Weltz, Adam; Gavron, Victor A.; Harris, Jason; Stewart, Trevor N.

    2015-01-12

    Improved non-destructive assay of isotopic masses in used nuclear fuel would be valuable for nuclear safeguards operations associated with the transport, storage and reprocessing of used nuclear fuel. Our collaboration is examining the feasibility of using lead slowing down spectrometry techniques to assay the isotopic fissile masses in used nuclear fuel assemblies. We present the application of our analysis algorithms on measurements conducted with a lead spectrometer. The measurements involved a single fresh fuel pin and discrete 239Pu and 235U samples. We are able to describe the isotopic fissile masses with root mean square errors over seven different configurations to 6.35% for 239Pu and 2.7% for 235U over seven different configurations. Funding Source(s):

  19. Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options

    SciTech Connect (OSTI)

    Brent W. Dixon; Steven J. Piet

    2004-10-01

    tripling market share by 2100 from the current 8.4% to 25%, equivalent to continuing the average market growth of last 50 years for an additional 100 years. Five primary spent fuel management strategies are assessed against each of the energy futures to determine the number of geological repositories needed and how the first repository would be used. The geological repository site at Yucca Mountain, Nevada, has the physical potential to accommodate all the spent fuel that will be generated by the current fleet of domestic commercial nuclear reactors, even with license extensions. If new nuclear plants are built in the future as replacements or additions, the United States will need to adopt spent fuel treatment to extend the life of the repository. Should a significant number of new nuclear plants be built, advanced fuel recycling will be needed to fully manage the spent fuel within a single repository. The analysis also considers the timeframe for most efficient implementation of new spent fuel management strategies. The mix of unprocessed spent fuel and processed high level waste in Yucca Mountain varies with each future and strategy. Either recycling must start before there is too much unprocessed waste emplaced or unprocessed waste will have to be retrieved later with corresponding costs. For each case, the latest date to implement reprocessing without subsequent retrieval is determined.

  20. Safe Advantage on Dry Interim Spent Nuclear Fuel Storage

    SciTech Connect (OSTI)

    Romanato, L.S.

    2008-07-01

    This paper aims to present the advantages of dry cask storage in comparison with the wet storage (cooling water pools) for SNF. When the nuclear fuel is removed from the core reactor, it is moved to a storage unit and it wait for a final destination. Generally, the spent nuclear fuel (SNF) remains inside water pools within the reactors facility for the radioactive activity decay. After some period of time in pools, SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing facilities, or still, wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet facilities, depending on the method adopted by the nuclear power plant or other plans of the country. Interim storage, up to 20 years ago, was exclusively wet and if the nuclear facility had to be decommissioned another storage solution had to be found. At the present time, after a preliminary cooling of the SNF elements inside the water pool, the elements can be stored in dry facilities. This kind of storage does not need complex radiation monitoring and it is safer then wet one. Casks, either concrete or metallic, are safer, especially on occurrence of earthquakes, like that occurred at Kashiwazaki-Kariwa nuclear power plant, in Japan on July 16, 2007. (authors)

  1. Nuclear Fuel Cycle Option Catalog SAND2015-2174 W

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    benefits and challenges of nuclear fuel cycle options (i.e., the complete nuclear ... of Energy, Office of Nuclear Energy, Fuel Cycle Research and Development program. ...

  2. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Bassett, C.H.

    1961-07-11

    Nuclear reactor fuel elements of the type in which the flssionsble material is in ceramic form, such as uranium dioxide, are described. The fuel element is comprised of elongated inner and outer concentric spaced tubular members providing an annular space therebetween for receiving the fissionable material, the annular space being closed at both ends and the inner tube being open at both ends. The fuel is in the form of compressed pellets of ceramic fissionsble material having the configuration of split bushings formed with wedge surfaces and arranged in seriated inner and outer concentric groups which are urged against the respective tubes in response to relative axial movement of the pellets in the direction toward each other. The pairs of pellets are axially urged together by a resilient means also enclosed within the annulus. This arrangement-permits relative axial displacement of the pellets during use dial stresses on the inner and outer tube members and yet maintains the fuel pellets in good thermal conductive relationship therewith.

  3. Nuclear core and fuel assemblies

    DOE Patents [OSTI]

    Downs, Robert E.

    1981-01-01

    A fast flux nuclear core of a plurality of rodded, open-lattice assemblies having a rod pattern rotated relative to a rod support structure pattern. Elongated fuel rods are oriented on a triangular array and laterally supported by grid structures positioned along the length of the assembly. Initial inter-assembly contact is through strongbacks at the corners of the support pattern and peripheral fuel rods between adjacent assemblies are nested so as to maintain a triangular pitch across a clearance gap between the other portions of adjacent assemblies. The rod pattern is rotated relative to the strongback support pattern by an angle .alpha. equal to sin .sup.-1 (p/2c), where p is the intra-assembly rod pitch and c is the center-to-center spacing among adjacent assemblies.

  4. EM Completes Project to Maintain Water Quality of Spent Nuclear...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ... This was an outstanding accomplishment." Addthis Related Articles Workers demolish the CPP 601-602 Spent Nuclear Fuel Reprocessing Complex at the Idaho Site. A Decade of Cleanup ...

  5. Deputy Secretary Poneman's Remarks at the Third Annual Nuclear...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ... HEU deal, would harness the power of the commercial market in nuclear fuel, to the national security imperative of minimizing the spread of enrichment and reprocessing facilities. ...

  6. Deputy Secretary Poneman's Remarks at the Nuclear Deterrence...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ... that seek to assure countries that they may gain reliable access to nuclear fuel services without developing their own uranium enrichment and plutonium reprocessing capabilities. ...

  7. Role of inorganic chemistry on nuclear energy examined

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    operation for electricity production; and the "back-end" steps, which are aimed toward the safe handling and subsequent reprocessing or disposition of spent nuclear fuel materials. ...

  8. 10 CFR Part 810 | National Nuclear Security Administration |...

    National Nuclear Security Administration (NNSA)

    A specific authorization is also required for any assistance involving sensitive nuclear technologies (enrichment, reprocessing, plutonium fuel, and heavy water production, ...

  9. West Valley Demonstration Project DOE Manual 435.1-1 Waste Incidental to Reprocessing Evaluations and Determinations

    Broader source: Energy.gov [DOE]

    DOE Manual 435.1-1, Radioactive Waste Management Manual provides that the DOE may determine that certain waste from reprocessing spent nuclear fuel is waste incidental to reprocessing, is not high-level waste and may be managed and disposed of as low-level waste if the waste meets the criteria in DOE Manual 435.1-1, Chapter II, Section B.

  10. Nuclear Fuel Cycle | Department of Energy

    Office of Environmental Management (EM)

    advice to DOE regarding research and development of nuclear fuel and waste management technologies that meet the nation's energy supply, environmental, and energy security needs. ...

  11. VISION -- A Dynamic Model of the Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    J. J. Jacobson; A. M. Yacout; S. J. Piet; D. E. Shropshire; G. E. Matthern

    2006-02-01

    The Advanced Fuel Cycle Initiative’s (AFCI) fundamental objective is to provide technology options that – if implemented – would enable long-term growth of nuclear power while improving sustainability and energy security. The AFCI organization structure consists of four areas; Systems Analysis, Fuels, Separations and Transmutations. The Systems Analysis Working Group is tasked with bridging the program technical areas and providing the models, tools, and analyses required to assess the feasibility of design and deploy¬ment options and inform key decision makers. An integral part of the Systems Analysis tool set is the development of a system level model that can be used to examine the implications of the different mixes of reactors, implications of fuel reprocessing, impact of deployment technologies, as well as potential “exit” or “off ramp” approaches to phase out technologies, waste management issues and long-term repository needs. The Verifiable Fuel Cycle Simulation Model (VISION) is a computer-based simulation model that allows performing dynamic simulations of fuel cycles to quantify infrastructure requirements and identify key trade-offs between alternatives. VISION is intended to serve as a broad systems analysis and study tool applicable to work conducted as part of the AFCI (including costs estimates) and Generation IV reactor development studies.

  12. On-Line Physical Property Process Measurements for Nuclear Fuel Recycling

    SciTech Connect (OSTI)

    Pappas, Richard A.; Bond, Leonard J.; Greenwood, Margaret S.; Hostick, Cody J.

    2007-07-01

    The Global Nuclear Energy Partnership (GNEP) is looking to close the nuclear fuel cycle and demonstrate key fuel recycling technologies, while at the same time reducing proliferation risks. A key element of GNEP is the demonstration of the uranium extraction (UREX) +1a process, and potentially other fuel reprocessing schemes. Advanced recycling of nuclear fuel will require improved on-line monitoring and process control. Advanced ultrasonic sensor technology can be a critical component of a process quality control strategy that is designed to determine the sources of variability and minimize their impact on the quality of the end product. PNNL ultrasonic devices and methodologies, many of which were initially developed and deployed to address the needs of the DOE Hanford site, provide on-line physical property measurement useful in optimizing plant capacity, assuring cost-effective analyses, and satisfying direct sampling requirements.. A select collection of PNNL ultrasonic technology is discussed in this context. (authors)

  13. Used Fuel Disposition Used Nuclear Fuel Storage and Transportation

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Used Nuclear Fuel Storage and Transportation Overview Steve Marschman Field Demonstration Lead Idaho National Laboratory NEET ASI Review Meeting September 17, 2014 Used Fuel Disposition Today's Discussion n Our R&D Objectives n What Guides Our Work n FY14 and FY15 Work - Full-Scale High Burn-Up Demo - Experiments - Transportation - Analysis Used Fuel Disposition 3 Overall Objectives * Develop the technical bases to demonstrate the continued safe and secure storage of used nuclear

  14. Electrorefining Experience For Pyrochemical Reprocessing of Spent EBR-II Driver Fuel

    SciTech Connect (OSTI)

    S. X. Li; T. A. Johnson; B. R. Westphal; K. M. Goff; R. W. Benedict

    2005-10-01

    Pyrochemical processing has been implemented for the treatment of spent fuel from the Experimental Breeder Reactor-II (EBR-II) at Idaho National Laboratory since 1996. This report summarizes technical advancements made in electrorefining of spent EBR-II driver fuel in the Mk-IV electrorefiner since the pyrochemical processing was integrated into the AFCI program in 2002. The significant advancements include improving uranium dissolution and noble metal retention from chopped fuel segments, increasing cathode current efficiency, and achieving co-collection of zirconium along with uranium from the cadmium pool.

  15. Human capital needs - teaching, training and coordination for nuclear fuel cycle

    SciTech Connect (OSTI)

    Retegan, T.; Ekberg, C.; John, J.; Nordlund, A.

    2013-07-01

    Human capital is the accumulation of competencies, knowledge, social and creativity skills and personality attributes, which are necessary to perform work so as to produce economic value. In the frame of the nuclear fuel cycle, this is of paramount importance that the right human capital exists and in Europe this is fostered by a series of integrated or directed projects. The teaching, training and coordination will be discussed in the frame of University curricula with examples from several programs, like e.g. the Master of Nuclear Engineering at Chalmers University, Sweden and two FP7 EURATOM Projects: CINCH - a project for cooperation in nuclear chemistry - and ASGARD - a research project on advanced or novel nuclear fuels and their reprocessing issues for generation IV reactors. The integration of the university curricula in the market needs but also the anchoring in the research and future fuel cycles will be also discussed, with examples from the ASGARD project. (authors)

  16. Specialized Disposal Sites for Different Reprocessing Plant Wastes

    SciTech Connect (OSTI)

    Forsberg, Charles W.; Driscoll, Michael J.

    2007-07-01

    Once-through fuel cycles have one waste form: spent nuclear fuel (SNF). In contrast, the reprocessed SNF yields multiple wastes with different chemical, physical, and radionuclide characteristics. The different characteristics of each waste imply that there are potential cost and performance benefits to developing different disposal sites that match the disposal requirements of different waste. Disposal sites as defined herein may be located in different geologies or in a single repository containing multiple sections, each with different characteristics. The paper describes disposal options for specific wastes and the potential for a waste management system that better couples various reprocessing plant wastes with disposal facilities. (authors)

  17. Noble gas atmospheric monitoring at reprocessing facilities

    SciTech Connect (OSTI)

    Nakhleh, C.W.; Perry, R.T. Jr.; Poths, J.; Stanbro, W.D.; Wilson, W.B.; Fearey, B.L.

    1997-05-01

    The discovery in Iraq after the Gulf War of the existence of a large clandestine nuclear-weapon program has led to an across-the-board international effort, dubbed Programme 93+2, to improve the effectiveness and efficiency of International Atomic Energy Agency (IAEA) safeguards. One particularly significant potential change is the introduction of environmental monitoring (EM) techniques as an adjunct to traditional safeguards methods. Monitoring of stable noble gas (Kr, Xe) isotopic abundances at reprocessing plant stacks appears to be able to yield information on the burnup and type of the fuel being processed. To estimate the size of these signals, model calculations of the production of stable Kr, Xe nuclides in reactor fuel and the subsequent dilution of these nuclides in the plant stack are carried out for two case studies: reprocessing of PWR fuel with a burnup of 35 GWd/tU, and reprocessing of CAND fuel with a burnup of 1 GWd/tU. For each case, a maximum-likelihood analysis is used to determine the fuel burnup and type from the isotopic data.

  18. Monitoring arrangement for vented nuclear fuel elements

    DOE Patents [OSTI]

    Campana, Robert J.

    1981-01-01

    In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

  19. Nuclear fuel: a new market dynamic

    SciTech Connect (OSTI)

    Kee, Edward D.

    2007-12-15

    After almost 20 years of low nuclear fuel prices, buyers have come to expect that these low and stable nuclear fuel prices will continue. This conventional wisdom may not reflect the significant changes and higher prices that growing demand, and the end of secondary sources of uranium and enrichment, will bring. (author)

  20. Spent Nuclear Fuel (SNF) Project Execution Plan

    SciTech Connect (OSTI)

    LEROY, P.G.

    2000-11-03

    The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

  1. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, V.T.

    1993-04-06

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  2. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, Victor T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  3. Behavior of iodine in the dissolution of spent nuclear fuels

    SciTech Connect (OSTI)

    Sakurai, Tsutomu; Komatsu, Kazunori; Takahashi, A.

    1997-08-01

    The results of laboratory-scale experiments concerning the behavior of iodine in the dissolution of spent nuclear fuels, which were carried out at the Japan Atomic Energy Research Institute, are summarized. Based on previous and new experimental results, the difference in quantity of residual iodine in the fuel solution between laboratory-scale experiments and reprocessing plants is discussed, Iodine in spent fuels is converted to the following four states: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid generated in the dissolution, (3) formation of a colloid of insoluble iodides such as AgI and PdI{sub 2}, and (4) deposition on insoluble residue. Nitrous acid controls the amount of colloid formed. As a result, up to 10% of iodine in spent fuels is retained in the fuel solution, up to 3% is deposited on insoluble residue, and the balance volatilizes to the off-gas, Contrary to earlier belief, when the dissolution is carried out in 3 to 4 M HNO{sub 3} at 100{degrees}C, the main iodine species in a fuel solution is a colloid, not iodate, Immediately after its formation, the colloid is unstable and decomposes partially in the hot nitric acid solution through the following reaction: AgI(s) + 2HNO{sub 3}(aq) = {1/2}I{sub 2}(aq) + AgNO{sub 3}(aq) + NO{sub 2}(g) + H{sub 2}O(1). For high concentrations of gaseous iodine, I{sub 2}(g), and NO{sub 2}, this reaction is reversed towards formation of the colloid (AgI). Since these concentrations are high near the liquid surface of a plant-scale dissolver, there is a possibility that the colloid is formed there through this reversal, Simulations performed in laboratory-scale experiments demonstrated this reversal, This phenomenon can be one reason the quantity of residual iodine in spent fuels is higher in reprocessing plants than in laboratory-scale experiments. 17 refs., 5 figs., 3 tabs.

  4. World nuclear fuel cycle requirements 1990

    SciTech Connect (OSTI)

    Not Available

    1990-10-26

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management.

  5. Comparison studies of head-end reprocessing using three LWR fuels

    SciTech Connect (OSTI)

    Goode, J.H.; Stacy, R.G.; Vaughen, V.C.A.

    1980-06-01

    The removal of {sup 3}H by voloxidation and the dissolution behavior of two PWR and one BWR fuels were compared in hot-cell studies. The experiments showed that >99% of the {sup 3}H contained in the irradiated UO{sub 2} was volatilized by oxidation in air at 753{sup 0}K (480{sup 0}C). The oxidation did not affect the dissolution of the uranium and plutonium in 7 M HNO{sub 3} (0.02 to 0.03% insoluble plutonium) but did create a fission-product residue that was two to three times more insoluble. From 40 to 69% of the ternary fission-product {sup 3}H was found in the Zircaloy cladding of the fuel rods. Voloxidation had little effect on the {sup 3}H held in the Zircaloy cladding; oxidation for 6 h at 753{sup 0}K released only 0.05% of the {sup 3}H.

  6. DOE/EIS-0396\tDraft Global Nuclear Energy Partnership Programmatic...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ... High-Level Radioactive Waste High-level radioactive waste is defined as: 1) the highly radioactive material resulting from the reprocessing of spent nuclear fuel, including liquid ...

  7. Fuel cycle analysis of once-through nuclear systems.

    SciTech Connect (OSTI)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  8. A Critical Step Toward Sustainable Nuclear Fuel Disposal | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    A Critical Step Toward Sustainable Nuclear Fuel Disposal A Critical Step Toward Sustainable Nuclear Fuel Disposal January 26, 2012 - 2:30pm Addthis Secretary Chu Secretary Chu...

  9. Review of Used Nuclear Fuel Storage and Transportation Technical...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis Review of Used Nuclear Fuel Storage and Transportation Technical Gap Analysis While both wet and dry...

  10. Current Status of the Spent Nuclear Fuel Management Program in...

    Office of Scientific and Technical Information (OSTI)

    Current Status of the Spent Nuclear Fuel Management Program in the United States. Citation Details In-Document Search Title: Current Status of the Spent Nuclear Fuel Management...

  11. Nuclear Fuel Cycle & Vulnerabilities (Technical Report) | SciTech...

    Office of Scientific and Technical Information (OSTI)

    Nuclear Fuel Cycle & Vulnerabilities Citation Details In-Document Search Title: Nuclear Fuel Cycle & Vulnerabilities The objective of safeguards is the timely detection of ...

  12. Nuclear Fuel Cycle & Vulnerabilities (Technical Report) | SciTech...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Nuclear Fuel Cycle & Vulnerabilities Citation Details In-Document Search Title: Nuclear Fuel Cycle & Vulnerabilities You are accessing a document from the ...

  13. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites Prepared for U.S. ... Evaluation of Removing Used Nuclear Fuel from Shutdown Sites October 1, 2014 iii ...

  14. Fundamental aspects of nuclear reactor fuel elements: solutions...

    Office of Scientific and Technical Information (OSTI)

    Fundamental aspects of nuclear reactor fuel elements: solutions to problems Citation Details In-Document Search Title: Fundamental aspects of nuclear reactor fuel elements: ...

  15. Fundamental aspects of nuclear reactor fuel elements (Technical...

    Office of Scientific and Technical Information (OSTI)

    Fundamental aspects of nuclear reactor fuel elements Citation Details In-Document Search Title: Fundamental aspects of nuclear reactor fuel elements You are accessing a document ...

  16. EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    79: Spent Nuclear Fuel Management, Aiken, South Carolina EIS-0279: Spent Nuclear Fuel ... Carolina, including placing these materials in forms suitable for ultimate disposition. ...

  17. Energy Department Announces New Investment in Nuclear Fuel Storage...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Investment in Nuclear Fuel Storage Research Energy Department Announces New Investment in Nuclear Fuel Storage Research April 16, 2013 - 12:19pm Addthis NEWS MEDIA CONTACT (202)...

  18. Used Nuclear Fuel Loading and Structural Performance Under Normal...

    Office of Environmental Management (EM)

    Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport - Modeling, Simulation and Experimental Integration RD&D Plan Used Nuclear Fuel Loading and ...

  19. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    SciTech Connect (OSTI)

    Dayem, H.A.; Ostenak, C.A.; Gutmacher, R.G.; Kern, E.A.; Markin, J.T.; Martinez, D.P.; Thomas, C.C. Jr.

    1982-07-01

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank.

  20. Nuclear fuel elements having a composite cladding

    DOE Patents [OSTI]

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  1. CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

    DOE Patents [OSTI]

    Weems, S.J.

    1963-09-24

    >A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

  2. Spent Nuclear Fuel (SNF) Removal Campaign Plan

    SciTech Connect (OSTI)

    PAJUNEN, A.L.

    2000-08-07

    The overall operation of the Spent Nuclear Fuel Project will include fuel removal, sludge removal, debris removal, and deactivation transition activities. Figure 1-1 provides an overview of the current baseline operating schedule for project sub-systems, indicating that a majority of fuel removal activities are performed over an approximately three-and-one-half year time period. The purpose of this document is to describe the strategy for operating the fuel removal process systems. The campaign plan scope includes: (1) identifying a fuel selection sequence during fuel removal activities, (2) identifying MCOs that are subjected to extra testing (process validation) and monitoring, and (3) discussion of initial MCO loading and monitoring in the Canister Storage Building (CSB). The campaign plan is intended to integrate fuel selection requirements for handling special groups of fuel within the basin (e.g., single pass reactor fuel), process validation activities identified for process systems, and monitoring activities during storage.

  3. Pyrochemical Treatment of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    K. M. Goff; K. L. Howden; G. M. Teske; T. A. Johnson

    2005-10-01

    Over the last 10 years, pyrochemical treatment of spent nuclear fuel has progressed from demonstration activities to engineering-scale production operations. As part of the Advanced Fuel Cycle Initiative within the U.S. Department of Energys Office of Nuclear Energy, Science and Technology, pyrochemical treatment operations are being performed as part of the treatment of fuel from the Experimental Breeder Reactor II at the Idaho National Laboratory. Integral to these treatment operations are research and development activities that are focused on scaling further the technology, developing and implementing process improvements, qualifying the resulting high-level waste forms, and demonstrating the overall pyrochemical fuel cycle.

  4. Method for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

    A method is disclosed for shearing spent nuclear fuel assemblies of the type wherein a plurality of long metal tubes packed with ceramic fuel are supported in a spaced apart relationship within an outer metal shell or shroud which provides structural support to the assembly. Spent nuclear fuel assemblies are first compacted in a stepwise manner between specially designed gag-compactors and then sheared into short segments amenable to chemical processing by shear blades contoured to mate with the compacted surface of the fuel assembly.

  5. LWR nuclear fuel bundle data for use in fuel bundle handling...

    Office of Scientific and Technical Information (OSTI)

    LWR nuclear fuel bundle data for use in fuel bundle handling Citation Details In-Document Search Title: LWR nuclear fuel bundle data for use in fuel bundle handling You are ...

  6. International Nuclear Fuel Cycle Fact Book

    SciTech Connect (OSTI)

    Leigh, I.W.; Patridge, M.D.

    1991-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECN/NEA activities reports; not reflect any one single source but frequently represent a consolidation/combination of information.

  7. Annotated Bibliography for Drying Nuclear Fuel

    SciTech Connect (OSTI)

    Rebecca E. Smith

    2011-09-01

    Internationally, the nuclear industry is represented by both commercial utilities and research institutions. Over the past two decades many of these entities have had to relocate inventories of spent nuclear fuel from underwater storage to dry storage. These efforts were primarily prompted by two factors: insufficient storage capacity (potentially precipitated by an open-ended nuclear fuel cycle) or deteriorating quality of existing underwater facilities. The intent of developing this bibliography is to assess what issues associated with fuel drying have been identified, to consider where concerns have been satisfactorily addressed, and to recommend where additional research would offer the most value to the commercial industry and the U. S. Department of Energy.

  8. Issues related to EM management of DOE spent nuclear fuel

    SciTech Connect (OSTI)

    Abbott, D.G.; Abashian, M.S.; Chakraborti, S.; Roberson, K.; Meloin, J.M.

    1993-07-01

    This document is a summary of the important issues involved in managing spent nuclear fuel (SNF) owned by the Department of Energy (DOE). Issues related to civilian SNF activities are not discussed. DOE-owned SNF is stored primarily at the Hanford Site, Idaho National Engineering Laboratory (INEL), Savannah River Site (SRS), Oak Ridge National Laboratory (ORNL), and West Valley Demonstration Project. Smaller quantities of SNF are stored at Brookhaven National Laboratory, Sandia National Laboratories, and Los Alamos National Laboratory (LANL). There is a wide variety of fuel types, including both low and high enrichment fuels from weapons production, DOE reactors, research and development programs, naval programs, and universities. Most fuel is stored in pools associated with reactor or reprocessing facilities. Smaller quantities are in dry storage. Physical conditions of the fuel range from excellent to poor or severely damaged. An issue is defined as an important question that must be answered or decision that must be made on a topic or subject relevant to achieving the complimentary objectives of (a) storing SNF in compliance with applicable regulations and orders until it can be disposed, and (b) safely disposing of DOE`s SNF. The purpose of this document is to define the issues; no recommendations are made on resolutions. As DOE`s national SNF management program is implemented, a system of issues identification, documentation, tracking, and resolution will be implemented. This document is an initial effort at issues identification. The first section of this document is an overview of issues that are common to several or all DOE facilities that manage SNF. The common issues are organized according to specific aspects of spent fuel management. This is followed by discussions of management issues that apply specifically to individual DOE facilities. The last section provides literature references.

  9. Nuclear Fuels Storage & Transportation Planning Project Documents |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy Fuel Cycle Technologies » Nuclear Fuels Storage & Transportation Planning Project » Nuclear Fuels Storage & Transportation Planning Project Documents Nuclear Fuels Storage & Transportation Planning Project Documents October 1, 2014 Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among

  10. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  11. Apparatus for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S.; Metz, III, Curtis F.

    1980-01-01

    A method and apparatus are described for shearing spent nuclear fuel assemblies of the type comprising an array of fuel pins disposed within an outer metal shell or shroud. A spent fuel assembly is first compacted in a known manner and then incrementally sheared using fixed and movable shear blades having matched laterally projecting teeth which slidably intermesh to provide the desired shearing action. Incremental advancement of the fuel assembly after each shear cycle is limited to a distance corresponding to the lateral projection of the teeth to ensure fuel assembly breakup into small uniform segments which are amenable to remote chemical processing.

  12. Land and Water Use, CO2 Emissions, and Worker Radiological Exposure Factors for the Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    Brett W Carlsen; Brent W Dixon; Urairisa Pathanapirom; Eric Schneider; Bethany L. Smith; Timothy M. AUlt; Allen G. Croff; Steven L. Krahn

    2013-08-01

    The Department of Energy Office of Nuclear Energy’s Fuel Cycle Technologies program is preparing to evaluate several proposed nuclear fuel cycle options to help guide and prioritize Fuel Cycle Technology research and development. Metrics are being developed to assess performance against nine evaluation criteria that will be used to assess relevant impacts resulting from all phases of the fuel cycle. This report focuses on four specific environmental metrics. • land use • water use • CO2 emissions • radiological Dose to workers Impacts associated with the processes in the front-end of the nuclear fuel cycle, mining through enrichment and deconversion of DUF6 are summarized from FCRD-FCO-2012-000124, Revision 1. Impact estimates are developed within this report for the remaining phases of the nuclear fuel cycle. These phases include fuel fabrication, reactor construction and operations, fuel reprocessing, and storage, transport, and disposal of associated used fuel and radioactive wastes. Impact estimates for each of the phases of the nuclear fuel cycle are given as impact factors normalized per unit process throughput or output. These impact factors can then be re-scaled against the appropriate mass flows to provide estimates for a wide range of potential fuel cycles. A companion report, FCRD-FCO-2013-000213, applies the impact factors to estimate and provide a comparative evaluation of 40 fuel cycles under consideration relative to these four environmental metrics.

  13. Nuclear Fuel Cycle | Department of Energy

    Energy Savers [EERE]

    ... In a fuel fabrication plant great care is taken with the size and shape of processing ... Generation of electricity in a nuclear reactor is similar to a coal-fired steam station. The ...

  14. Electrochemical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    K. M. Goff; J. C. Wass; G. M. Teske

    2011-08-01

    As part of the Department of Energys Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

  15. Fuel cycle cost uncertainty from nuclear fuel cycle comparison

    SciTech Connect (OSTI)

    Li, J.; McNelis, D.; Yim, M.S.

    2013-07-01

    This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for the discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC.

  16. Fuel assembly for nuclear reactors

    DOE Patents [OSTI]

    Creagan, Robert J.; Frisch, Erling

    1977-01-01

    A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

  17. Nuclear Fuel Facts: Uranium | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Facts: Uranium Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Uranium is a silvery-white metallic chemical element in the periodic table, with atomic number 92. It is assigned the chemical symbol U. A uranium atom has 92 protons and 92 electrons, of which 6 are valence electrons. Uranium has the highest atomic weight (19 kg m) of all naturally occurring elements. Uranium occurs naturally in low concentrations in soil, rock and water, and is commercially extracted from uranium-bearing

  18. Composite construction for nuclear fuel containers

    DOE Patents [OSTI]

    Cheng, B. C.; Rosenbaum, H. S.; Armijo, J. S.

    1987-04-21

    Disclosed is an improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof. 1 fig.

  19. Composite construction for nuclear fuel containers

    DOE Patents [OSTI]

    Cheng, Bo-Ching [Fremont, CA; Rosenbaum, Herman S [Fremont, CA; Armijo, Joseph S [Saratoga, CA

    1987-01-01

    An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

  20. Fresh nuclear fuel measurements at Ukrainian nuclear power plants

    SciTech Connect (OSTI)

    Kuzminski, Jozef; Ewing, Tom; Dickman, Debbie; Gavrilyuk, Victor; Drapey, Sergey; Kirischuk, Vladimir; Strilchuk, Nikolay

    2009-01-01

    In 2005, the Provisions on Nuclear Material Measurement System was enacted in Ukraine as an important regulatory driver to support international obligations in nuclear safeguards and nonproliferation. It defines key provisions and requirements for material measurement and measurement control programs to ensure the quality and reliability of measurement data within the framework of the State MC&A System. Implementing the Provisions requires establishing a number of measurement techniques for both fresh and spent nuclear fuel for various types of Ukrainian reactors. Our first efforts focused on measurements of fresh nuclear fuel from a WWR-1000 power reactor.

  1. Pyrolytic carbon-coated nuclear fuel

    DOE Patents [OSTI]

    Lindemer, Terrence B.; Long, Jr., Ernest L.; Beatty, Ronald L.

    1978-01-01

    An improved nuclear fuel kernel having at least one pyrolytic carbon coating and a silicon carbon layer is provided in which extensive interaction of fission product lanthanides with the silicon carbon layer is avoided by providing sufficient UO.sub.2 to maintain the lanthanides as oxides during in-reactor use of said fuel.

  2. Spent Nuclear Fuel Project Technical Databook

    SciTech Connect (OSTI)

    Reilly, M.A.

    1998-10-23

    The Spent Nuclear Fuel (SNF) Project Technical Databook is developed for use as a common authoritative source of fuel behavior and material parameters in support of the Hanford SNF Project. The Technical Databook will be revised as necessary to add parameters as their Databook submittals become available.

  3. Dry Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    K. M. Goff; M. F. Simpson

    2009-09-01

    Dry (non-aqueous) separations technologies have been used for treatment of used nuclear fuel since the 1960s, and they are still being developed and demonstrated in many countries. Dry technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. Within the Department of Energys Advanced Fuel Cycle Initiative, an electrochemical process employing molten salts is being developed for recycle of fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. Much of the development of this technology is based on treatment of used Experimental Breeder Reactor II (EBR-II) fuel, which is metallic. Electrochemical treatment of the EBR-II fuel has been ongoing in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory since 1996. More than 3.8 metric tons of heavy metal of metallic fast reactor fuel have been treated using this technology. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including high-level waste work. A historic perspective on the background of dry processing will also be provided.

  4. Rack for storing spent nuclear fuel elements

    DOE Patents [OSTI]

    Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

    1978-06-20

    A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

  5. Utilization of Used Nuclear Fuel (UNF) in a Potential Future...

    Office of Scientific and Technical Information (OSTI)

    Utilization of Used Nuclear Fuel (UNF) in a Potential Future U.S. Fuel Cycle Scenario Citation Details In-Document Search Title: Utilization of Used Nuclear Fuel (UNF) in a ...

  6. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOE Patents [OSTI]

    Dickson, J.J.

    1963-09-24

    A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

  7. Vermont Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable." "- No data reported." "Notes: Totals may not equal sum of ...

  8. Kansas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable." "* Absolute percentage less than 0.05." "Notes: Totals may not ...

  9. Washington Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable." "* Absolute percentage less than 0.05." "- No data reported." ...

  10. Virginia Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable." "* Absolute percentage less than 0.05." "- No data reported." ...

  11. Iowa Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable." "Notes: Totals may not equal sum of components due to independent ...

  12. Wisconsin Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    components of the fuel; however, all Municipal Solid Waste summer capacity is classified as Renewable." "Notes: Totals may not equal sum of components due to independent ...

  13. THE ATTRACTIVENESS OF MATERIAS ASSOCIATED WITH THORIUM-BASED NUCLEAR FUEL CYCLES FOR PHWRS

    SciTech Connect (OSTI)

    Prichard, Andrew W.; Niehus, Mark T.; Collins, Brian A.; Bathke, Charles G.; Ebbinghaus, Bartley B.; Hase, Kevin R.; Sleaford, Brad W.; Robel, Martin; Smith, Brian W.

    2011-07-17

    This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with thorium based nuclear fuel cycles. Specifically, this paper examines a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of natural uranium/233U/thorium. This paper uses a PHWR fueled with natural uranium as a base fuel cycle, and then compares material attractiveness of fuel cycles that use 233U/thorium salted with natural uranium. The results include the material attractiveness of fuel at beginning of life (BoL), end of life (EoL), and the number of fuel assemblies required to collect a bare critical mass of plutonium or uranium. This study indicates what is required to render the uranium as having low utility for use in nuclear weapons; in addition, this study estimates the increased number of assemblies required to accumulate a bare critical mass of plutonium that has a higher utility for use in nuclear weapons. This approach identifies that some fuel cycles may be easier to implement the International Atomic Energy Agency (IAEA) safeguards approach and have a more effective safeguards by design outcome. For this study, approximately one year of fuel is required to be reprocessed to obtain one bare critical mass of plutonium. Nevertheless, the result of this paper suggests that all spent fuel needs to be rigorously safeguarded and provided with high levels of physical protection. This study was performed at the request of the United States Department of Energy /National Nuclear Security Administration (DOE/NNSA). The methodology and key findings will be presented.

  14. COMPLETION OF THE FIRST INTEGRATED SPENT NUCLEAR FUEL TRANSSHIPMENT/INTERIM STORAGE FACILITY IN NW RUSSIA

    SciTech Connect (OSTI)

    Dyer, R.S.; Barnes, E.; Snipes, R.L.; Hoeibraaten, S.; Gran, H.C.; Foshaug, E.; Godunov, V.

    2003-02-27

    Northwest and Far East Russia contain large quantities of unsecured spent nuclear fuel (SNF) from decommissioned submarines that potentially threaten the fragile environments of the surrounding Arctic and North Pacific regions. The majority of the SNF from the Russian Navy, including that from decommissioned nuclear submarines, is currently stored in on-shore and floating storage facilities. Some of the SNF is damaged and stored in an unstable condition. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Additional interim storage capacity is required. Most of the existing storage facilities being used in Northwest Russia do not meet health and safety, and physical security requirements. The United States and Norway are currently providing assistance to the Russian Federation (RF) in developing systems for managing these wastes. If these wastes are not properly managed, they could release significant concentrations of radioactivity to these sensitive environments and could become serious global environmental and physical security issues. There are currently three closely-linked trilateral cooperative projects: development of a prototype dual-purpose transport and storage cask for SNF, a cask transshipment interim storage facility, and a fuel drying and cask de-watering system. The prototype cask has been fabricated, successfully tested, and certified. Serial production is now underway in Russia. In addition, the U.S. and Russia are working together to improve the management strategy for nuclear submarine reactor compartments after SNF removal.

  15. Functionalized ultra-porous titania nanofiber membranes as nuclear waste separation and sequestration scaffolds for nuclear fuels recycle.

    SciTech Connect (OSTI)

    Liu, Haiqing; Bell, Nelson Simmons; Cipiti, Benjamin B.; Lewis, Tom Goslee,; Sava, Dorina Florentina; Nenoff, Tina Maria

    2012-09-01

    Advanced nuclear fuel cycle concept is interested in reducing separations to a simplified, one-step process if possible. This will benefit from the development of a one-step universal getter and sequestration material so as a simplified, universal waste form was proposed in this project. We have developed a technique combining a modified sol-gel chemistry and electrospinning for producing ultra-porous ceramic nanofiber membranes with controllable diameters and porous structures as the separation/sequestration materials. These ceramic nanofiber materials have been determined to have high porosity, permeability, loading capacity, and stability in extreme conditions. These porous fiber membranes were functionalized with silver nanoparticles and nanocrystal metal organic frameworks (MOFs) to introduce specific sites to capture gas species that are released during spent nuclear fuel reprocessing. Encapsulation into a durable waste form of ceramic composition was also demonstrated.

  16. Nuclear Fuels & Materials Spotlight Volume 4

    SciTech Connect (OSTI)

    I. J. van Rooyen,; T. M. Lillo; Y. Q. WU; P.A. Demkowicz; L. Scott; D.M. Scates; E. L. Reber; J. H. Jackson; J. A. Smith; D.L. Cottle; B.H. Rabin; M.R. Tonks; S.B. Biner; Y. Zhang; R.L. Williamson; S.R. Novascone; B.W. Spencer; J.D. Hales; D.R. Gaston; C.J. Permann; D. Anders; S.L. Hayes; P.C. Millett; D. Andersson; C. Stanek; R. Ali; S.L. Garrett; J.E. Daw; J.L. Rempe; J. Palmer; B. Tittmann; B. Reinhardt; G. Kohse; P. Ramuhali; H.T. Chien; T. Unruh; B.M. Chase; D.W. Nigg; G. Imel; J. T. Harris

    2014-04-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

  17. fuel | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    fuel Cheaper catalyst may lower fuel costs for hydrogen-powered cars Sandia National Laboratories post-doctoral fellow Stan Chou demonstrates the reaction of more efficiently catalyzing hydrogen. In this simulation, the color is from dye excited by light and generating electrons for the catalyst molybdenum disulfide to evolve hydrogen. ALBUQUERQUE, N.M. -Sandia

  18. Dry Transfer Systems for Used Nuclear Fuel

    SciTech Connect (OSTI)

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

    The potential need for a dry transfer system (DTS) to enable retrieval of used nuclear fuel (UNF) for inspection or repackaging will increase as the duration and quantity of fuel in dry storage increases. This report explores the uses for a DTS, identifies associated general functional requirements, and reviews existing and proposed systems that currently perform dry fuel transfers. The focus of this paper is on the need for a DTS to enable transfer of bare fuel assemblies. Dry transfer systems for UNF canisters are currently available and in use for transferring loaded canisters between the drying station and storage and transportation casks.

  19. President Reagan Calls for a National Spent Fuel Storage Facility...

    National Nuclear Security Administration (NNSA)

    nuclear energy policy that anticipates the establishment of a facility for the storage of high-level radioactive waste and lifts the ban on commercial reprocessing of nuclear fuel

  20. INERT-MATRIX FUEL: ACTINIDE ''BURINGIN'' AND DIRECT DISPOSAL...

    Office of Scientific and Technical Information (OSTI)

    Excess actinides result from the dismantlement of nuclear weapons (Pu) and the reprocessing of commercial spent nuclear fuel (mainly 241 Am, 244 Cm and 237 Np). In Europe, Canada ...

  1. Nuclear fuel particles and method of making nuclear fuel compacts therefrom

    DOE Patents [OSTI]

    DeVelasco, Rubin I.; Adams, Charles C.

    1991-01-01

    Methods for making nuclear fuel compacts exhibiting low heavy metal contamination and fewer defective coatings following compact fabrication from a mixture of hardenable binder, such as petroleum pitch, and nuclear fuel particles having multiple layer fission-product-retentive coatings, with the dense outermost layer of the fission-product-retentive coating being surrounded by a protective overcoating, e.g., pyrocarbon having a density between about 1 and 1.3 g/cm.sup.3. Such particles can be pre-compacted in molds under relatively high pressures and then combined with a fluid binder which is ultimately carbonized to produce carbonaceous nuclear fuel compacts having relatively high fuel loadings.

  2. Waste Stream Analyses for Nuclear Fuel Cycles

    SciTech Connect (OSTI)

    N. R. Soelberg

    2010-08-01

    A high-level study was performed in Fiscal Year 2009 for the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) Advanced Fuel Cycle Initiative (AFCI) to provide information for a range of nuclear fuel cycle options (Wigeland 2009). At that time, some fuel cycle options could not be adequately evaluated since they were not well defined and lacked sufficient information. As a result, five families of these fuel cycle options are being studied during Fiscal Year 2010 by the Systems Analysis Campaign for the DOE NE Fuel Cycle Research and Development (FCRD) program. The quality and completeness of data available to date for the fuel cycle options is insufficient to perform quantitative radioactive waste analyses using recommended metrics. This study has been limited thus far to qualitative analyses of waste streams from the candidate fuel cycle options, because quantitative data for wastes from the front end, fuel fabrication, reactor core structure, and used fuel for these options is generally not yet available.

  3. Nuclear Energy Systems Laboratory (NESL) / Transient Nuclear Fuels Testing

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Transient Nuclear Fuels Testing - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Energy Defense Waste Management Programs

  4. Fuel cycles for the 80's

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    Papers presented at the American Nuclear Society's topical meeting on the fuel cycle are summarized. Present progress and goals in the areas of fuel fabrication, fuel reprocessing, spent fuel storage, accountability, and safeguards are reported. Present governmental policies which affect the fuel cycle are also discussed. Individual presentations are processed for inclusion in the Energy Data Base.(DMC)

  5. Nuclear Fuel Cycle Options Catalog

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Nuclear Security Administration | (NNSA) Nuclear Facility Risk Reduction project moves forward at Y-12 December 02, 2011 OAK RIDGE, Tenn. - The Y-12 National Security Complex has received final approval for a $76 million project that aims to maintain decades-old equipment - some dating to World War II - until the site constructs a new facility to ensure that the nation has essential uranium processing capability long-term. The Nuclear Facility Risk Reduction (NFRR) project includes two Y-12

  6. Advanced LWR Nuclear Fuel Development

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    ... through LWRS program funding and industry cost-sharing. * Coordinate project development among research organizations associated with the U.S commercial nuclear industry, to the ...

  7. Method of monitoring stored nuclear fuel elements

    SciTech Connect (OSTI)

    Borloo, E.; Buergers, W.; Crutzen, S.; Vinche, C.

    1983-05-24

    To monitor a nuclear fuel element or fuel elements located in a store, e.g. a pond in a swimming pool reactor, the store is illuminated ultrasonically using one or more transducers transmitting ultrasonic signals in one or more predetermined directions to obtain an output which, because it depends on the number and relative location of the fuel elements in the store, and the structure of the store itself is distinctive to the fuel elements or elements stored therein. From this distinctive output is derived an identity unique to the stored fuel element or elements and a reference signal indicative of the whole structure when intact, the reference signal and identity being recorded. Subsequent ultrasonic testing of the store and its contents under identical operating conditions produces a signal which is compared to the recorded reference signal and if different therefrom reveals the occurrence of tampering with the store and/or the fuel element or elements.

  8. MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

    DOE Patents [OSTI]

    Barton, C.J.; Grimes, W.R.

    1960-01-01

    Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

  9. Systems Analysis of an Advanced Nuclear Fuel Cycle Based on a Modified UREX+3c Process

    SciTech Connect (OSTI)

    E. R. Johnson; R. E. Best

    2009-12-28

    The research described in this report was performed under a grant from the U.S. Department of Energy (DOE) to describe and compare the merits of two advanced alternative nuclear fuel cycles -- named by this study as the “UREX+3c fuel cycle” and the “Alternative Fuel Cycle” (AFC). Both fuel cycles were assumed to support 100 1,000 MWe light water reactor (LWR) nuclear power plants operating over the period 2020 through 2100, and the fast reactors (FRs) necessary to burn the plutonium and minor actinides generated by the LWRs. Reprocessing in both fuel cycles is assumed to be based on the UREX+3c process reported in earlier work by the DOE. Conceptually, the UREX+3c process provides nearly complete separation of the various components of spent nuclear fuel in order to enable recycle of reusable nuclear materials, and the storage, conversion, transmutation and/or disposal of other recovered components. Output of the process contains substantially all of the plutonium, which is recovered as a 5:1 uranium/plutonium mixture, in order to discourage plutonium diversion. Mixed oxide (MOX) fuel for recycle in LWRs is made using this 5:1 U/Pu mixture plus appropriate makeup uranium. A second process output contains all of the recovered uranium except the uranium in the 5:1 U/Pu mixture. The several other process outputs are various waste streams, including a stream of minor actinides that are stored until they are consumed in future FRs. For this study, the UREX+3c fuel cycle is assumed to recycle only the 5:1 U/Pu mixture to be used in LWR MOX fuel and to use depleted uranium (tails) for the makeup uranium. This fuel cycle is assumed not to use the recovered uranium output stream but to discard it instead. On the other hand, the AFC is assumed to recycle both the 5:1 U/Pu mixture and all of the recovered uranium. In this case, the recovered uranium is reenriched with the level of enrichment being determined by the amount of recovered plutonium and the combined amount

  10. International nuclear fuel cycle fact book

    SciTech Connect (OSTI)

    Leigh, I.W.

    1988-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source or information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  11. International Nuclear Fuel Cycle Fact Book

    SciTech Connect (OSTI)

    Leigh, I W; Mitchell, S J

    1990-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops, etc. The data listed do not reflect any one single source but frequently represent a consolidation/combination of information.

  12. International Nuclear Fuel Cycle Fact Book

    SciTech Connect (OSTI)

    Leigh, I.W.

    1992-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

  13. Spent nuclear fuel project integrated schedule plan

    SciTech Connect (OSTI)

    Squires, K.G.

    1995-03-06

    The Spent Nuclear Fuel Integrated Schedule Plan establishes the organizational responsibilities, rules for developing, maintain and status of the SNF integrated schedule, and an implementation plan for the integrated schedule. The mission of the SNFP on the Hanford site is to provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it to final disposition. This particularly involves K Basin fuel.

  14. Double-clad nuclear fuel safety rod

    DOE Patents [OSTI]

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  15. Arizona Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",3937,14.9,"31,200",27.9 "Coal","6,233",23.6,"43,644",39.1 "Hydro and Pumped ...

  16. Maryland Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,705",13.6,"13,994",32.1 "Coal","4,886",39.0,"23,668",54.3 "Hydro and Pumped ...

  17. Mississippi Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,251",8.0,"9,643",17.7 "Coal","2,526",16.1,"13,629",25.0 "Natural ...

  18. Pennsylvania Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","9,540",20.9,"77,828",33.9 "Coal","18,481",40.6,"110,369",48.0 "Hydro and Pumped ...

  19. Connecticut Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,103",25.4,"16,750",50.2 "Coal",564,6.8,"2,604",7.8 "Hydro and Pumped ...

  20. Illinois Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","11,441",25.9,"96,190",47.8 "Coal","15,551",35.2,"93,611",46.5 "Hydro and Pumped ...

  1. Iowa Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",601,4.1,"4,451",7.7 "Coal","6,956",47.7,"41,283",71.8 "Hydro and Pumped ...

  2. Louisiana Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,142",8.0,"18,639",18.1 "Coal","3,417",12.8,"23,924",23.3 "Hydro and Pumped ...

  3. Michigan Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,947",13.2,"29,625",26.6 "Coal","11,531",38.7,"65,604",58.8 "Hydro and Pumped ...

  4. Missouri Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,190",5.5,"8,996",9.7 "Coal","12,070",55.5,"75,047",81.3 "Hydro and Pumped ...

  5. Ohio Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,134",6.5,"15,805",11.0 "Coal","21,360",64.6,"117,828",82.1 "Hydro and Pumped ...

  6. Vermont Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",620,55.0,"4,782",72.2 "Hydro and Pumped Storage",324,28.7,"1,347",20.3 "Natural ...

  7. Wisconsin Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,584",8.9,"13,281",20.7 "Coal","8,063",45.2,"40,169",62.5 "Hydro and Pumped ...

  8. Arkansas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,835",11.5,"15,023",24.6 "Coal","4,535",28.4,"28,152",46.2 "Hydro and Pumped ...

  9. Minnesota Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,594",10.8,"13,478",25.1 "Coal","4,789",32.5,"28,083",52.3 "Hydro and Pumped ...

  10. Florida Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,924",6.6,"23,936",10.4 "Coal","9,975",16.9,"59,897",26.1 "Hydro and Pumped ...

  11. Texas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,966",4.6,"41,335",10.0 "Coal","22,335",20.6,"150,173",36.5 "Hydro and Pumped ...

  12. Virginia Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,501",14.5,"26,572",36.4 "Coal","5,868",24.3,"25,459",34.9 "Hydro and Pumped ...

  13. Massachusetts Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",685,5.0,"5,918",13.8 "Coal","1,669",12.2,"8,306",19.4 "Hydro and Pumped ...

  14. Tennessee Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,401",15.9,"27,739",33.7 "Coal","8,805",41.1,"43,670",53.0 "Hydro and Pumped ...

  15. Washington Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,097",3.6,"9,241",8.9 "Coal","1,340",4.4,"8,527",8.2 "Hydro and Pumped ...

  16. Nebraska Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,245",15.8,"11,054",30.2 "Coal","3,932",50.0,"23,363",63.8 "Hydro and Pumped ...

  17. Georgia Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,061",11.1,"33,512",24.4 "Coal","13,230",36.1,"73,298",53.3 "Hydro and Pumped ...

  18. California Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,390",6.5,"32,201",15.8 "Coal",374,0.6,"2,100",1.0 "Hydro and Pumped ...

  19. Kansas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,160",9.2,"9,556",19.9 "Coal","5,179",41.3,"32,505",67.8 "Hydro and Pumped ...

  20. Alabama Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","5,043",15.6,"37,941",24.9 "Coal","11,441",35.3,"63,050",41.4 "Hydro and Pumped ...

  1. Nuclear fuel elements made from nanophase materials

    DOE Patents [OSTI]

    Heubeck, N.B.

    1998-09-08

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

  2. Nuclear fuel elements made from nanophase materials

    DOE Patents [OSTI]

    Heubeck, Norman B.

    1998-01-01

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

  3. Computational Design of Advanced Nuclear Fuels

    SciTech Connect (OSTI)

    Savrasov, Sergey; Kotliar, Gabriel; Haule, Kristjan

    2014-06-03

    The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

  4. Nuclear Fuels Storage and Transportation Planning Project (NFST...

    Office of Environmental Management (EM)

    Fuels Storage and Transportation Planning Project (NFST) Program Status Nuclear Fuels Storage and Transportation Planning Project (NFST) Program Status Presentation made by Jeff ...

  5. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    DOE Patents [OSTI]

    Schluderberg, D.C.; Ryon, J.W.

    1962-05-01

    A fuel assembly is designed for use in a gas-suspension cooled nuclear fuel reactor. The coolant fluid is an inert gas such as nitrogen or helium with particles such as carbon suspended therein. The fuel assembly is contained within an elongated pressure vessel extending down into the reactor. The fuel portion is at the lower end of the vessel and is constructed of cylindrical segments through which the coolant passes. Turbulence promotors within the passageways maintain the particles in agitation to increase its ability to transfer heat away from the outer walls. Shielding sections and alternating passageways above the fueled portion limit the escape of radiation out of the top of the vessel. (AEC)

  6. Locking support for nuclear fuel assemblies

    DOE Patents [OSTI]

    Ledin, Eric

    1980-01-01

    A locking device for supporting and locking a nuclear fuel assembly within a cylindrical bore formed by a support plate, the locking device including a support and locking sleeve having upwardly extending fingers forming wedge shaped contact portions arranged for interaction between an annular tapered surface on the fuel assembly and the support plate bore as well as downwardly extending fingers having wedge shaped contact portions arranged for interaction between an annularly tapered surface on the support plate bore and the fuel assembly whereby the sleeve tends to support and lock the fuel assembly in place within the bore by its own weight while facilitating removal and/or replacement of the fuel assembly.

  7. The role of Z-pinch fusion transmutation of waste in the nuclear fuel cycle.

    SciTech Connect (OSTI)

    Smith, James Dean; Drennen, Thomas E.; Rochau, Gary Eugene; Martin, William Joseph; Kamery, William; Phruksarojanakun, Phiphat; Grady, Ryan; Cipiti, Benjamin B.; Wilson, Paul Philip Hood; Mehlhorn, Thomas Alan; Guild-Bingham, Avery; Tsvetkov, Pavel Valeryevich

    2007-10-01

    The resurgence of interest in reprocessing in the United States with the Global Nuclear Energy Partnership has led to a renewed look at technologies for transmuting nuclear waste. Sandia National Laboratories has been investigating the use of a Z-Pinch fusion driver to burn actinide waste in a sub-critical reactor. The baseline design has been modified to solve some of the engineering issues that were identified in the first year of work, including neutron damage and fuel heating. An on-line control feature was added to the reactor to maintain a constant neutron multiplication with time. The transmutation modeling effort has been optimized to produce more accurate results. In addition, more attention was focused on the integration of this burner option within the fuel cycle including an investigation of overall costs. This report presents the updated reactor design, which is able to burn 1320 kg of actinides per year while producing 3,000 MWth.

  8. President Reagan Calls for a National Spent Fuel Storage Facility |

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration | (NNSA) Reagan Calls for a National Spent Fuel Storage Facility President Reagan Calls for a National Spent Fuel Storage Facility Washington, DC The Reagan Administration announces a nuclear energy policy that anticipates the establishment of a facility for the storage of high-level radioactive waste and lifts the ban on commercial reprocessing of nuclear fuel

  9. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities, Sections 15-19

    SciTech Connect (OSTI)

    Schneider, K.J.

    1982-09-01

    Information is presented under the following section headings: fuel reprocessing; spent fuel and high-level and transuranic waste storage; spent fuel and high-level and transuranic waste disposal; low-level and intermediate-level waste disposal; and, transportation of radioactive materials in the nuclear fuel cycle. In each of the first three sections a description is given on the mainline process, effluent processing and waste management systems, plant layout, and alternative process schemes. Safety information and a summary are also included in each. The section on transport of radioactive materials includes information on the transportation of uranium ore, uranium ore concentrate, UF/sub 6/, PuO/sub 2/ powder, unirradiated uranium and mixed-oxide fuel assemblies, spent fuel, solidified high-level waste, contact-handled transuranic waste, remote-handled transuranic waste, and low and intermediate level nontransuranic waste. A glossary is included. (JGB)

  10. Categorization of Used Nuclear Fuel Inventory in Support of a...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    in Support of a Comprehensive National Nuclear Fuel Cycle Strategy Categorization of Used Nuclear ... of research, development, demonstration, and national security interests. ...

  11. Strategy for the Management and Disposal of Used Nuclear Fuel...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    waste from civilian nuclear power generation, defense, national security and other activities. Strategy for the Management and Disposal of Used Nuclear Fuel and High ...

  12. Activities Related to Storage of Spent Nuclear Fuel | Department...

    Office of Environmental Management (EM)

    Related to Storage of Spent Nuclear Fuel More Documents & Publications Nuclear Regulatory Commission Fifth National Report for the Joint Convention on the Safety of Spent...

  13. Comparison of spent nuclear fuel management alternatives

    SciTech Connect (OSTI)

    Beebe, C.L.; Caldwell, M.A,

    1996-09-01

    This paper reports the process an results of a trade study of spent nuclear fuel (SNF)management alternatives. The purpose of the trade study was to provide: (1) a summary of various SNF management alternatives, (2) an objective comparison of the various alternatives to facilitate the decision making process, and (3) documentation of trade study rational and the basis for decisions.

  14. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Bevard, Bruce Balkcom; Howard, Rob L

    2014-01-01

    Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

  15. Future nuclear fuel cycles: prospects and challenges

    SciTech Connect (OSTI)

    Boullis, Bernard

    2008-07-01

    Solvent extraction has played, from the early steps, a major role in the development of nuclear fuel cycle technologies, both in the front end and back end. Today's stakes in the field of energy enhance further than before the need for a sustainable management of nuclear materials. Recycling actinides appears as a main guideline, as much for saving resources as for minimizing the final waste impact, and many options can be considered. Strengthened by the important and outstanding performance of recent PUREX processing plants, solvent-extraction processes seem a privileged route to meet the new and challenging requirements of sustainable future nuclear systems. (author)

  16. Spent Nuclear Fuel Alternative Technology Decision Analysis

    SciTech Connect (OSTI)

    Shedrow, C.B.

    1999-11-29

    The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

  17. Next-generation nuclear fuel withstands high-temperature accident...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Next-generation nuclear fuel withstands high-temperature accident conditions IDAHO FALLS - A safer and more efficient nuclear fuel is on the horizon. A team of researchers at the ...

  18. Behavior of Spent Nuclear Fuel in Water Pool Storage

    Office of Scientific and Technical Information (OSTI)

    Behavior of Spent Nuclear Fuel in Water Pool Storage A. 0; Johnson, jr. , I ..: . Prepared ... I . , I BEHAVIOR OF SPENT NUCLEAR FUEL I N WATER POOL STORAGE by A. B. Johnson, J r . ...

  19. Behavior of spent nuclear fuel in water pool storage (Technical...

    Office of Scientific and Technical Information (OSTI)

    Behavior of spent nuclear fuel in water pool storage Citation Details In-Document Search Title: Behavior of spent nuclear fuel in water pool storage You are accessing a document ...

  20. RADIOACTIVE WASTE STREAMS FROM VARIOUS POTENTIAL NUCLEAR FUEL CYCLE OPTIONS

    SciTech Connect (OSTI)

    Nick Soelberg; Steve Piet

    2010-11-01

    Five fuel cycle options, about which little is known compared to more commonly known options, have been studied in the past year for the United States Department of Energy. These fuel cycle options, and their features relative to uranium-fueled light water reactor (LWR)-based fuel cycles, include: • Advanced once-through reactor concepts (Advanced Once-Through, or AOT) – intended for high uranium utilization and long reactor operating life, use depleted uranium in some cases, and avoid or minimize used fuel reprocessing • Fission-fusion hybrid (FFH) reactor concepts – potential variations are intended for high uranium or thorium utilization, produce fissile material for use in power generating reactors, or transmute transuranic (TRU) and some radioactive fission product (FP) isotopes • High temperature gas reactor (HTGR) concepts - intended for high uranium utilization, high reactor thermal efficiencies; they have unique fuel designs • Molten salt reactor (MSR) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, use on-line reprocessing of the used fuel, produce lesser amounts of long-lived, highly radiotoxic TRU elements, and avoid fuel assembly fabrication • Thorium/U-233 fueled LWR (Th/U-233) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, and produce lesser amounts of long-lived, highly radiotoxic TRU elements. These fuel cycle options could result in widely different types and amounts of used or spent fuels, spent reactor core materials, and waste streams from used fuel reprocessing, such as: • Highly radioactive, high-burnup used metal, oxide, or inert matrix U and/or Th fuels, clad in Zr, steel, or composite non-metal cladding or coatings • Spent radioactive-contaminated graphite, SiC, carbon-carbon-composite, metal, and Be reactor core materials • Li-Be-F salts containing U, TRU, Th, and fission products • Ranges of separated or un-separated activation

  1. Supply Security in Future Nuclear Fuel Markets

    SciTech Connect (OSTI)

    Seward, Amy M.; Wood, Thomas W.; Gitau, Ernest T.; Ford, Benjamin E.

    2013-11-18

    Previous PNNL work has shown the existing nuclear fuel markets to provide a high degree of supply security, including the ability to respond to supply disruptions that occur for technical and non-technical reasons. It is in the context of new reactor designs – that is, reactors likely to be licensed and market ready over the next several decades – that fuel supply security is most relevant. Whereas the fuel design and fabrication technology for existing reactors are well known, the construction of a new set of reactors could stress the ability of the existing market to provide adequate supply redundancy. This study shows this is unlikely to occur for at least thirty years, as most reactors likely to be built in the next three decades will be evolutions of current designs, with similar fuel designs to existing reactors.

  2. Nuclear Fuel Storage and Transportation Planning Project Overview |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy Fuel Storage and Transportation Planning Project Overview Nuclear Fuel Storage and Transportation Planning Project Overview Nuclear Fuel Storage and Transportation Planning Project Overview (956.77 KB) More Documents & Publications Section 180(c) Ad Hoc Working Group DOE Office of Nuclear Energy Transportation Plan Ad Hoc Working Group

  3. Nebraska Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,245",15.8,"11,054",30.2 "Coal","3,932",50.0,"23,363",63.8 "Hydro and Pumped Storage",278,3.5,"1,314",3.6

  4. Ohio Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,134",6.5,"15,805",11.0 "Coal","21,360",64.6,"117,828",82.1 "Hydro and Pumped Storage",101,0.3,429,0.3 "Natural

  5. Pennsylvania Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","9,540",20.9,"77,828",33.9 "Coal","18,481",40.6,"110,369",48.0 "Hydro and Pumped

  6. Alabama Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","5,043",15.6,"37,941",24.9 "Coal","11,441",35.3,"63,050",41.4 "Hydro and Pumped

  7. Louisiana Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,142",8.0,"18,639",18.1 "Coal","3,417",12.8,"23,924",23.3 "Hydro and Pumped Storage",192,0.7,"1,109",1.1

  8. Maryland Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,705",13.6,"13,994",32.1 "Coal","4,886",39.0,"23,668",54.3 "Hydro and Pumped Storage",590,4.7,"1,667",3.8

  9. Massachusetts Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, smmer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",685,5.0,"5,918",13.8 "Coal","1,669",12.2,"8,306",19.4 "Hydro and Pumped Storage","1,942",14.2,659,1.5 "Natural

  10. Michigan Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy Source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,947",13.2,"29,625",26.6 "Coal","11,531",38.7,"65,604",58.8 "Hydro and Pumped Storage","2,109",7.1,228,0.2

  11. Minnesota Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,594",10.8,"13,478",25.1 "Coal","4,789",32.5,"28,083",52.3 "Hydro and Pumped Storage",193,1.3,840,1.6 "Natural

  12. Mississippi Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,251",8.0,"9,643",17.7 "Coal","2,526",16.1,"13,629",25.0 "Natural Gas","11,640",74.2,"29,619",54.4

  13. Arizona Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (nw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear",3937,14.9,"31,200",27.9 "Coal","6,233",23.6,"43,644",39.1 "Hydro and Pumped Storage","2,937",11.1,"6,831",6.1

  14. Arkansas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,835",11.5,"15,023",24.6 "Coal","4,535",28.4,"28,152",46.2 "Hydro and Pumped

  15. California Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,390",6.5,"32,201",15.8 "Coal",374,0.6,"2,100",1.0 "Hydro and Pumped Storage","13,954",20.7,"33,260",16.3

  16. Connecticut Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","2,103",25.4,"16,750",50.2 "Coal",564,6.8,"2,604",7.8 "Hydro and Pumped Storage",151,1.8,400,1.2 "Natural

  17. Florida Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (nw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,924",6.6,"23,936",10.4 "Coal","9,975",16.9,"59,897",26.1 "Hydro and Pumped Storage",55,0.1,177,0.1 "Natural

  18. Georgia Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,061",11.1,"33,512",24.4 "Coal","13,230",36.1,"73,298",53.3 "Hydro and Pumped

  19. Illinois Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","11,441",25.9,"96,190",47.8 "Coal","15,551",35.2,"93,611",46.5 "Hydro and Pumped Storage",34,0.1,119,0.1 "Natural

  20. Tennessee Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","3,401",15.9,"27,739",33.7 "Coal","8,805",41.1,"43,670",53.0 "Hydro and Pumped

  1. Texas Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","4,966",4.6,"41,335",10.0 "Coal","22,335",20.6,"150,173",36.5 "Hydro and Pumped Storage",689,0.6,"1,262",0.3

  2. Nuclear power generation and fuel cycle report 1996

    SciTech Connect (OSTI)

    1996-10-01

    This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

  3. Missouri Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    total electric power industry, summer capacity and net generation, by energy source, 2010" "Primary energy source","Summer capacity (mw)","Share of State total (percent)","Net generation (thousand mwh)","Share of State total (percent)" "Nuclear","1,190",5.5,"8,996",9.7 "Coal","12,070",55.5,"75,047",81.3 "Hydro and Pumped Storage","1,221",5.6,"2,427",2.6

  4. Categorization of Used Nuclear Fuel Inventory in Support of a Comprehensive National Nuclear Fuel Cycle Strategy

    Broader source: Energy.gov [DOE]

    The Office of Nuclear Energy has conducted a technical review and assessment of the total current inventory [~70,150 MTHM as of 2011] of domestic discharged used nuclear fuel (UNF) and estimated an amount to be considered for retention in support of research, development, demonstration, and national security interests.

  5. Nuclear reactor fuel rod attachment system

    DOE Patents [OSTI]

    Not Available

    1980-09-17

    A reusable system is described for removably attaching a nuclear reactor fuel rod to a support member. A locking cap is secured to the fuel rod and a locking strip is fastened to the support member. The locking cap has two opposing fingers shaped to form a socket having a body portion. The locking strip has an extension shaped to rigidly attach to the socket's body portion. The locking cap's fingers are resiliently deflectable. For attachment, the locking cap is longitudinally pushed onto the locking strip causing the extension to temporarily deflect open the fingers to engage the socket's body portion. For removal, the process is reversed.

  6. Waste Management Planned for the Advanced Fuel Cycle Facility

    SciTech Connect (OSTI)

    Soelberg

    2007-09-01

    The U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program has been proposed to develop and employ advanced technologies to increase the proliferation resistance of spent nuclear fuels, recover and reuse nuclear fuel resources, and reduce the amount of wastes requiring permanent geological disposal. In the initial GNEP fuel cycle concept, spent nuclear fuel is to be reprocessed to separate re-useable transuranic elements and uranium from waste fission products, for fabricating new fuel for fast reactors. The separated wastes would be converted to robust waste forms for disposal. The Advanced Fuel Cycle Facility (AFCF) is proposed by DOE for developing and demonstrating spent nuclear fuel recycling technologies and systems. The AFCF will include capabilities for receiving and reprocessing spent fuel and fabricating new nuclear fuel from the reprocessed spent fuel. Reprocessing and fuel fabrication activities will generate a variety of radioactive and mixed waste streams. Some of these waste streams are unique and unprecedented. The GNEP vision challenges traditional U.S. radioactive waste policies and regulations. Product and waste streams have been identified during conceptual design. Waste treatment technologies have been proposed based on the characteristics of the waste streams and the expected requirements for the final waste forms. Results of AFCF operations will advance new technologies that will contribute to safe and economical commercial spent fuel reprocessing facilities needed to meet the GNEP vision. As conceptual design work and research and design continues, the waste management strategies for the AFCF are expected to also evolve.

  7. Method and apparatus for close packing of nuclear fuel assemblies

    DOE Patents [OSTI]

    Newman, Darrell F.

    1993-01-01

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  8. Method and apparatus for close packing of nuclear fuel assemblies

    DOE Patents [OSTI]

    Newman, D.F.

    1993-03-30

    The apparatus of the present invention is a plate of neutron absorbing material. The plate may have a releasable locking feature permitting the plate to be secured within a nuclear fuel assembly between nuclear fuel rods during storage or transportation then removed for further use or destruction. The method of the present invention has the step of placing a plate of neutron absorbing material between nuclear fuel rods within a nuclear fuel assembly, preferably between the two outermost columns of nuclear fuel rods. Additionally, the plate may be releasably locked in place.

  9. Pyrochemical processing of DOE spent nuclear fuel

    SciTech Connect (OSTI)

    Laidler, J.J.

    1995-02-01

    A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or {open_quotes}pyroprocessing,{close_quotes} provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (>99.9%) separation of transuranics. The resultant waste forms from the pyroprocess, are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

  10. Current Comparison of Advanced Nuclear Fuel Cycles

    SciTech Connect (OSTI)

    Steven Piet; Trond Bjornard; Brent Dixon; Robert Hill; Gretchen Matthern; David Shropshire

    2007-04-01

    This paper compares potential nuclear fuel cycle strategies – once-through, recycling in thermal reactors, sustained recycle with a mix of thermal and fast reactors, and sustained recycle with fast reactors. Initiation of recycle starts the draw-down of weapons-usable material and starts accruing improvements for geologic repositories and energy sustainability. It reduces the motivation to search for potential second geologic repository sites. Recycle in thermal-spectru

  11. NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

    DOE Patents [OSTI]

    Rasor, N.S.; Hirsch, R.L.

    1963-12-01

    The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

  12. POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

    DOE Patents [OSTI]

    Dwyer, O.E.

    1958-12-23

    A nuclear reactor system is described wherein the reactor is the type using a liquid metal fuel, such as a dispersion of fissile material in bismuth. The reactor is designed ln the form of a closed loop having a core sectlon and heat exchanger sections. The liquid fuel is clrculated through the loop undergoing flssion in the core section to produce heat energy and transferrlng this heat energy to secondary fluids in the heat exchanger sections. The fission in the core may be produced by a separate neutron source or by a selfsustained chain reaction of the liquid fuel present in the core section. Additional auxiliary heat exchangers are used in the system to convert water into steam which drives a turbine.

  13. Report on interim storage of spent nuclear fuel

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  14. Nuclear Fuels Storage & Transportation Planning Project | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Energy Nuclear Fuels Storage & Transportation Planning Project Nuclear Fuels Storage & Transportation Planning Project Independent Spent Fuel Storage Installation (ISFSI) at the shutdown Connecticut Yankee site. The ISFSI includes 40 multi-purpose canisters, within vertical concrete storage casks, containing 1019 used nuclear fuel assemblies [412.3 metric ton heavy metal (MTHM)] and 3 canisters of greater-than-class-C (GTCC) low-level radioactive waste. Photo courtesy of Connecticut

  15. Determination of plutonium in spent nuclear fuel using high resolution X-ray

    SciTech Connect (OSTI)

    McIntosh, Kathryn G.; Reilly, Sean D.; Havrilla, George J.

    2015-05-30

    Characterization of Pu is an essential aspect of safeguards operations at nuclear fuel reprocessing facilities. A novel analysis technique called hiRX (high resolution X-ray) has been developed for the direct measurement of Pu in spent nuclear fuel dissolver solutions. hiRX is based on monochromatic wavelength dispersive X-ray fluorescence (MWDXRF), which provides enhanced sensitivity and specificity compared with conventional XRF techniques. A breadboard setup of the hiRX instrument was calibrated using spiked surrogate spent fuel (SSF) standards prepared as dried residues. Samples of actual spent fuel were utilized to evaluate the performance of the hiRX. The direct detection of just 39 ng of Pu is demonstrated. Initial quantitative results, with error of 4–27% and precision of 2% relative standard deviation (RSD), were obtained for spent fuel samples. The limit of detection for Pu (100 s) within an excitation spot of 200 μm diameter was 375 pg. This study demonstrates the potential for the hiRX technique to be utilized for the rapid, accurate, and precise determination of Pu. Moreover, the results highlight the analytical capability of hiRX for other applications requiring sensitive and selective nondestructive analyses.

  16. Determination of plutonium in spent nuclear fuel using high resolution X-ray

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    McIntosh, Kathryn G.; Reilly, Sean D.; Havrilla, George J.

    2015-05-30

    Characterization of Pu is an essential aspect of safeguards operations at nuclear fuel reprocessing facilities. A novel analysis technique called hiRX (high resolution X-ray) has been developed for the direct measurement of Pu in spent nuclear fuel dissolver solutions. hiRX is based on monochromatic wavelength dispersive X-ray fluorescence (MWDXRF), which provides enhanced sensitivity and specificity compared with conventional XRF techniques. A breadboard setup of the hiRX instrument was calibrated using spiked surrogate spent fuel (SSF) standards prepared as dried residues. Samples of actual spent fuel were utilized to evaluate the performance of the hiRX. The direct detection of just 39more » ng of Pu is demonstrated. Initial quantitative results, with error of 4–27% and precision of 2% relative standard deviation (RSD), were obtained for spent fuel samples. The limit of detection for Pu (100 s) within an excitation spot of 200 μm diameter was 375 pg. This study demonstrates the potential for the hiRX technique to be utilized for the rapid, accurate, and precise determination of Pu. Moreover, the results highlight the analytical capability of hiRX for other applications requiring sensitive and selective nondestructive analyses.« less

  17. Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

    SciTech Connect (OSTI)

    Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.; Schwantes, Jon M.

    2014-01-01

    The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this

  18. Optimally moderated nuclear fission reactor and fuel source therefor

    DOE Patents [OSTI]

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  19. Transportation capabilities study of DOE-owned spent nuclear fuel

    SciTech Connect (OSTI)

    Clark, G.L.; Johnson, R.A.; Smith, R.W.; Abbott, D.G.; Tyacke, M.J.

    1994-10-01

    This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

  20. Dynamic Systems Analysis Report for Nuclear Fuel Recycle

    SciTech Connect (OSTI)

    Brent Dixon; Sonny Kim; David Shropshire; Steven Piet; Gretchen Matthern; Bill Halsey

    2008-12-01

    This report examines the time-dependent dynamics of transitioning from the current United States (U.S.) nuclear fuel cycle where used nuclear fuel is disposed in a repository to a closed fuel cycle where the used fuel is recycled and only fission products and waste are disposed. The report is intended to help inform policy developers, decision makers, and program managers of system-level options and constraints as they guide the formulation and implementation of advanced fuel cycle development and demonstration efforts and move toward deployment of nuclear fuel recycling infrastructure.

  1. Pyroprocess for processing spent nuclear fuel

    DOE Patents [OSTI]

    Miller, William E.; Tomczuk, Zygmunt

    2002-01-01

    This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

  2. Nuclear fuel rods along the sand dunes

    SciTech Connect (OSTI)

    Regan, M.B.

    1993-05-10

    Just north of the small town of Covert, Michigan, Consumers Power Co. officials and environmental activists are locked in a battle that marks a new phase in the nation's long-running struggle over nuclear power. The company's Palisades power plant reactor needs refueling. But the utility has no more room for the spent fuel rods it must place in its water-filled storage pool. So Consumers is taking advantage of a 1990 Nuclear Regulatory Commission rule that lets utilities store waste above ground without agency review. Palisades officials plan to transfer older radioactive fuel rods from its storage pool into concrete and steel silo-like casks on a site overlooking Lake Michigan. Over the next decade, nearly half of the nation's 109 operating nuclear plants will run out of space in water-filled storage pools and be forced to consider aboveground storage. The Palisades plant is causing a stir because it is the first to exploit the 1990 NRC rule, which doesn't require utilities to seek approval for waste-storage sites as long as the waste is stored in an approved container. Before 1990, five other utilities had received the agency O.K. for above-ground storage - but only after a lengthy and exhaustive analysis of each site.

  3. Quantifying the passive gamma signal from spent nuclear fuel in support of determining the plutonium content in spent nuclear fuel with nondestructive assay

    SciTech Connect (OSTI)

    Fensin, Michael L; Tobin, Steven J; Menlove, Howard O; Swinhoe, Martyn T

    2009-01-01

    The objective of safeguarding nuclear material is to deter diversions of significant quantities of nuclear materials by timely monitoring and detection. There are a variety of motivations for quantifying plutonium in spent fuel (SF), by means of nondestructive assay (NDA), in order to meet this goal. These motivations include the following: strengthening the capabilities of the International Atomic Energy Agencies ability to safeguard nuclear facilities, shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories. Many NDA techniques exist for measuring signatures from SF; however, no single NDA technique can, in isolation, quantify elemental plutonium in SF. A study has been undertaken to determine the best integrated combination of 13 NDA techniques for characterizing Pu mass in spent fuel. This paper focuses on the development of a passive gamma measurement system in support the spent fuel assay system. Gamma ray detection for fresh nuclear fuel focuses on gamma ray emissions that directly coincide with the actinides of interest to the assay. For example, the 186-keV gamma ray is generally used for {sup 235}U assay and the 384-keV complex is generally used for assaying plutonium. In spent nuclear fuel, these signatures cannot be detected as the Compton continuum created from the fission products dominates the signal in this energy range. For SF, the measured gamma signatures from key fission products ({sup 134}Cs, {sup 137}Cs, {sup 154}Eu) are used to ascertain burnup, cooling time, and fissile content information. In this paper the Monte Carlo modeling set-up for a passive gamma spent fuel assay system will be described. The set-up of the system includes a germanium detector and an ion chamber and will be used to gain passive gamma information that will be integrated into a system for determining Pu in SF. The passive gamma signal will be determined from a library of {approx} 100 assemblies that have been

  4. What to Expect When Readying to Move Spent Nuclear Fuel from...

    Energy Savers [EERE]

    Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move ...

  5. Spent Nuclear Fuel Vibration Integrity Study

    SciTech Connect (OSTI)

    Wang, Jy-An John; Wang, Hong; Jiang, Hao; Yan, Yong; Bevard, Bruce Balkcom

    2016-01-01

    The objective of this research is to collect dynamic experimental data on spent nuclear fuel (SNF) under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT), the hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL). The collected CIRFT data will be utilized to support ongoing spent fuel modeling activities, and support SNF transportation related licensing issues. Recent testing to understand the effects of hydride reorientation on SNF vibration integrity is also being evaluated. CIRFT results have provided insight into the fuel/clad system response to transportation related loads. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance, Fuel structure contributes to the SNF system stiffness, There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interaction, and SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous. Because of the non-homogeneous composite structure of the SNF system, finite element analyses (FEA) are needed to translate the global moment-curvature measurement into local stress-strain profiles. The detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained directly from a CIRFT system measurement. Therefore, detailed FEA is used to understand the global test response, and that data will also be presented.

  6. Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads

    SciTech Connect (OSTI)

    2013-07-01

    The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

  7. Nuclear reactor fuel rod attachment system

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA)

    1982-01-01

    A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

  8. Fuel handling system for a nuclear reactor

    DOE Patents [OSTI]

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-01-01

    A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

  9. Radiation chemistry in the reprocessing and recycling of spent nuclear fuels

    SciTech Connect (OSTI)

    Bruce J. Mincher

    2015-04-01

    The interaction of ionizing radiation with solvent extraction solutions results in the ionization, excitation, and decay to neutral radicals of mainly diluent molecules. These produced reactive species diffuse into the bulk solution to react with solvent extraction ligands. Ligand reactions often result in deleterious effects such as loss in ligand concentration or the production of decomposition products that may also be complexing agents. This often interferes with desired separations. The common radiolysis reactions and their potential effects on solvent extraction are reviewed here.

  10. Spent nuclear fuel project path forward preliminary safety evaluation

    SciTech Connect (OSTI)

    Brehm, J.R.; Crowe, R.D.; Siemer, J.M.; Wojdac, L.F.; Hosler, A.G.

    1995-03-01

    This preliminary safety evaluation (PSE) provides validation of the initial project design criteria for the Spent Nuclear Fuel Project (SNFP) Path Forward for removal of fuel from K Basins.

  11. Fuel assembly transfer basket for pool type nuclear reactor vessels

    DOE Patents [OSTI]

    Fanning, Alan W.; Ramsour, Nicholas L.

    1991-01-01

    A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.

  12. Gap Analysis to Support Extended Storage of Used Nuclear Fuel

    Office of Energy Efficiency and Renewable Energy (EERE)

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development...

  13. EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing...

    Energy Savers [EERE]

    Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing U.S.-Origin...

  14. Data validation and security for reprocessing.

    SciTech Connect (OSTI)

    Tolk, Keith Michael; Merkle, Peter Benedict; DurÔan, Felicia Angelica; Cipiti, Benjamin B.

    2008-10-01

    Next generation nuclear fuel cycle facilities will face strict requirements on security and safeguards of nuclear material. These requirements can result in expensive facilities. The purpose of this project was to investigate how to incorporate safeguards and security into one plant monitoring system early in the design process to take better advantage of all plant process data, to improve confidence in the operation of the plant, and to optimize costs. An existing reprocessing plant materials accountancy model was examined for use in evaluating integration of safeguards (both domestic and international) and security. International safeguards require independent, secure, and authenticated measurements for materials accountability--it may be best to design stand-alone systems in addition to domestic safeguards instrumentation to minimize impact on operations. In some cases, joint-use equipment may be appropriate. Existing domestic materials accountancy instrumentation can be used in conjunction with other monitoring equipment for plant security as well as through the use of material assurance indicators, a new metric for material control that is under development. Future efforts will take the results of this work to demonstrate integration on the reprocessing plant model.

  15. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    F. Habashi

    1998-06-26

    The DOE Spent Nuclear Fuel Disposal Container (SNF DC) supports the confinement and isolation of waste within the Engineered Barrier System of the Mined Geologic Disposal System (MGDS). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the access mains, and emplaced in emplacement drifts. The DOE Spent Nuclear Fuel Disposal Container provides long term confinement of DOE SNF waste, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The DOE SNF Disposal Containers provide containment of waste for a designated period of time, and limit radionuclide release thereafter. The disposal containers maintain the waste in a designated configuration, withstand maximum handling and rockfall loads, limit the individual waste canister temperatures after emplacement. The disposal containers also limit the introduction of moderator into the disposal container during the criticality control period, resist corrosion in the expected repository environment, and provide complete or limited containment of waste in the event of an accident. Multiple disposal container designs may be needed to accommodate the expected range of DOE Spent Nuclear Fuel. The disposal container will include outer and inner barrier walls and outer and inner barrier lids. Exterior labels will identify the disposal container and contents. Differing metal barriers will support the design philosophy of defense in depth. The use of materials with different failure mechanisms prevents a single mode failure from breaching the waste package. The corrosion-resistant inner barrier and inner barrier lid will be constructed of a high-nickel alloy and the corrosion-allowance outer barrier and outer barrier lid will be made of carbon steel. The DOE Spent Nuclear Fuel Disposal Containers interface with the emplacement drift environment by transferring heat from the waste to the external environment and by protecting

  16. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles

    SciTech Connect (OSTI)

    Finck, P.; Edelstein, N.; Allen, T.; Burns, C.; Chadwick, M.; Corradini, M.; Dixon, D.; Goff, M.; Laidler, J.; McCarthy, K.; Moyer, B.; Nash, K.; Navrotsky, A.; Oblozinsky, P.; Pasamehmetoglu, K.; Peterson, P.; Sackett, J.; Sickafus, K. E.; Tulenko, J.; Weber, W.; Morss, L.; Henry, G.

    2005-09-01

    The objective of this report is to identify new basic science that will be the foundation for advances in nuclear fuel-cycle technology in the near term, and for changing the nature of fuel cycles and of the nuclear energy industry in the long term. The goals are to enhance the development of nuclear energy, to maximize energy production in nuclear reactor parks, and to minimize radioactive wastes, other environmental impacts, and proliferation risks. The limitations of the once-through fuel cycle can be overcome by adopting a closed fuel cycle, in which the irradiated fuel is reprocessed and its components are separated into streams that are recycled into a reactor or disposed of in appropriate waste forms. The recycled fuel is irradiated in a reactor, where certain constituents are partially transmuted into heavier isotopes via neutron capture or into lighter isotopes via fission. Fast reactors are required to complete the transmutation of long-lived isotopes. Closed fuel cycles are encompassed by the Department of Energy?s Advanced Fuel Cycle Initiative (AFCI), to which basic scientific research can contribute. Two nuclear reactor system architectures can meet the AFCI objectives: a ?single-tier? system or a ?dual-tier? system. Both begin with light water reactors and incorporate fast reactors. The ?dual-tier? systems transmute some plutonium and neptunium in light water reactors and all remaining transuranic elements (TRUs) in a closed-cycle fast reactor. Basic science initiatives are needed in two broad areas: ? Near-term impacts that can enhance the development of either ?single-tier? or ?dual-tier? AFCI systems, primarily within the next 20 years, through basic research. Examples: Dissolution of spent fuel, separations of elements for TRU recycling and transmutation Design, synthesis, and testing of inert matrix nuclear fuels and non-oxide fuels Invention and development of accurate on-line monitoring systems for chemical and nuclear species in the nuclear

  17. International nuclear fuel cycle fact book. Revision 6

    SciTech Connect (OSTI)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1986-01-01

    The International Fuel Cycle Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2.

  18. Models Help Pinpoint Material for Better Nuclear Fuel Recycling

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Models Help Pinpoint Material for Better Nuclear Fuel Recycling Models Help Pinpoint Material for Better Nuclear Fuel Recycling Sifting 125,000 Candidates Yields Ideal Candidate for Xenon, Krypton Recovery June 13, 2016 Contact: Jon Bashor, jbashor@lbl.gov, +1 510.486.5849 SBMOF-1 illlustration A new material, dubbed SBMOF-1 illustrated here, could be used to separate xenon and krypton gases from the waste produced in recycling spent nuclear fuels using less energy than conventional methods. The

  19. Spent Nuclear Fuel project integrated safety management plan

    SciTech Connect (OSTI)

    Daschke, K.D.

    1996-09-17

    This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

  20. EM Safely and Efficiently Manages Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    EM's mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal in a geologic repository.

  1. Office of Used Nuclear Fuel Disposition International Program...

    Office of Environmental Management (EM)

    strategy that combines safe storage of spent nuclear fuel with expeditious progress toward siting ... involves information exchanges, expert panels, reviews, and training and education. ...

  2. EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory ...

  3. THE ATTRACTIVENESS OF MATERIALS IN ADVANCED NUCLEAR FUEL CYCLES FOR VARIOUS PROLIFERATION AND THEFT SCENARIOS

    SciTech Connect (OSTI)

    Bathke, C. G.; Ebbinghaus, Bartley B.; Collins, Brian A.; Sleaford, Brad W.; Hase, Kevin R.; Robel, Martin; Wallace, R. K.; Bradley, Keith S.; Ireland, J. R.; Jarvinen, G. D.; Johnson, M. W.; Prichard, Andrew W.; Smith, Brian W.

    2012-08-29

    We must anticipate that the day is approaching when details of nuclear weapons design and fabrication will become common knowledge. On that day we must be particularly certain that all special nuclear materials (SNM) are adequately accounted for and protected and that we have a clear understanding of the utility of nuclear materials to potential adversaries. To this end, this paper examines the attractiveness of materials mixtures containing SNM and alternate nuclear materials associated with the plutonium-uranium reduction extraction (Purex), uranium extraction (UREX), coextraction (COEX), thorium extraction (THOREX), and PYROX (an electrochemical refining method) reprocessing schemes. This paper provides a set of figures of merit for evaluating material attractiveness that covers a broad range of proliferant state and subnational group capabilities. The primary conclusion of this paper is that all fissile material must be rigorously safeguarded to detect diversion by a state and must be provided the highest levels of physical protection to prevent theft by subnational groups; no 'silver bullet' fuel cycle has been found that will permit the relaxation of current international safeguards or national physical security protection levels. The work reported herein has been performed at the request of the U.S. Department of Energy (DOE) and is based on the calculation of 'attractiveness levels' that are expressed in terms consistent with, but normally reserved for, the nuclear materials in DOE nuclear facilities. The methodology and findings are presented. Additionally, how these attractiveness levels relate to proliferation resistance and physical security is discussed.

  4. EA-1977: Acceptance and Disposition of Spent Nuclear Fuel Containing...

    Energy Savers [EERE]

    This spent nuclear fuel is composed of kernels containing thorium and U.S.-origin highly enriched uranium (HEU) embedded in small graphite spheres that were irradiated in nuclear ...

  5. CHARACTERIZATION OF HYDROGEN CONTENT IN ZIRCALOY-4 NUCLEAR FUEL CLADDING

    SciTech Connect (OSTI)

    Pfeif, E. A.; Mishra, B.; Olson, D. L.; Lasseigne, A. N.; Krzywosz, K.; Mader, E. V.

    2010-02-22

    Assessment of hydrogen uptake of underwater nuclear fuel clad and component materials will enable improved monitoring of fuel health. Zirconium alloys are used in nuclear reactors as fuel cladding, fuel channels, guide tubes and spacer grids, and are available for inspection in spent fuel pools. With increasing reactor exposure zirconium alloys experience hydrogen ingress due to neutron interactions and water-side corrosion that is not easily quantified without destructive hot cell examination. Contact and non-contact nondestructive techniques, using Seebeck coefficient measurements and low frequency impedance spectroscopy, to assess the hydrogen content and hydride formation within zircaloy 4 material that are submerged to simulate spent fuel pools are presented.

  6. Method and means of packaging nuclear fuel rods for handling

    DOE Patents [OSTI]

    Adam, Milton F.

    1979-01-01

    Nuclear fuel rods, especially spent nuclear fuel rods that may show physical distortion, are encased within a metallic enclosing structure by forming a tube about the fuel rod. The tube has previously been rolled to form an overlapping tubular structure and then unrolled and coiled about an axis perpendicular to the tube. The fuel rod is inserted into the tube as the rolled tube is removed from a coiled strip and allowed to reassume its tubular shape about the fuel rod. Rollers support the coiled strip in an open position as the coiled strip is uncoiled and allowed to roll about the fuel rod.

  7. A Critical Step Toward Sustainable Nuclear Fuel Disposal | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Energy A Critical Step Toward Sustainable Nuclear Fuel Disposal A Critical Step Toward Sustainable Nuclear Fuel Disposal January 26, 2012 - 2:30pm Addthis Secretary Chu Secretary Chu Former Secretary of Energy The Blue Ribbon Commission on America's Nuclear Future was formed at the direction of the President to conduct a comprehensive review of polices for managing the back end of the nuclear fuel cycle. If we are going to ensure that the United States remains at the forefront of nuclear

  8. Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

    SciTech Connect (OSTI)

    Phillips, Chris; Willis, William; Carter, Robert; Baker, Stephen

    2013-07-01

    Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions

  9. Overview of the United States spent nuclear fuel program

    SciTech Connect (OSTI)

    Hurt, W.L.

    1997-12-01

    As a result of the end of the Cold War, the mission of the US Department of Energy (DOE) has shifted from an emphasis on nuclear weapons development and production to an emphasis on the safe management and disposal of excess nuclear materials including spent nuclear fuel from both production and research reactors. Within the US, there are two groups managing spent nuclear fuel. Commercial nuclear power plants are managing their spent nuclear fuel at the individual reactor sites until the planned repository is opened. All other spent nuclear fuel, including research reactors, university reactors, naval reactors, and legacy material from the Cold War is managed by DOE. DOE`s mission is to safely and efficiently manage its spent nuclear fuel and prepare it for disposal. This mission involves correcting existing vulnerabilities in spent fuel storage; moving spent fuel from wet basins to dry storage; processing at-risk spent fuel; and preparing spent fuel in road-ready condition for repository disposal. Most of DOE`s spent nuclear fuel is stored in underwater basins (wet storage). Many of these basins are outdated, and spent fuel is to be removed and transferred to more modern basins or to new dry storage facilities. In 1995, DOE completed a complex-wide environmental impact analysis that resulted in spent fuel being sent to one of three principal DOE sites for interim storage (up to 40 years) prior to shipment to a repository. This regionalization by fuel type will allow for economies of scale yet minimize unnecessary transportation. This paper discusses the national SNF program, ultimate disposition of SNF, and the technical challenges that have yet to be resolved, namely, release rate testing, non-destructive assay, alternative treatments, drying, and chemical reactivity.

  10. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  11. Iodine Pathways and Off-Gas Stream Characteristics for Aqueous Reprocessing Plants – A Literature Survey and Assessment

    SciTech Connect (OSTI)

    R. T. Jubin; D. M. Strachan; N. R. Soelberg

    2013-09-01

    Used nuclear fuel is currently being reprocessed in only a few countries, notably France, England, Japan, and Russia. The need to control emissions of the gaseous radionuclides to the air during nuclear fuel reprocessing has already been reported for the entire plant. But since the gaseous radionuclides can partition to various different reprocessing off-gas streams, for example, from the head end, dissolver, vessel, cell, and melter, an understanding of each of these streams is critical. These off-gas streams have different flow rates and compositions and could have different gaseous radionuclide control requirements, depending on how the gaseous radionuclides partition. This report reviews the available literature to summarize specific engineering data on the flow rates, forms of the volatile radionuclides in off-gas streams, distributions of these radionuclides in these streams, and temperatures of these streams. This document contains an extensive bibliography of the information contained in the open literature.

  12. Basis for Section 3116 Determination for the Idaho Nuclear Technology...

    Office of Environmental Management (EM)

    ... Spent nuclear fuel reprocessing was curtailed in 1992, and the first-cycle extraction process wastes stored in the TFF were removed and solidified (by calcination) by February 1998 ...

  13. Plutonium partitioning in uranium and plutonium co-recovery system for fast reactor fuel recycling with enhanced nuclear proliferation resistance

    SciTech Connect (OSTI)

    Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo

    2013-07-01

    For enhancement of nuclear proliferation resistance, a 'co-processing' method for U and Pu co-recovery was studied. Two concepts, no U scrubbing and no Pu reduction partitioning, were employed to formulate two types of flow sheets by using a calculation code. Their process performance was demonstrated using radioactive solutions derived from an irradiated fast reactor fuel. These experimental results indicated that U and Pu were co-recovered in the U/Pu product, and the Pu content in the U/Pu product increased approximately 2.3 times regardless of using reductant. The proposed no U scrubbing and no Pu reductant flow sheet is applicable to fast reactor fuel reprocessing and enhances its resistance to nuclear proliferation. (authors)

  14. Spent nuclear fuel project quality assurance program plan

    SciTech Connect (OSTI)

    Lacey, R.E.

    1997-05-09

    This main body of this document describes how the requirements of 10 CFR 830.120 are met by the Spent Nuclear Fuel Project through implementation of WHC-SP-1131. Appendix A describes how the requirements of DOE/RW-0333P are met by the Spent Nuclear Fuel Project through implementation of specific policies, manuals, and procedures.

  15. Preliminary Evaluation of Removing Used Nuclear Fuel From Shutdown Sites

    Broader source: Energy.gov [DOE]

    In January 2013, the Department of Energy issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites.

  16. Separator assembly for use in spent nuclear fuel shipping cask

    DOE Patents [OSTI]

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  17. Belgium Nuclear Security Summit: Fact Sheet | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration | (NNSA) Nuclear Security Summit: Fact Sheet March 27, 2012 As one of the leaders in nuclear technology development, Belgium's nuclear program has covered all aspects of nuclear fuel cycle including reprocessing and operated a reprocessing plant between 1966 and 1974. Belgium signed the NPT in 1975 as a non-weapons state, but has retained a leading nuclear technology research center and derives over 50% of its energy from nuclear power using 7 power reactors. SCK-CEN is one of

  18. CHARACTERISTICS OF NEXT-GENERATION SPENT NUCLEAR FUEL (SNF) TRANSPORT AND STORAGE CASKS

    SciTech Connect (OSTI)

    Haire, M.J.; Forsberg, C.W.; Matveev, V.Z.; Shapovalov, V.I.

    2004-10-03

    The design of spent nuclear fuel (SNF) casks used in the present SNF disposition systems has evolved from early concepts about the nuclear fuel cycle. The reality today is much different from that envisioned by early nuclear scientists. Most SNF is placed in pool storage, awaiting reprocessing (as in Russia) or disposal at a geologic SNF repository (as in the United States). Very little transport of SNF occurs. This paper examines the requirements for SNF casks from today's perspective and attempts to answer this question: What type of SNF cask would be produced if we were to start over and design SNF casks based on today's requirements? The characteristics for a next-generation SNF cask system are examined and are found to be essentially the same in Russia and the United States. It appears that the new depleted uranium dioxide (DUO2)-steel cermet material will enable these requirements to be met. Depleted uranium (DU) is uranium in which a portion of the 235U isotope has been removed during a uranium enrichment process. The DUO2-steel cermet material is described. The United States and Russia are cooperating toward the development of a next-generation, dual-purpose, storage and transport SNF system.

  19. FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOE Patents [OSTI]

    Duffy, J.G. Jr.

    1961-05-30

    A lattice type fissionable fuel structure for a nuclear reactor is described. The fissionable material is formed into a plurality of rod-llke bodies with each encased in a fluid-tight jacket. A plurality of spaced longitudinal fins are mounted on the exterior and extend radially from each jacket, with a portion of the fins extending radially beyond the remainder of the fins. A collar of short length for each body is mounted on the extended fins for spacing the bodies, and adjacent bodies abut each other through these collars. Should distortion of the bodies take place, coilapse of the outer fins is limited by the shorter flns, thereby insuring some coolant flow at all times. (AEC)

  20. Electric heater for nuclear fuel rod simulators

    DOE Patents [OSTI]

    McCulloch, Reginald W.; Morgan, Jr., Chester S.; Dial, Ralph E.

    1982-01-01

    The present invention is directed to an electric cartridge-type heater for use as a simulator for a nuclear fuel pin in reactor studies. The heater comprises an elongated cylindrical housing containing a longitudinally extending helically wound heating element with the heating element radially inwardly separated from the housing. Crushed cold-pressed preforms of boron nitride electrically insulate the heating element from the housing while providing good thermal conductivity. Crushed cold-pressed preforms of magnesia or a magnesia-15 percent boron nitride mixture are disposed in the cavity of the helical heating element. The coefficient of thermal expansion of the magnesia or the magnesia-boron nitride mixture is higher than that of the boron nitride disposed about the heating element for urging the boron nitride radially outwardly against the housing during elevated temperatures to assure adequate thermal contact between the housing and the boron nitride.

  1. Spent nuclear fuel project technical databook

    SciTech Connect (OSTI)

    Reilly, M.A.

    1998-07-22

    The Spent Nuclear Fuel (SNF) project technical databook provides project-approved summary tables of selected parameters and derived physical quantities, with nominal design and safety basis values. It contains the parameters necessary for a complete documentation basis of the SNF Project technical and safety baseline. The databook is presented in two volumes. Volume 1 presents K Basins SNF related information. Volume 2 (not yet available) will present selected sludge and water information, as it relates to the sludge and water removal projects. The values, within this databook, shall be used as the foundation for analyses, modeling, assumptions, or other input to SNF project safety analyses or design. All analysis and modeling using a parameter available in this databook are required to use and cite the appropriate associated value, and document any changes to those values (i.e., analysis assumptions, equipment conditions, etc). Characterization and analysis efforts are ongoing to validate, or update these values.

  2. Technology Readiness Levels for Advanced Nuclear Fuels and Materials Development

    SciTech Connect (OSTI)

    Jon Carmack

    2014-01-01

    The Technology Readiness Level (TRL) process is used to quantitatively assess the maturity of a given technology. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Advanced nuclear fuels and materials development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the advanced fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Advanced Fuel Campaign (AFC).

  3. METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOE Patents [OSTI]

    Roake, W.E.; Evans, E.A.; Brite, D.W.

    1960-06-21

    A method of preparing a fuel element for a nuclear reactor is given in which an internally and externally cooled fuel element consisting of two coaxial tubes having a plurality of integral radial ribs extending between the tubes and containing a powdered fuel material is isostatically pressed to form external coolant channels and compact the powder simultaneously.

  4. Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin

    SciTech Connect (OSTI)

    Mickalonis, J. I.; Murphy, T. R.; Deible, R.

    2012-10-01

    Aluminum and stainless steel coupons were put into a nuclear fuel basin to monitor the effect of water chemistry on the corrosion of fuel cladding. These coupons have been monitored for over ten years. The corrosion and pitting data is being used to model the kinetics and estimate the damage that is occurring to the fuel cladding.

  5. World nuclear capacity and fuel cycle requirements, November 1993

    SciTech Connect (OSTI)

    Not Available

    1993-11-30

    This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy`s activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration`s annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment.

  6. Impact of Nuclear Energy Futures on Advanced Fuel Cycle Options

    SciTech Connect (OSTI)

    Dixon, B.W.; Piet, S.J.

    2004-10-03

    The Nuclear Waste Policy Act requires the Secretary of Energy to inform Congress before 2010 on the need for a second geologic repository for spent nuclear fuel. By that time, the spent fuel discharged from current commercial reactors will exceed the statutory limit of the first repository. There are several approaches to eliminate the need for another repository in this century. This paper presents a high-level analysis of these spent fuel management options in the context of a full range of possible nuclear energy futures. The analysis indicates the best option to implement varies depending on the nuclear energy future selected.

  7. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect (OSTI)

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  8. Nuclear fuel cycle facility accident analysis handbook

    SciTech Connect (OSTI)

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  9. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    SciTech Connect (OSTI)

    Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

    2011-06-30

    In nuclear resonance fluorescence (NRF) measurements, resonances are excited by an external photon beam leading to the emission of gamma rays with specific energies that are characteristic of the emitting isotope. NRF promises the unique capability of directly quantifying a specific isotope without the need for unfolding the combined responses of several fissile isotopes as is required in other measurement techniques. We have analyzed the potential of NRF as a non-destructive analysis technique for quantitative measurements of Pu isotopes in spent nuclear fuel (SNF). Given the low concentrations of 239Pu in SNF and its small integrated NRF cross sections, the main challenge in achieving precise and accurate measurements lies in accruing sufficient counting statistics in a reasonable measurement time. Using analytical modeling, and simulations with the radiation transport code MCNPX that has been experimentally tested recently, the backscatter and transmission methods were quantitatively studied for differing photon sources and radiation detector types. Resonant photon count rates and measurement times were estimated for a range of photon source and detection parameters, which were used to determine photon source and gamma-ray detector requirements. The results indicate that systems based on a bremsstrahlung source and present detector technology are not practical for high-precision measurements of 239Pu in SNF. Measurements that achieve the desired uncertainties within hour-long measurements will either require stronger resonances, which may be expressed by other Pu isotopes, or require quasi-monoenergetic photon sources with intensities that are approximately two orders of magnitude higher than those currently being designed or proposed.This work is part of a larger effort sponsored by the Next Generation Safeguards Initiative to develop an integrated instrument, comprised of individual NDA techniques with complementary features, that is fully capable of

  10. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    SciTech Connect (OSTI)

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  11. Fuel rod retention device for a nuclear reactor

    DOE Patents [OSTI]

    Hylton, Charles L.

    1984-01-01

    A device is described for supporting a nuclear fuel rod in a fuel rod assembly which allows the rod to be removed without disturbing other rods in the assembly. A fuel rod cap connects the rod to a bolt which is supported in the assembly end fitting by means of a locking assembly. The device is designed so that the bolt is held securely during normal reactor operation yet may be easily disengaged and the fuel rod removed when desired.

  12. Methods for making a porous nuclear fuel element

    DOE Patents [OSTI]

    Youchison, Dennis L; Williams, Brian E; Benander, Robert E

    2014-12-30

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  13. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  14. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    SciTech Connect (OSTI)

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

    2013-09-30

    This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

  15. Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Randall A. Alli; Steven L. Garrett

    2013-05-01

    A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

  16. Nuclear fuel alloys or mixtures and method of making thereof

    DOE Patents [OSTI]

    Mariani, Robert Dominick; Porter, Douglas Lloyd

    2016-04-05

    Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

  17. Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; G.L. Fredrickson; G.G. Galbreth; D. Vaden; M.D. Elliott; J.C. Price; E.M. Honeyfield; M.N. Patterson; L. A. Wurth

    2013-10-01

    Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.

  18. Pyroprocessing of fast flux test facility nuclear fuel

    SciTech Connect (OSTI)

    Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.; Galbreth, G.G.; Vaden, D.; Elliott, M.D.; Price, J.C.; Honeyfield, E.M.; Patterson, M.N.

    2013-07-01

    Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electro-refined uranium products exceeded 99%. (authors)

  19. Laser-based characterization of nuclear fuel plates

    SciTech Connect (OSTI)

    Smith, James A.; Cottle, Dave L.; Rabin, Barry H.

    2014-02-18

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  20. Laser-Based Characterization of Nuclear Fuel Plates

    SciTech Connect (OSTI)

    James A. Smith; David L. Cottle; Barry H. Rabin

    2013-07-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  1. Method of increasing the deterrent to proliferation of nuclear fuels

    DOE Patents [OSTI]

    Rampolla, Donald S.

    1982-01-01

    A process of recycling protactinium-231 to enhance the utilization of radioactively hot uranium-232 in nuclear fuel for the purpose of making both fresh and spent fuel more resistant to proliferation. The uranium-232 may be obtained by the irradiation of protactinium-231 which is normally found in the spent fuel rods of a thorium base nuclear reactor. The production of protactinium-231 and uranium-232 would be made possible by the use of the thorium uranium-233 fuel cycle in power reactors.

  2. What to Expect When Readying to Move Spent Nuclear Fuel from Commercial

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Nuclear Power Plants | Department of Energy What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants (397.39 KB) More Documents & Publications Nuclear Fuel Storage and Transportation Planning Project Overview Indiana Department of Homeland Security - NNPP Exercise

  3. Evaluation of Waste Arising from Future Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    Jubin, Robert Thomas; Taiwo, Temitope; Wigeland, Roald

    2015-01-01

    A comprehensive study was recently completed at the request of the US Department of Energy Office of Nuclear Energy (DOE-NE) to evaluate and screen nuclear fuel cycles. The final report was issued in October 2014. Uranium- and thorium-based fuel cycles were evaluated using both fast and thermal spectrum reactors. Once-through, limited-recycle, and continuous-recycle cases were considered. This study used nine evaluation criteria to identify promising fuel cycles. Nuclear waste management was one of the nine evaluation criteria. The waste generation criterion from this study is discussed herein.

  4. A Comparison Study of Various Nuclear Fuel Cycle Alternatives

    SciTech Connect (OSTI)

    Kwon, Eun-ha; Ko, Won-il

    2007-07-01

    As a nation develops its nuclear strategies, it must consider various aspects of nuclear energy such as sustainability, environmental-friendliness, proliferation-resistance, economics, technologies, and so on. Like all the policy decision, however, a nuclear fuel cycle option can not be superior in all aspects; the nation must identify its top priority and accordingly evaluate all the possible nuclear fuel cycle options. For such a purpose, this paper takes four different fuel cycle options that are likely adopted by the Korean government, considering the current status of nuclear power generation and the 3. Comprehensive Nuclear Energy Promotion Plan (CNEPP) - Once-through Cycle, DUPIC Recycle, Thermal Recycle and GEN-IV Recycle. The paper then evaluates each option in terms of resource utilization and waste generation. The analysis shows that the GEN-IV Recycle appears to be most competitive from these aspects. (authors)

  5. Fabrication of high exposure nuclear fuel pellets

    DOE Patents [OSTI]

    Frederickson, James R.

    1987-01-01

    A method is disclosed for making a fuel pellet for a nuclear reactor. A mixture is prepared of PuO.sub.2 and UO.sub.2 powders, where the mixture contains at least about 30% PuO.sub.2, and where at least about 12% of the Pu is the Pu.sup.240 isotope. To this mixture is added about 0.3 to about 5% of a binder having a melting point of at least about 250.degree. F. The mixture is pressed to form a slug and the slug is granulated. Up to about 4.7% of a lubricant having a melting point of at least about 330.degree. F. is added to the granulated slug. Both the binder and the lubricant are selected from a group consisting of polyvinyl carboxylate, polyvinyl alcohol, naturally occurring high molecular weight cellulosic polymers, chemically modified high molecular weight cellulosic polymers, and mixtures thereof. The mixture is pressed to form a pellet and the pellet is sintered.

  6. Disposition of ORNL's Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Turner, D. W.; DeMonia, B. C.; Horton, L. L.

    2002-02-26

    This paper describes the process of retrieving, repackaging, and preparing Oak Ridge spent nuclear fuel (SNF) for off-site disposition. The objective of the Oak Ridge SNF Project is to safely, reliably, and efficiently manage SNF that is stored on the Oak Ridge Reservation until it can be shipped off-site. The project required development of several unique processes and the design and fabrication of special equipment to enable the successful retrieval, transfer, and repackaging of Oak Ridge SNF. SNF was retrieved and transferred to a hot cell for repackaging. After retrieval of SNF packages, the storage positions were decontaminated and stainless steel liners were installed to resolve the vulnerability of water infiltration. Each repackaged SNF canister has been transferred from the hot cell back to dry storage until off-site shipments can be made. Three shipments of aluminum-clad SNF were made to the Savannah River Site (SRS), and five shipments of non-aluminum-clad SNF are planned to the Idaho National Engineering and Environmental Laboratory (INEEL). Through the integrated cooperation of several organizations including the U.S. Department of Energy (DOE), Bechtel Jacobs Company LLC (BJC), Oak Ridge National Laboratory (ORNL), and various subcontractors, preparations for the disposition of SNF in Oak Ridge have been performed in a safe and successful manner.

  7. New Design for an HLW Repository (for Spent Fuel and Waste from Reprocessing) in a Salt Formation in Germany - 12213

    SciTech Connect (OSTI)

    Bollingerfehr, Wilhelm; Filbert, Wolfgang; Lerch, Christian; Mueller-Hoeppe, Nina; Charlier, Frank

    2012-07-01

    In autumn 2010, after a 10-year moratorium, exploration was resumed in Gorleben, the potential site for a German HLW repository. At the same time, the Federal Government launched a two-year preliminary safety analysis to assess whether the salt dome at Gorleben is suitable to host all heat-generating radioactive waste generated by German NPPs based on the waste amounts expected at that time. The revised Atomic Energy Act of June 2011 now stipulates a gradual phase-out of nuclear energy production by 2022, which is 13 years earlier than expected in 2010. A repository design was developed which took into account an updated set of data on the amounts and types of expected heat-generating waste, the documented results of the exploration of the Gorleben salt dome, and the new 'Safety Requirements Governing the Final Disposal of Heat-Generating Radioactive Waste' of 30 September, 2010. The latter has a strong influence on the conceptual designs as it requires that retrievability of all waste containers is possible within the repository lifetime. One design considered that all waste containers will be disposed of in horizontal drifts of a geologic repository, while the other design considered that all waste containers will be disposed of in deep vertical boreholes. For both options (emplacement in drifts/emplacement in vertical boreholes), the respective design includes a selection of waste containers, the layout of drifts, respectively lined boreholes, a description of emplacement fields, and backfilling and sealing measures. The design results were described and displayed and the differences between the two main concepts were elaborated and discussed. For the first time in both repository designs the requirement was implemented to retrieve waste canisters during the operational phase. The measures to fulfill this requirement and eventually the consequences were highlighted. It was pointed out that there arises the need to keep transport- and storage casks in adequate

  8. 105 K East and 105 K West fuel transfer bay crane use strategy for spent nuclear fuel path-forward

    SciTech Connect (OSTI)

    Ard, K.E.

    1996-04-02

    The purpose of this document is to outline the K Basins 30 ton crane qualification strategy for use in the Spent Nuclear Fuel Project fuel relocation campaign.

  9. Development of FLUOREX Process as a Progressive LWR Reprocessing System

    SciTech Connect (OSTI)

    Sasahira, Akira; Kani, Yuko; Iino, Kenji; Hoshino, Kuniyoshi; Kawamura, Fumio

    2007-07-01

    New LWR fuel reprocessing technology named FLUOREX, the hybrid process of fluoride volatility and solvent extraction, proposed here is suitable for future thermal/fast reactors (coexistence) cycle. Recently we developed the flame reactor which was the main equipment to reduce volume of the spent nuclear fuel in order to down-size the PUREX like purification system. Newly constructed flame reactor, having 300 gU/h to 1 kgU/h production rate, was employed to examine the fluorination efficiency of UO{sub 2}. It was successfully demonstrated that we can control the residual amount of U within a range of 2 to 10 % by changing the excess amount of F{sub 2} gas during fluorination. This amount indicated that spent LWR fuel would be reduced to 8 to 15% by fluorination. (authors)

  10. Improved nuclear fuel assembly grid spacer

    DOE Patents [OSTI]

    Marshall, John; Kaplan, Samuel

    1977-01-01

    An improved fuel assembly grid spacer and method of retaining the basic fuel rod support elements in position within the fuel assembly containment channel. The improvement involves attachment of the grids to the hexagonal channel and of forming the basic fuel rod support element into a grid structure, which provides a design which is insensitive to potential channel distortion (ballooning) at high fluence levels. In addition the improved method eliminates problems associated with component fabrication and assembly.

  11. R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

    SciTech Connect (OSTI)

    Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina

    2006-07-01

    The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

  12. Energy Return on Investment from Recycling Nuclear Fuel

    SciTech Connect (OSTI)

    2011-08-17

    This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.

  13. Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; Chichester, David L.; Williams, Walter J.; Papaioannou, Glen C.; Smolinski, Andrew T.

    2015-09-10

    Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities,more » the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.« less

  14. Neutron radiography of irradiated nuclear fuel at Idaho National Laboratory

    SciTech Connect (OSTI)

    Craft, Aaron E.; Wachs, Daniel M.; Okuniewski, Maria A.; Chichester, David L.; Williams, Walter J.; Papaioannou, Glen C.; Smolinski, Andrew T.

    2015-09-10

    Neutron radiography of irradiated nuclear fuel provides more comprehensive information about the internal condition of irradiated nuclear fuel than any other non-destructive technique to date. Idaho National Laboratory (INL) has multiple nuclear fuels research and development programs that routinely evaluate irradiated fuels using neutron radiography. The Neutron Radiography reactor (NRAD) sits beneath a shielded hot cell facility where neutron radiography and other evaluation techniques are performed on these highly radioactive objects. The NRAD currently uses the foil-film transfer technique for imaging fuel that is time consuming but provides high spatial resolution. This study describes the NRAD and hot cell facilities, the current neutron radiography capabilities available at INL, planned upgrades to the neutron imaging systems, and new facilities being brought online at INL related to neutron imaging.

  15. Discovery sheds light on nuclear reactor fuel behavior during...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Discovery sheds light on nuclear reactor fuel behavior during a severe event By Angela Hardin * November 20, 2014 Tweet EmailPrint A new discovery about the atomic structure of...

  16. Nuclear reactor fuel element having improved heat transfer

    DOE Patents [OSTI]

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  17. The Economic, repository and proliferation implications of advanced nuclear fuel cycle

    SciTech Connect (OSTI)

    Deinert, Mark; Cady, K B

    2011-09-04

    The goal of this project was to compare the effects of recycling actinides using fast burner reactors, with recycle that would be done using inert matrix fuel burned in conventional light water reactors. In the fast reactor option, actinides from both spent light water and fast reactor fuel would be recycled. In the inert matrix fuel option, actinides from spent light water fuel would be recycled, but the spent inert matrix fuel would not be reprocessed. The comparison was done over a limited 100-year time horizon. The economic, repository and proliferation implications of these options all hinge on the composition of isotopic byproducts of power production. We took the perspective that back-end economics would be affected by the cost of spent fuel reprocessing (whether conventional uranium dioxide fuel, or fast reactor fuel), fuel manufacture, and ultimate disposal of high level waste in a Yucca Mountain like geological repository. Central to understanding these costs was determining the overall amount of reprocessing needed to implement a fast burner, or inert matrix fuel, recycle program. The total quantity of high level waste requiring geological disposal (along with its thermal output), and the cost of reprocessing were also analyzed. A major advantage of the inert matrix fuel option is that it could in principle be implemented using the existing fleet of commercial power reactors. A central finding of this project was that recycling actinides using an inert matrix fuel could achieve reductions in overall actinide production that are nearly very close to those that could be achieved by recycling the actinides using a fast burner reactor.

  18. Coolant mass flow equalizer for nuclear fuel

    DOE Patents [OSTI]

    Betten, Paul R.

    1978-01-01

    The coolant mass flow distribution in a liquid metal cooled reactor is enhanced by restricting flow in sub-channels defined in part by the peripheral fuel elements of a fuel assembly. This flow restriction, which results in more coolant flow in interior sub-channels, is achieved through the use of a corrugated liner positioned between the bundle of fuel elements and the inner wall of the fuel assembly coolant duct. The corrugated liner is expandable to accommodate irradiation induced growth of fuel assembly components.

  19. Means for supporting fuel elements in a nuclear reactor

    DOE Patents [OSTI]

    Andrews, Harry N.; Keller, Herbert W.

    1980-01-01

    A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

  20. Metallic Fast Reactor Fuel Fabrication for Global Nuclear Energy Partnership

    SciTech Connect (OSTI)

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter

    2009-07-01

    Fast reactors are once again being considered for nuclear power generation, in addition to transmutation of long-lived fission products resident in spent nuclear fuels. This re-consideration follows with intense developmental programs for both fuel and reactor design. One of the two leading candidates for next generation fast reactor fuel is metal alloys, resulting primarily from the successes achieved in the 1960s to early 1990s with both the experimental breeding reactor-II and the fast flux test facility. The goal of the current program is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional, fast-spectrum nuclear fuel while destroying recycled actinides, thereby closing the nuclear fuel cycle. In order to meet this goal, the program must develop efficient and safe fuel fabrication processes designed for remote operation. This paper provides an overview of advanced casting processes investigated in the past, and the development of a gaseous diffusion calculation that demonstrates how straightforward process parameter modification can mitigate the loss of volatile minor actinides in the metal alloy melt.

  1. Support grid for fuel elements in a nuclear reactor

    DOE Patents [OSTI]

    Finch, Lester M.

    1977-01-01

    A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates.

  2. Advanced nuclear fuel cycles - Main challenges and strategic choices

    SciTech Connect (OSTI)

    Le Biez, V.; Machiels, A.; Sowder, A.

    2013-07-01

    A graphical conceptual model of the uranium fuel cycles has been developed to capture the present, anticipated, and potential (future) nuclear fuel cycle elements. The once-through cycle and plutonium recycle in fast reactors represent two basic approaches that bound classical options for nuclear fuel cycles. Chief among these other options are mono-recycling of plutonium in thermal reactors and recycling of minor actinides in fast reactors. Mono-recycling of plutonium in thermal reactors offers modest savings in natural uranium, provides an alternative approach for present-day interim management of used fuel, and offers a potential bridging technology to development and deployment of future fuel cycles. In addition to breeder reactors' obvious fuel sustainability advantages, recycling of minor actinides in fast reactors offers an attractive concept for long-term management of the wastes, but its ultimate value is uncertain in view of the added complexity in doing so,. Ultimately, there are no simple choices for nuclear fuel cycle options, as the selection of a fuel cycle option must reflect strategic criteria and priorities that vary with national policy and market perspectives. For example, fuel cycle decision-making driven primarily by national strategic interests will likely favor energy security or proliferation resistance issues, whereas decisions driven primarily by commercial or market influences will focus on economic competitiveness.

  3. Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF)

    Broader source: Energy.gov [DOE]

    GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor...

  4. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-07-17

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  5. Mox fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    2001-05-15

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion. characteristics of the assembly.

  6. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  7. MOX fuel arrangement for nuclear core

    DOE Patents [OSTI]

    Kantrowitz, M.L.; Rosenstein, R.G.

    1998-10-13

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. 38 figs.

  8. Fuels

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing ... Heavy Duty Fuels DISI Combustion HCCISCCI Fundamentals Spray Combustion Modeling ...

  9. Electrochemical fluorination for processing of used nuclear fuel

    DOE Patents [OSTI]

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2016-07-05

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  10. South Carolina Nuclear Profile - All Fuels

    U.S. Energy Information Administration (EIA) Indexed Site

    solid waste, batteries, chemicals, hydrogen, pitch, purchased steam, sulfur, tire-derived fuel, and miscellaneous technologies." "Other Renewable: Wood, black liquor, other wood ...

  11. Technical bases for interim storage of spent nuclear fuel

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.

    1981-06-01

    The experience base for water storage of spent nuclear fuel has evolved since 1943. The technology base includes licensing documentation, standards, technology studies, pool operator experience, and documentation from public hearings. That base reflects a technology which is largely successful and mundane. It projects probable satisfactory water storage of spent water reactor fuel for several decades. Interim dry storage of spent water reactor fuel is not yet licensed in the US, but a data base and documentation have developed. There do not appear to be technological barriers to interim dry storage, based on demonstrations with irradiated fuel. Water storage will continue to be a part of spent fuel management at reactors. Whether dry storage becomes a prominent interim fuel management option depends on licensing and economic considerations. National policies will strongly influence how long the spent fuel remains in interim storage and what its final disposition will be.

  12. International nuclear fuel cycle fact book. Revision 4

    SciTech Connect (OSTI)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-03-01

    This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  13. International Nuclear Fuel Cycle Fact Book. Revision 5

    SciTech Connect (OSTI)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1985-01-01

    This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  14. New Nuclear Fuel Disposition Opportunities in a Changed World

    SciTech Connect (OSTI)

    Barrett, L.H.

    2006-07-01

    The world's economic, security, environmental, and technological situation has changed significantly in the last several years and these changes bring new opportunities for substantial policy improvements and redirections in the used nuclear fuel management arena. The passage of new energy legislation; the need for more US nuclear energy; growing state, national and international momentum for carbon emission and other air pollutant reductions; post September 11. Homeland Security threat reduction improvements; desires to improve global nuclear security; rapidly emerging needs for clean electricity supplies in developing countries; and the technological advancements in advanced fuel cycle technologies provide a substantial foundation for future enhancements and improvements in current used nuclear fuel management programs. Past progress, lessons learned, and new used fuel/waste management technological innovations coupled with current and future economic, security, and environmental issues can create new approaches that can help the Federal government meet its obligations while simultaneously addressing many of the difficult regional/state issues that have historically hindered progress. This paper will examine and integrate the synergy of these issues to explore options and discuss possible new opportunities in the vitally important area of spent fuel management and the entire back end of the nuclear fuel cycle. (authors)

  15. Spent nuclear fuel discharges from US reactors 1992

    SciTech Connect (OSTI)

    Not Available

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  16. Recent Sandia International Used Nuclear Fuel Management Collaborations

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    International Used Nuclear Fuel Management Collaborations - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Energy Defense Waste

  17. Multidimensional Multiphysics Simulation of Nuclear Fuel Behavior

    SciTech Connect (OSTI)

    R. L. Williamson; J. D. Hales; S. R. Novascone; M. R. Tonks; D. R. Gaston; C. J. Permann; D. Andrs; R. C. Martineau

    2012-04-01

    Important aspects of fuel rod behavior, for example pellet-clad mechanical interaction (PCMI), fuel fracture, oxide formation, non- axisymmetric cooling, and response to fuel manufacturing defects, are inherently multidimensional in addition to being complicated multiphysics problems. Many current modeling tools are strictly 2D axisymmetric or even 1.5D. This paper outlines the capabilities of a new fuel modeling tool able to analyze either 2D axisymmetric or fully 3D models. These capabilities include temperature-dependent thermal conductivity of fuel; swelling and densification; fuel creep; pellet fracture; fission gas release; cladding creep; irradiation growth; and gap mechanics (contact and gap heat transfer). The need for multiphysics, multidimensional modeling is then demonstrated through a discussion of results for a set of example problems. The first, a 10-pellet rodlet, demonstrates the viability of the solution method employed. This example highlights the effect of our smeared cracking model and also shows the multidimensional nature of discrete fuel pellet modeling. The second example relies on our multidimensional, multiphysics approach to analyze a missing pellet surface problem. The next example is the analysis of cesium diffusion in a TRISO fuel particle with defects. As a final example, we show a lower-length-scale simulation coupled to a continuum-scale simulation.

  18. Estimating Source Terms for Diverse Spent Nuclear Fuel Types

    SciTech Connect (OSTI)

    Brett Carlsen; Layne Pincock

    2004-11-01

    The U.S. Department of Energy (DOE) National Spent Nuclear Fuel Program is responsible for developing a defensible methodology for determining the radionuclide inventory for the DOE spent nuclear fuel (SNF) to be dispositioned at the proposed Monitored Geologic Repository at the Yucca Mountain Site. SNF owned by DOE includes diverse fuels from various experimental, research, and production reactors. These fuels currently reside at several DOE sites, universities, and foreign research reactor sites. Safe storage, transportation, and ultimate disposal of these fuels will require radiological source terms as inputs to safety analyses that support design and licensing of the necessary equipment and facilities. This paper summarizes the methodology developed for estimating radionuclide inventories associated with DOE-owned SNF. The results will support development of design and administrative controls to manage radiological risks and may later be used to demonstrate conformance with repository acceptance criteria.

  19. Eddy Current for Sizing Cracks in Canisters for Dry Storage of Used Nuclear Fuel

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Jones, Anthony M.; Pardini, Allan F.

    2014-01-01

    The storage of used nuclear fuel (UNF) in dry canister storage systems (DCSSs) at Independent Spent Fuel Storage Installations (ISFSI) sites is a temporary measure to accommodate UNF inventory until it can be reprocessed or transferred to a repository for permanent disposal. Policy uncertainty surrounding the long-term management of UNF indicates that DCSSs will need to store UNF for much longer periods than originally envisioned. Meanwhile, the structural and leak-tight integrity of DCSSs must not be compromised. The eddy current technique is presented as a potential tool for inspecting the outer surfaces of DCSS canisters for degradation, particularly atmospheric stress corrosion cracking (SCC). Results are presented that demonstrate that eddy current can detect flaws that cannot be detected reliably using standard visual techniques. In addition, simulations are performed to explore the best parameters of a pancake coil probe for sizing of SCC flaws in DCSS canisters and to identify features in frequency sweep curves that may potentially be useful for facilitating accurate depth sizing of atmospheric SCC flaws from eddy current measurements.

  20. Damaged Spent Nuclear Fuel at U.S. DOE Facilities Experience and Lessons Learned

    SciTech Connect (OSTI)

    Brett W. Carlsen; Eric Woolstenhulme; Roger McCormack

    2005-11-01

    From a handling perspective, any spent nuclear fuel (SNF) that has lost its original technical and functional design capabilities with regard to handling and confinement can be considered as damaged. Some SNF was damaged as a result of experimental activities and destructive examinations; incidents during packaging, handling, and transportation; or degradation that has occurred during storage. Some SNF was mechanically destroyed to protect proprietary SNF designs. Examples of damage to the SNF include failed cladding, failed fuel meat, sectioned test specimens, partially reprocessed SNFs, over-heated elements, dismantled assemblies, and assemblies with lifting fixtures removed. In spite of the challenges involved with handling and storage of damaged SNF, the SNF has been safely handled and stored for many years at DOE storage facilities. This report summarizes a variety of challenges encountered at DOE facilities during interim storage and handling operations along with strategies and solutions that are planned or were implemented to ameliorate those challenges. A discussion of proposed paths forward for moving damaged and nondamaged SNF from interim storage to final disposition in the geologic repository is also presented.

  1. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  2. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1998-01-01

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  3. Between-cycle laser system for depressurization and resealing of modified design nuclear fuel assemblies

    DOE Patents [OSTI]

    Bradley, John G.

    1982-01-01

    A laser beam is used to puncture fuel cladding for release of contained pressurized fission gas from plenum sections or irradiated fuel pins. Exhausted fission gases are collected and trapped for safe disposal. The laser beam, adjusted to welding mode, is subsequently used to reseal the puncture holes. The fuel assembly is returned to additional irradiation or, if at end of reactivity lifetime, is routed to reprocess. The fuel assembly design provides graded cladding lengths, by rows or arrays, such that the cladding of each component fuel element of the assembly is accessible to laser beam reception.

  4. Realization of the German Concept for Interim Storage of Spent Nuclear Fuel - Current Situation and Prospects

    SciTech Connect (OSTI)

    Thomauske, B. R.

    2003-02-25

    The German government has determined a phase out of nuclear power. With respect to the management of spent fuel it was decided to terminate transports to reprocessing plants by 2005 and to set up interim storage facilities on power plant sites. This paper gives an overview of the German concept for spent fuel management focused on the new on-site interim storage concept and the applied interim storage facilities. Since the end of the year 1998, the utilities have applied for permission of on-site interim storage in 13 storage facilities and 5 storage areas; one application for the interim storage facility Stade was withdrawn due to the planned final shut down of Stade nuclear power plant in autumn 2003. In 2001 and 2002, 3 on-site storage areas and 2 on-site storage facilities for spent fuel were licensed by the Federal Office for Radiation Protection (BfS). A main task in 2002 and 2003 has been the examination of the safety and security of the planned interim storage facilities and the verification of the licensing prerequisites. In the aftermath of September 11, 2001, BfS has also examined the attack with a big passenger airplane. Up to now, these aircraft crash analyses have been performed for three on-site interim storage facilities; the fundamental results will be presented. It is the objective of BfS to conclude the licensing procedures for the applied on-site interim storage facilities in 2003. With an assumed construction period for the storage buildings of about two years, the on-site interim storage facilities could then be available in the year 2005.

  5. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  6. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  7. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    DOE Patents [OSTI]

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  8. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    DOE Patents [OSTI]

    Travelli, Armando

    1988-01-01

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  9. Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

    DOE Patents [OSTI]

    Travelli, A.

    1985-10-25

    A flat or curved plate structure, to be used as fuel in a nuclear reactor, comprises elongated fissionable wires or strips embedded in a metallic continuous non-fissionable matrix plate. The wires or strips are made predominantly of a malleable uranium alloy, such as uranium silicide, uranium gallide or uranium germanide. The matrix plate is made predominantly of aluminum or an aluminum alloy. The wires or strips are located in a single row at the midsurface of the plate, parallel with one another and with the length dimension of the plate. The wires or strips are separated from each other, and from the surface of the plate, by sufficient thicknesses of matrix material, to provide structural integrity and effective fission product retention, under neutron irradiation. This construction makes it safely feasible to provide a high uranium density, so that the uranium enrichment with uranium 235 may be reduced below about 20%, to deter the reprocessing of the uranium for use in nuclear weapons.

  10. An evaluation of the nuclear fuel performance code BISON

    SciTech Connect (OSTI)

    Perez, D. M.; Williamson, R. L.; Novascone, S. R.; Larson, T. K.; Hales, J. D.; Spencer, B. W.; Pastore, G.

    2013-07-01

    BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behavior and is used to analyze either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods and other well known fuel performance codes. Results from several assessment cases are reported, with emphasis on fuel centerline temperatures at various stages of fuel life, fission gas release, and clad deformation during pellet clad mechanical interaction (PCMI). BISON comparisons to fuel centerline temperature measurements are very good at beginning of life and reasonable at high burnup. Although limited to date, fission gas release comparisons are very good. Comparisons of rod diameter following significant power ramping are also good and demonstrate BISON's unique ability to model discrete pellet behavior and accurately predict clad ridging from PCMI. (authors)

  11. Temperature measuring analysis of the nuclear reactor fuel assembly

    SciTech Connect (OSTI)

    Urban, F. E-mail: zdenko.zavodny@stuba.sk; Kučák, L. E-mail: zdenko.zavodny@stuba.sk; Bereznai, J. E-mail: zdenko.zavodny@stuba.sk; Závodný, Z. E-mail: zdenko.zavodny@stuba.sk; Muškát, P. E-mail: zdenko.zavodny@stuba.sk

    2014-08-06

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  12. International nuclear fuel cycle fact book: Revision 9

    SciTech Connect (OSTI)

    Leigh, I.W.

    1989-01-01

    The International Nuclear Fuel Cycle Fact Book has been compiled in an effort to provide current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. The Fact Book contains: national summaries in which a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; and international agencies in which a section for each of the international agencies which has significant fuel cycle involvement, and a listing of nuclear societies. The national summaries, in addition to the data described above, feature a small map for each country as well as some general information. The latter is presented from the perspective of the Fact Book user in the United States.

  13. International Nuclear Fuel Cycle Fact Book. Revision 12

    SciTech Connect (OSTI)

    Leigh, I.W.

    1992-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need exists costs for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book has been compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NMEA activities reports; and proceedings of conferences and workshops. The data listed typically do not reflect any single source but frequently represent a consolidation/combination of information.

  14. International nuclear fuel cycle fact book. [Contains glossary

    SciTech Connect (OSTI)

    Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

    1987-01-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained has been obtained from nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECD/NEA activities reports; proceedings of conferences and workshops; and so forth. Sources do not agree completely with each other, and the data listed herein does not reflect any one single source but frequently is a consolidation/combination of information. Lack of space as well as the intent and purpose of the Fact Book limit the given information to that pertaining to the Nuclear Fuel Cycle and to data considered of primary interest or most helpful to the majority of users.

  15. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    SciTech Connect (OSTI)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A. Ignatiev, V. V.; Subbotin, S. A. Tsibulskiy, V. F.

    2015-12-15

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  16. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOE Patents [OSTI]

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  17. Apparatus for injection casting metallic nuclear energy fuel rods

    DOE Patents [OSTI]

    Seidel, Bobby R.; Tracy, Donald B.; Griffiths, Vernon

    1991-01-01

    Molds for making metallic nuclear fuel rods are provided which present reduced risks to the environment by reducing radioactive waste. In one embodiment, the mold is consumable with the fuel rod, and in another embodiment, part of the mold can be re-used. Several molds can be arranged together in a cascaded manner, if desired, or several long cavities can be integrated in a monolithic multiple cavity re-usable mold.

  18. Double-clad nuclear-fuel safety rod

    DOE Patents [OSTI]

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  19. Evaluation of radioactivity release at Rokkasho reprocessing plant

    SciTech Connect (OSTI)

    Sugiyama, Hiroshi; Ishihara, Noriyuki; Maki, Akira

    2007-07-01

    JNFL have been conducting Active Test with spent fuels at Rokkasho Reprocessing Plant (RRP). In Active Test, the evaluation of radioactivity release to the environment (atmosphere and sea) was obtained. (authors)

  20. Railroad transportation of spent nuclear fuel

    SciTech Connect (OSTI)

    Wooden, D.G.

    1986-03-01

    This report documents a detailed analysis of rail operations that are important for assessing the risk of transporting high-level nuclear waste. The major emphasis of the discussion is towards ''general freight'' shipments of radioactive material. The purpose of this document is to provide a basis for selecting models and parameters that are appropriate for assessing the risk of rail transportation of nuclear waste.

  1. Preparation of nuclear fuel spheres by flotation-internal gelation

    DOE Patents [OSTI]

    Haas, P.A.; Fowler, V.L.; Lloyd, M.H.

    1984-12-21

    A simplified internal gelation process is claimed for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85/sup 0/C) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column. 3 figs.

  2. SUPPLEMENT ANALYSIS PROPOSED SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    SUPPLEMENT ANALYSIS PROPOSED SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL TO DOE NATIONAL LABORATORIES FOR RESEARCH AND DEVELOPMENT PURPOSES Office of Nuclear Energy U.S. DEPARTMENT OF ENERGY DECEMBER 2015 DOE/EIS-0203-SA-07 DOE/EIS-0250F-S-1-SA-02 Commercial Fuel Shipment SA DOE/EIS-0203-SA-07 December 2015 CONVERSION FACTORS Metric to English English to Metric Multiply by To get Multiply by To get Area Square kilometers 247.1 Acres Square kilometers 0.3861 Square miles Square meters 10.764 Square

  3. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites Prepared for U.S. Department of Energy Nuclear Fuels Storage and Transportation Planning Project Steven J. Maheras (PNNL) Ralph E. Best (PNNL) Steven B. Ross (PNNL) Kenneth A. Buxton (PNNL) Jeffery L. England (SRNL) Paul E. McConnell (SNL) Lawrence M. Massaro (FRA) Philip J. Jensen (PNNL) October 1, 2014 FCRD- NFST-2014-000091 Rev. 1 PNNL-22676 Rev. 4 DISCLAIMER This information was prepared as an account of work sponsored

  4. Preparation of nuclear fuel spheres by flotation-internal gelation

    DOE Patents [OSTI]

    Haas, Paul A.; Fowler, Victor L.; Lloyd, Milton H.

    1987-01-01

    A simplified internal gelation process for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85.degree. C.) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column.

  5. Sizing up nuclear fuel | Y-12 National Security Complex

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Sizing up nuclear fuel Sizing up nuclear fuel Posted: May 7, 2014 - 5:50pm | Y-12 Report | Volume 10, Issue 2 | 2014 When is small too small? Y-12 researchers analyzing samples of uranium materials would answer, "Almost never." In response to a request from the International Atomic Energy Agency, a team from Y-12's Analytical Chemistry and Development organizations evaluated natural and low-enriched-uranium samples - most no larger than 15 grams, or roughly comparable to a teaspoon of

  6. Technology Insights and Perspectives for Nuclear Fuel Cycle Concepts

    SciTech Connect (OSTI)

    S. Bays; S. Piet; N. Soelberg; M. Lineberry; B. Dixon

    2010-09-01

    The following report provides a rich resource of information for exploring fuel cycle characteristics. The most noteworthy trends can be traced back to the utilization efficiency of natural uranium resources. By definition, complete uranium utilization occurs only when all of the natural uranium resource can be introduced into the nuclear reactor long enough for all of it to undergo fission. Achieving near complete uranium utilization requires technologies that can achieve full recycle or at least nearly full recycle of the initial natural uranium consumed from the Earth. Greater than 99% of all natural uranium is fertile, and thus is not conducive to fission. This fact requires the fuel cycle to convert large quantities of non-fissile material into fissile transuranics. Step increases in waste benefits are closely related to the step increase in uranium utilization going from non-breeding fuel cycles to breeding fuel cycles. The amount of mass requiring a disposal path is tightly coupled to the quantity of actinides in the waste stream. Complete uranium utilization by definition means that zero (practically, near zero) actinide mass is present in the waste stream. Therefore, fuel cycles with complete (uranium and transuranic) recycle discharge predominately fission products with some actinide process losses. Fuel cycles without complete recycle discharge a much more massive waste stream because only a fraction of the initial actinide mass is burned prior to disposal. In a nuclear growth scenario, the relevant acceptable frequency for core damage events in nuclear reactors is inversely proportional to the number of reactors deployed in a fuel cycle. For ten times the reactors in a fleet, it should be expected that the fleet-average core damage frequency be decreased by a factor of ten. The relevant proliferation resistance of a fuel cycle system is enhanced with: decreasing reliance on domestic fuel cycle services, decreasing adaptability for technology misuse

  7. Quantitative assessment of proposals on assurance of nuclear fuel supply

    SciTech Connect (OSTI)

    Tanaka, T.; Kuno, Y.; Tanaka, S.

    2013-07-01

    The assurance of nuclear fuel supply has the potential to contribute to balancing peaceful use of nuclear power and nuclear nonproliferation. 5 proposals which provide the backup supply of the enrichment service in case of supply disruption, are investigated in this study. We investigated the 20 NPT countries which are non-nuclear-weapon states and possess operable commercial LWRs in October 2012 as potential participants for each proposal. As a result of literature researching, we have extracted factors that can be considered as important for a country to participate or not participate in the assurance of nuclear fuel supply. Then we have computed incentive and disincentive parameters for each country. The results show that the participation expectancy decreases in the order of IAEA Fuel Bank proposal, Russian LEU Reserve proposal, AFS proposal, WNA proposal and 6-Country proposal. The 'IAEA fuel bank proposal' would be triggered in case of the supply disruption which cannot be solved by the market mechanism and bilateral agreements.

  8. Nuclear waste actinides as fissile fuel in hybrid blankets

    SciTech Connect (OSTI)

    Sahin, S.; Al-Kusayer, T.A.

    1983-12-01

    The widespread use of the present LWRs produces substantial quantities of nuclear waste materials. Among those, actinide nuclear waste poses a serious problem of stockage because the associated half life times for actinides is measured in terms of geological time periods (several millions of years) so that no waste disposal guarantee over such time intervals can be given, except for space disposal. On the other hand, these nuclear waste actinides are very good fissionable materials for high energetic (D,T) fusion neutrons. It is therefore worthwhile to investigate their quality as potential nuclear fuel in hybrid blankets. The present study investigates the neutronic performance of hybrid blankets containing Np/sup 237/ and Cm/sup 244/ as fissile materials. The isotopic composition of Americium has been adjusted to the spent fuel isotope composition of a LWR. The geometrical design has been made, according to the AYMAN fussion-fission (hybrid) experimental facility, now in the very early phase of planning.

  9. Manufacture of bonded-particle nuclear fuel composites

    DOE Patents [OSTI]

    Stradley, J.G.; Sease, J.D.

    1973-10-01

    A preselected volume of nuclear fuel particles are placed in a cylindrical mold cavity followed by a solid pellet of resin--carbon matrix material of preselected volume. The mold is heated to liquefy the pellet and the liquefied matrix forced throughout the interstices of the fuel particles by advancing a piston into the mold cavity. Excess matrix is permitted to escape through a vent hole in the end of the mold opposite to that end where the pellet was originally disposed. After the matrix is resolidified by cooling, the resultant fuel composite is removed from the mold and the resin component of the matrix carbonized. (Official Gazette)

  10. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    SciTech Connect (OSTI)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

    The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

  11. EIS-0218: Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    This study analyzes the potential environmental impacts of adopting a policy to manage foreign research reactor spent nuclear fuel containing uranium enriched in the United States. In particular, the study examines the comparative impacts of several alternative approaches to managing the spent fuel.

  12. Nuclear fuel elements and method of making same

    DOE Patents [OSTI]

    Schweitzer, Donald G.

    1992-01-01

    A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

  13. Nuclear Fuel Cycle Reasoner: PNNL FY13 Report

    SciTech Connect (OSTI)

    Hohimer, Ryan E.; Strasburg, Jana D.

    2013-09-30

    In Fiscal Year 2012 (FY12) PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In Fiscal Year 2013 (FY13) the SNAP demonstration was enhanced with respect to query and navigation usability issues.

  14. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    SciTech Connect (OSTI)

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

  15. Systems for the Intermodal Routing of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Peterson, Steven K; Liu, Cheng

    2015-01-01

    The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable system for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of

  16. Zone approaches to international safeguards of a nuclear fuel cycle

    SciTech Connect (OSTI)

    Fishbone, L.G.; Higinbotham, W.A.

    1986-01-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the results of safeguards verifications for the individual facilities within it. We have examined safeguards approaches for a state nuclear fuel cycle that take into account the existence of all of the nuclear facilities in the state. We have focussed on the fresh-fuel zone of an advanced nuclear fuel cycle, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. The intention is to develop an approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the zone approach and for some reasonable intermediate safeguards approaches. Technical effectiveness, in these cases, means an estimate of the assurance that all nuclear material has been accounted for.

  17. fuel cells | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    fuel cells Sandia National Laboratories wins TechConnect Innovation Award NNSA's Sandia National Laboratories received an Innovation Award from global technology outreach and development organization TechConnect for the development of prototype Hydrocarbon Membranes for Energy and Water Electrochemical Systems. Sandia's submission placed in the top 15... Bay Area national labs team to tackle long-standing automotive hydrogen storage challenge Sandia National Laboratories chemist Mark Allendorf,

  18. Modeling Deep Burn TRISO Particle Nuclear Fuel

    SciTech Connect (OSTI)

    Besmann, Theodore M [ORNL; Stoller, Roger E [ORNL; Samolyuk, German D [ORNL; Schuck, Paul C [ORNL; Rudin, Sven [Los Alamos National Laboratory (LANL); Wills, John [Los Alamos National Laboratory (LANL); Wirth, Brian D. [University of California, Berkeley; Kim, Sungtae [University of Wisconsin, Madison; Morgan, Dane [University of Wisconsin, Madison; Szlufarska, Izabela [University of Wisconsin, Madison

    2012-01-01

    Under the DOE Deep Burn program TRISO fuel is being investigated as a fuel form for consuming plutonium and minor actinides, and for greater efficiency in uranium utilization. The result will thus be to drive TRISO particulate fuel to very high burn-ups. In the current effort the various phenomena in the TRISO particle are being modeled using a variety of techniques. The chemical behavior is being treated utilizing thermochemical analysis to identify phase formation/transformation and chemical activities in the particle, including kernel migration. First principles calculations are being used to investigate the critical issue of fission product palladium attack on the SiC coating layer. Density functional theory is being used to understand fission product diffusion within the plutonia oxide kernel. Kinetic Monte Carlo techniques are shedding light on transport of fission products, most notably silver, through the carbon and SiC coating layers. The diffusion of fission products through an alternative coating layer, ZrC, is being assessed via DFT methods. Finally, a multiscale approach is being used to understand thermal transport, including the effect of radiation damage induced defects, in a model SiC material.

  19. LWR spent fuel reduction by the removal of U and the compact storage of Pu with FP for long-term nuclear sustainability

    SciTech Connect (OSTI)

    Fukasawa, T.; Hoshino, K.; Takano, M.; Sato, S.; Shimazu, Y.

    2013-07-01

    Fast breeder reactors (FBR) nuclear fuel cycle is needed for long-term nuclear sustainability while preventing global warming and maximum utilizing the limited uranium (U) resources. The 'Framework for Nuclear Energy Policy' by the Japanese government on October 2005 stated that commercial FBR deployment will start around 2050 under its suitable conditions by the successive replacement of light water reactors (LWR) to FBR. Even after Fukushima Daiichi Nuclear Power Plant accident which made Japanese tendency slow down the nuclear power generation activities, Japan should have various options for energy resources including nuclear, and also consider the delay of FBR deployment and increase of LWR spent fuel (LWR-SF) storage amounts. As plutonium (Pu) for FBR deployment will be supplied from LWR-SF reprocessing and Japan will not possess surplus Pu, the authors have developed the flexible fuel cycle initiative (FFCI) for the transition from LWR to FBR. The FFCI system is based on the possibility to stored recycled materials (U, Pu)temporarily for a suitable period according to the FBR deployment rate to control the Pu demand/supply balance. This FFCI system is also effective after the Fukushima accident for the reduction of LWR-SF and future LWR-to-FBR transition. (authors)

  20. Interim report spent nuclear fuel retrieval system fuel handling development testing

    SciTech Connect (OSTI)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.