Sample records for reactor spent fuel

  1. Spent nuclear fuel discharges from U.S. reactors 1994

    SciTech Connect (OSTI)

    NONE

    1996-02-01T23:59:59.000Z

    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.

  2. Reactor-specific spent fuel discharge projections, 1987-2020

    SciTech Connect (OSTI)

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01T23:59:59.000Z

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.

  3. Reactor-specific spent fuel discharge projections: 1985 to 2020

    SciTech Connect (OSTI)

    Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

    1986-09-01T23:59:59.000Z

    The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel.

  4. Development of Technical Nuclear Forensics for Spent Research Reactor Fuel

    E-Print Network [OSTI]

    Sternat, Matthew Ryan 1982-

    2012-11-20T23:59:59.000Z

    , an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict...

  5. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect (OSTI)

    Not Available

    1995-02-01T23:59:59.000Z

    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.

  6. Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems 

    E-Print Network [OSTI]

    Szakaly, Frank Joseph

    2004-09-30T23:59:59.000Z

    The purpose of this work is to investigate the implementation of nitride fuels containing little or no uranium in a fast-spectrum nuclear reactor to reduce the amount of plutonium and minor actinides in spent nuclear fuel ...

  7. EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries.  The spent fuel would be shipped across...

  8. Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

    SciTech Connect (OSTI)

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)] [Moore Nuclear Energy, LLC (United States)

    2013-07-01T23:59:59.000Z

    Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

  9. Distribution of characteristics of LWR [light water reactor] spent fuel

    SciTech Connect (OSTI)

    Reich, W.J.; Notz, K.J. [Oak Ridge National Lab., TN (USA); Moore, R.S. [Automated Sciences Group, Inc., Oak Ridge, TN (USA)

    1991-01-01T23:59:59.000Z

    The purpose of this report is to develop a collective description of the entire spent fuel inventory in terms of various fuel properties relevant to Approved Testing Materials (ATMs) using information available from the Characteristics Data Base (CBD), which is sponsored by the US Department of Energy`s (DOE`s) Office of Civilian Radioactive Waste Management. A number of light-water reactor (LWR) characteristics were analyzed including assembly class representation, fuel burnup, enrichment, fuel fabrication data, defective fuel quantities, and, at PNL`s specific request, linear heat generation rate (LHGR) and the utilization of burnable poisons. A quantitative relationships was developed between burnup and enrichment for BWRs and PWRs. The relationship shows that the existing BWR ATM is near the center of the burnup-enrichment distribution, while the four PWR ATMs bracket the center of the burnup range but are on the low side of the enrichment range. Fuel fabrication data are based on vendor specifications for new fuel. Defective fuel distributions were analyzed in terms of assembly class and vendor design. LHGR values were calculated from utility data on burnup and effective full-power days; these calculations incorporate some unavoidable assumptions which may compromise the value of the results. Only a limited amount of data are available on burnable poisons at this time. Based on this distribution study, suggestions for additional ATMs are made. These are based on the class and design concepts and include BWR/2,3 barrier fuel, and the WE 17 {times} 17 class with integral burnable poison. Both should be at relatively high burnups. 16 refs., 5 figs., 15 tabs.

  10. Assessment of uranium-free nitride fuels for spent fuel transmutation in fast reactor systems

    E-Print Network [OSTI]

    Szakaly, Frank Joseph

    2004-09-30T23:59:59.000Z

    The purpose of this work is to investigate the implementation of nitride fuels containing little or no uranium in a fast-spectrum nuclear reactor to reduce the amount of plutonium and minor actinides in spent nuclear fuel destined for the Yucca...

  11. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect (OSTI)

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18T23:59:59.000Z

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

  12. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect (OSTI)

    Chad Pope

    2007-05-01T23:59:59.000Z

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day operations as well as obtaining historical information. Over 12,000 driver fuel elements have been processed resulting in the production of 2500 kg of 20% enriched uranium. Also, over one thousand blanket fuel elements have been processed resulting in the production of 2400 kg of depleted uranium. These operations required over 35,000 fissile material transfers between zones and over 6000 transfers between containers. Throughout all of these movements, no mass limit violations occurred. Numerous lessons were learned over the ten year operating history. From a criticality safety perspective, the most important lesson learned was the involvement of a criticality safety practitioner in daily operations. A criticality safety engineer was assigned directly to facility operations, and was responsible for implementation of limits and controls including upkeep of the associated computerized tracking files. The criticality safety engineer was also responsible for conducting fuel handler training activities including serving on fuel handler qualification oral boards, and continually assessing operations from a criticality control perspective. The criticality safety engineer also attended bimonthly project planning meetings to identify upcoming process changes that would require criticality safety evaluation. Finally, the excellent criticality safety record was due in no small part to the continual support, involvement, trust, and confidence of project and operations mana

  13. Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel

    E-Print Network [OSTI]

    A. C. Hayes; Gerard Jungman

    2012-05-30T23:59:59.000Z

    The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

  14. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect (OSTI)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01T23:59:59.000Z

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactorsspent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  15. Spent nuclear fuel discharges from US reactors 1992

    SciTech Connect (OSTI)

    Not Available

    1994-05-05T23:59:59.000Z

    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. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11T23:59:59.000Z

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  17. Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria

    SciTech Connect (OSTI)

    K. J. Allen; T. G. Apostolov; I. S. Dimitrov

    2009-03-01T23:59:59.000Z

    The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

  18. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

  19. 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.

    2014-06-10T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

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

    2013-11-05T23:59:59.000Z

    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.

  1. Melt-Dilute Treatment Technology for Aluminum Based Research Reactor Spent Fuel

    SciTech Connect (OSTI)

    Adams, T.

    1999-11-05T23:59:59.000Z

    The United States Department of Energy has selected the Savannah River Site (SRS) as the location to consolidate and store Aluminum Spent Nuclear Fuel (SNF), originating in the United States, from Foreign Research Reactor (FRR) and Domestic Research Reactor (DRR) through the Environmental Impact Statement (EIS) process. These SNF are either in service, being stored in water basins or in dry storage casks at the reactor sites, or have been transferred to SRS and stored in water basins. A portion of this inventory contains HEU. Since the fuel receipts would continue for several decades beyond projected SRS canyon operations, it is anticipated that it will be necessary to develop disposal technologies that do not rely on reprocessing. The Research Reactor Spent Nuclear Fuel Task Team, appointed by the Office of Spent Fuel Management of DOE, assessed and identified the most promising technology options for the alternative disposition of aluminum based domestic and foreign research reactor SNF in a geologic repository. The most promising options identified by the task team were direct/ co-disposal and melt-dilute technologies. The DOE through the SRS has evaluated the two options and has identified Melt-Dilute Treatment Technology as the preferred alternative in the Draft Environmental Impact Statement for the ultimate disposal of Al-SNF in the Mined Geologic Disposal System.

  2. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    SciTech Connect (OSTI)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01T23:59:59.000Z

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  3. Analysis of spent, highly enriched reactor fuel by delayed neutron interrogation

    SciTech Connect (OSTI)

    Piper, T.C.; Kirkham, R.J. (Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States)); Eccleston, G.W.; Menlove, H.O. (Los Alamos National Lab., NM (United States))

    1989-06-22T23:59:59.000Z

    Design aspects are given of a neutron shuffler designed to measure fissile material content of spent, highly enriched reactor fuel. The mode of operation used, results of analyzing 176 fuel packages and recommended system improvements are also discussed. Four measurements were made on each of the fuel packages with the mean of the 176 standard deviations being 1.7 percent of value. The maximum individual standard deviation was 6.3%. Use of a stronger neutron source, an improved neutron source shuffler, an improved fuel package motion system and modernized computer system should permit significant improvement of present performance. 2 refs.

  4. Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel

    SciTech Connect (OSTI)

    L. Angers

    2001-01-31T23:59:59.000Z

    The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

  5. Nuclide Composition Benchmark Data Set for Verifying Burnup Codes on Spent Light Water Reactor Fuels

    SciTech Connect (OSTI)

    Nakahara, Yoshinori; Suyama, Kenya; Inagawa, Jun; Nagaishi, Ryuji; Kurosawa, Setsumi; Kohno, Nobuaki; Onuki, Mamoru; Mochizuki, Hiroki [Japan Atomic Energy Research Institute (Japan)

    2002-02-15T23:59:59.000Z

    To establish a nuclide composition benchmark data set for the verification of burnup codes, destructive analyses of light water reactor spent-fuel samples, which were cut out from several heights of spent-fuel rods, were carried out at the analytical laboratory at the Japan Atomic Energy Research Institute. The 16 samples from three kinds of pressurized water reactor (PWR) fuel rods and the 18 samples from two boiling water reactor (BWR) fuel rods were examined. Their initial {sup 235}U enrichments and burnups were from 2.6 to 4.1% and from 4 to 50 GWd/t, respectively. One PWR fuel rod and one BWR fuel rod contained gadolinia as a burnable poison. The measurements for more than 40 nuclides of uranium, transuranium, and fission product elements were performed by destructive analysis using mass spectrometry, and alpha-ray and gamma-ray spectrometry. Burnup for each sample was determined by the {sup 148}Nd method. The analytical methods and the results as well as the related irradiation condition data are compiled as a complete benchmark data set.

  6. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01T23:59:59.000Z

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  7. Categorization of failed and damaged spent LWR (light-water reactor) fuel currently in storage

    SciTech Connect (OSTI)

    Bailey, W.J.

    1987-11-01T23:59:59.000Z

    The results of a study that was jointly sponsored by the US Department of Energy and the Electric Power Research Institute are described in this report. The purpose of the study was to (1) estimate the number of failed fuel assemblies and damaged fuel assemblies (i.e., ones that have sustained mechanical or chemical damage but with fuel rod cladding that is not breached) in storage, (2) categorize those fuel assemblies, and (3) prepare this report as an authoritative, illustrated source of information on such fuel. Among the more than 45,975 spent light-water reactor fuel assemblies currently in storage in the United States, it appears that there are nearly 5000 failed or damaged fuel assemblies. 78 refs., 23 figs., 19 tabs.

  8. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil)] [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil)] [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina)] [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)] [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01T23:59:59.000Z

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  9. Tools Developed to Prepare and Stabilize Reactor Spent Fuel for Retrieval from Tile Holes - 12251

    SciTech Connect (OSTI)

    Horne, Michael; Clough, Malcolm [Atomic Energy of Canada Limited (Canada)

    2012-07-01T23:59:59.000Z

    Spent fuel from the Chalk River Laboratories (CRL) nuclear reactors is stored in the waste management areas on site. This fuel is contained within carbon steel spent fuel cans that are stored inside vertical carbon steel lined concrete pipes in the ground known as tile holes. The fuel cans have been stored in the tile holes for greater than 30 years. Some of the fuel cans have experienced corrosion which may have affected their structural integrity as well as the potential to form hydrogen gas. In addition to these potential hazards, there was a need to clean contaminated surfaces inside of and around the exposed upper surface of the tile holes. As part of the site waste management remediation plan spent fuel will be retrieved from degraded tile holes, dried, and relocated to a new purpose built above ground storage facility. There have been a number of tools that are required to be developed to ensure spent fuel cans are in a safe condition prior to retrieval and re-location. A series of special purpose tools have been designed and constructed to stabilize the contents of the tile holes, to determine the integrity of the fuel containers and to decontaminate inside and around the tile holes. Described herein are the methods and types of tools used. Tools that have been presented here have been used, or will be used in the near future, in the waste management areas of the CRL Site in preparation for storage of spent fuel in a new above ground facility. The stabilization tools have been demonstrated on mock-up facilities prior to successful use in the field to remove hydrogen gas and uranium hydrides from the fuel cans. A lifting tool has been developed and used successfully in the field to confirm the integrity of the fuel cans for future relocation. A tool using a commercial dry ice blaster has been developed and is ready to start mock-up trials and is scheduled to be used in the field during the summer of 2012. (authors)

  10. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

    1993-01-01T23:59:59.000Z

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  11. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    DOE Patents [OSTI]

    Corletti, M.M.; Lau, L.K.; Schulz, T.L.

    1993-12-14T23:59:59.000Z

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

  12. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    SciTech Connect (OSTI)

    Not Available

    1992-07-01T23:59:59.000Z

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

  13. Radiation Exposures Associated with Shipments of Foreign Research Reactor Spent Nuclear Fuel

    SciTech Connect (OSTI)

    MASSEY,CHARLES D.; MESSICK,C.E.; MUSTIN,T.

    1999-11-01T23:59:59.000Z

    Experience has shown that the analyses of marine transport of spent fuel in the Environmental Impact Statement (EIS) were conservative. It is anticipated that for most shipments. The external dose rate for the loaded transportation cask will be more in line with recent shipments. At the radiation levels associated with these shipments, we would not expect any personnel to exceed radiation exposure limits for the public. Package dose rates usually well below the regulatory limits and personnel work practices following ALARA principles are keeping human exposures to minimal levels. However, the potential for Mure shipments with external dose rates closer to the exclusive-use regulatory limit suggests that DOE should continue to provide a means to assure that individual crew members do not receive doses in excess of the public dose limits. As a minimum, the program will monitor cask dose rates and continue to implement administrative procedures that will maintain records of the dose rates associated with each shipment, the vessel used, and the crew list for the vessel. DOE will continue to include a clause in the contract for shipment of the foreign research reactor spent nuclear fuel requiring that the Mitigation Action Plan be followed.

  14. Spent graphite fuel element processing

    SciTech Connect (OSTI)

    Holder, N.D.; Olsen, C.W.

    1981-07-01T23:59:59.000Z

    The Department of Energy currently sponsors two programs to demonstrate the processing of spent graphite fuel elements. General Atomic in San Diego operates a cold pilot plant to demonstrate the processing of both US and German high-temperature reactor fuel. Exxon Nuclear Idaho Company is demonstrating the processing of spent graphite fuel elements from Rover reactors operated for the Nuclear Rocket Propulsion Program. This work is done at Idaho National Engineering Laboratory, where a hot facility is being constructed to complete processing of the Rover fuel. This paper focuses on the graphite combustion process common to both programs.

  15. The U.S.-Russian joint studies on using power reactors to disposition surplus weapon plutonium as spent fuel

    SciTech Connect (OSTI)

    Chebeskov, A.; Kalashnikov, A. [State Scientific Center, Obninsk (Russian Federation). Inst. of Physics and Power Engineering; Bevard, B.; Moses, D. [Oak Ridge National Lab., TN (United States); Pavlovichev, A. [State Scientific Center, Moscow (Russian Federation). Kurchatov Inst.

    1997-09-01T23:59:59.000Z

    In 1996, the US and the Russian Federation completed an initial joint study of the candidate options for the disposition of surplus weapons plutonium in both countries. The options included long term storage, immobilization of the plutonium in glass or ceramic for geologic disposal, and the conversion of weapons plutonium to spent fuel in power reactors. For the latter option, the US is only considering the use of existing light water reactors (LWRs) with no new reactor construction for plutonium disposition, or the use of Canadian deuterium uranium (CANDU) heavy water reactors. While Russia advocates building new reactors, the cost is high, and the continuing joint study of the Russian options is considering only the use of existing VVER-1000 LWRs in Russia and possibly Ukraine, the existing BN-60O fast neutron reactor at the Beloyarsk Nuclear Power Plant in Russia, or the use of the Canadian CANDU reactors. Six of the seven existing VVER-1000 reactors in Russia and the eleven VVER-1000 reactors in Ukraine are all of recent vintage and can be converted to use partial MOX cores. These existing VVER-1000 reactors are capable of converting almost 300 kg of surplus weapons plutonium to spent fuel each year with minimum nuclear power plant modifications. Higher core loads may be achievable in future years.

  16. HFIR spent fuel management alternatives

    SciTech Connect (OSTI)

    Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

    1992-10-15T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

  17. HFIR spent fuel management alternatives

    SciTech Connect (OSTI)

    Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

    1992-10-15T23:59:59.000Z

    The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems` Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

  18. Spent fuel integrity during transportation

    SciTech Connect (OSTI)

    Funk, C.W.; Jacobson, L.D.

    1980-01-01T23:59:59.000Z

    The conditions of recent shipments of light water reactor spent fuel were surveyed. The radioactivity level of cask coolant was examined in an attempt to find the effects of transportation on LWR fuel assemblies. Discussion included potential cladding integrity loss mechanisms, canning requirements, changes of radioactivity levels, and comparison of transportation in wet or dry media. Although integrity loss or degradation has not been identified, radioactivity levels usually increase during transportation, especially for leaking assemblies.

  19. Process and apparatus for recovery of fissionable materials from spent reactor fuel by anodic dissolution

    DOE Patents [OSTI]

    Tomczuk, Zygmunt (Orland Park, IL); Miller, William E. (Naperville, IL); Wolson, Raymond D. (Lockport, IL); Gay, Eddie C. (Park Forest, IL)

    1991-01-01T23:59:59.000Z

    An electrochemical process and apparatus for the recovery of uranium and plutonium from spent metal clad fuel pins is disclosed. The process uses secondary reactions between U.sup.+4 cations and elemental uranium at the anode to increase reaction rates and improve anodic efficiency compared to prior art processes. In another embodiment of the process, secondary reactions between Cd.sup.+2 cations and elemental uranium to form uranium cations and elemental cadmium also assists in oxidizing the uranium at the anode.

  20. Wastes associated with recycling spent MOX fuel into fast reactor oxide fuel

    SciTech Connect (OSTI)

    Foare, G.; Meze, F. [AREVA EP, SGN - 1, rue des Herons, 18182 Montigny-le-Bretonneux (France); McGee, D.; Murray, P.; Bader, S. [AREVA Federal Services LLC - 7207 IBM Drive, Charlotte, NC 28262 (United States)

    2013-07-01T23:59:59.000Z

    A study sponsored by the DOE has been performed by AREVA to estimate the process and secondary wastes produced from an 800 MTIHM/yr (initial metric tons heavy metal a year) recycling plant proposed to be built in the U.S. utilizing the COEX process and utilized some DOE defined assumptions and constraints. In this paper, this plant has been analyzed for a recycling campaign that included 89% UO{sub x} and 11% MOX UNF to estimate process and secondary waste quantities produced while manufacturing 28 MTIHM/yr of SFR fuel. AREVA utilized operational data from its backend facilities in France (La Hague and MELOX), and from recent advances in waste treatment technology to estimate the waste quantities. A table lists the volumes and types of the different final wastes for a recycling plant. For instance concerning general fission products the form of the final wastes is vitrified glass and its volume generation rate is 135 l/MTHM, concerning Iodine 129 waste its final form is synthetic rock and its volume generation rate is 0.625 l/MTIHM.

  1. Spent fuel storage requirements 1993--2040

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges.

  2. Systems report on the analysis of spent, highly enriched U-235 reactor fuel by delayed neutron interrogation

    SciTech Connect (OSTI)

    Piper, T.C.; Kirkham, R.J.

    1990-05-01T23:59:59.000Z

    Design aspects are briefly given of a neutron source shuffler used to measure fissile material content of spent, highly enriched reactor fuel. The mode of operation used, results of analyzing 176 fuel packages and recommended system improvements are discussed. Four measurements were made on each of the fuel packages with the mean of the 176 standard deviations being 2.03 percent of value. The maximum individual standard deviation was 9.27 percent. Appendixes concerning imprecisions introduced by counting statistics and crane speed irregularities are given. Use of an improved neutron source shuffler, an improved fuel package motion system and modernized computer system should permit system performance to be limited mainly by counting statistics, to about 1.5 percent of measured value. A stronger source could then be installed to further enhance system operation. 16 figs., 3 tabs.

  3. Characteristics of spent nuclear fuel

    SciTech Connect (OSTI)

    Notz, K.J.

    1988-04-01T23:59:59.000Z

    The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for the spent fuels and other wastes that will, or may, eventually be disposed of in a geological repository. The two major sources of these materials are commercial light-water reactor (LWR) spent fuel and immobilized high-level waste (HLW). Other wastes that may require long-term isolation include non-LWR spent fuels and miscellaneous sources such as activated metals. This report deals with spent fuels, but for completeness, the other sources are described briefly. Detailed characterizations are required for all of these potential repository wastes. These characteristics include physical, chemical, and radiological properties. The latter must take into account decay as a function of time. In addition, the present inventories and projected quantities of the various wastes are needed. This information has been assembled in a Characteristics Data Base which provides data in four formats: hard copy standard reports, menu-driven personal computer (PC) data bases, program-level PC data bases, and mainframe computer files. 5 refs., 3 figs., 4 tabs.

  4. Electrochemical separation of aluminum from uranium for research reactor spent nuclear fuel applications.

    SciTech Connect (OSTI)

    Slater, S. A.; Willit, J. L.; Gay, E. C.; Chemical Engineering

    1999-01-01T23:59:59.000Z

    Researchers at Argonne National Laboratory (ANL) are developing an electrorefining process to treat aluminum-based spent nuclear fuel by electrochemically separating aluminum from uranium. The aluminum electrorefiner is modeled after the high-throughput electrorefiner developed at ANL. Aluminum is electrorefined, using a fluoride salt electrolyte, in a potential range of -0.1 V to -0.2 V, while uranium is electrorefined in a potential range of -0.3 V to -0.4 V; therefore, aluminum can be selectively separated electrochemically from uranium. A series of laboratory-scale experiments was performed to demonstrate the aluminum electrorefining concept. These experiments involved selecting an electrolyte (determining a suitable fluoride salt composition); selecting a crucible material for the electrochemical cell; optimizing the operating conditions; determining the effect of adding alkaline and rare earth elements to the electrolyte; and demonstrating the electrochemical separation of aluminum from uranium, using a U-Al-Si alloy as a simulant for aluminum-based spent nuclear fuel. Results of the laboratory-scale experiments indicate that aluminum can be selectively electrotransported from the anode to the cathode, while uranium remains in the anode basket.

  5. Scintillator spent fuel monitor

    SciTech Connect (OSTI)

    Moss, C.E.; Nixon, K.V.; Bernard, W.

    1980-01-01T23:59:59.000Z

    A monitor for rapidly measuring the gross gamma-ray flux immediately above spent fuel assemblies in underwater storage racks has been developed. It consists of a plastic scintillator, photomultiplier, collimator, and a small battery-powered electronics package. The crosstalk from an isolated fuel assembly to an adjacent void is only about 2%. The mean difference between the measured gamma-ray flux and the flux estimated from the declared burnup and cooling time with a simple formula is 22%.

  6. EIS-0015: U.S. Spent Fuel Policy

    Broader source: Energy.gov [DOE]

    Subsumed DOE/EIS-0040 and DOE/EIS-0041. The Savannah River Laboratory prepared this EIS to analyze the impacts of implementing or not implementing the policy for interim storage of spent power reactor fuel. This Final EIS is a compilation of three Draft EISs and one Supplemental Draft EIS: DOE/EIS-0015-D, Storage of U.S. Spent Power Reactor Fuel; DOE/EIS-0015-DS, Storage of U.S. Spent Power Reactor Fuel - Supplement; DOE/EIS-0040-D, Storage of Foreign Spent Power Reactor Fuel; and DOE/EIS-0041-D, Charge for Spent Fuel Storage.

  7. AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR

    SciTech Connect (OSTI)

    Dewes, J.

    2014-02-24T23:59:59.000Z

    The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

  8. Examination of Spent Pressurized Water Reactor Fuel Rods After 15 Years in Dry Storage

    SciTech Connect (OSTI)

    Einziger, Robert E. [Argonne National Laboratory (United States); Tsai Hanchung [Argonne National Laboratory (United States); Billone, Michael C. [Argonne National Laboratory (United States); Hilton, Bruce A. [Argonne National Laboratory-West (United States)

    2003-11-15T23:59:59.000Z

    For [approximately equal to]15 yr Dominion Generation's Surry Nuclear Station 15 x 15 Westinghouse pressurized water reactor (PWR) fuel was stored in a dry inert-atmosphere Castor V/21 cask at the Idaho National Environmental and Engineering Laboratory at peak cladding temperatures that decreased from {approx}350 to 150 deg. C. Before storage, the loaded cask was subjected to thermal-benchmark tests, during which time the peak temperatures were greater than 400 deg. C. The cask was opened to examine the fuel rods for degradation and to determine if they were suitable for extended storage. No fuel rod breaches and no visible degradation or crud/oxide spallation from the fuel rod surface were observed. The results from profilometry, gas release measurements, metallographic examinations, microhardness determination, and cladding hydrogen behavior are reported in this paper.It appears that little or no fission gas was released from the fuel pellets during either the thermal-benchmark tests or the long-term storage. In the central region of the fuel column, where the axial temperature gradient in storage is small, the measured hydrogen content in the cladding is consistent with the thickness of the oxide layer. At {approx}1 m above the fuel midplane, where a steep temperature gradient existed in the cask, less hydrogen is present than would be expected from the oxide thickness that developed in-reactor. Migration of hydrogen during dry storage probably occurred and may signal a higher-than-expected concentration at the cooler ends of the rod. The volume of hydrides varies azimuthally around the cladding, and at some elevations, the hydrides appear to have segregated somewhat to the inner and outer cladding surfaces. It is, however, impossible to determine if this segregation occurred in-reactor or during transportation, thermal-benchmark tests, or the dry storage period. The hydrides retained the circumferential orientation typical of prestorage PWR fuel rods. Little or no cladding creep occurred during thermal-benchmark testing and dry storage. It is anticipated that the creep would not increase significantly during additional storage because of the lower temperature after 15 yr, continual decrease in temperature from the reduction in decay heat, and concurrent reductions in internal rod pressure and stress. This paper describes the results of the characterization of the fuel and intact cladding, as well as the implications of these results for long-term (i.e., beyond 20 yr) dry-cask storage.

  9. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01T23:59:59.000Z

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM'S program specifications. (CBS)

  10. Possible effects of UO/sub 2/ oxidation on light water reactor spent fuel performance in long-term geologic disposal

    SciTech Connect (OSTI)

    Almassy, M.Y.; Woodley, R.E.

    1982-08-01T23:59:59.000Z

    Disposal of spent nuclear fuel in a conventionally mined geologic formation is the nearest-term option for permanently isolating radionuclides from the biosphere. Because irradiated uranium dioxide (UO/sub 2/) fuel pellets retain 95 to 99% of the radionuclides generated during normal light water reactor operation, they may represent a significant barrier to radionuclide release. This document presents a technical assessment of published literature representing the current level of understanding of spent fuel characteristics and conditions that may degrade pellet integrity during a geologic disposal sequence. A significant deterioration mechanism is spent UO/sub 2/ oxidation with possible consequences identified as fission gas release, rod diameter increases, cladding breach extension, and release of solid fuel particles containing radionuclides. Areas requiring further study to support development of a comprehensive spent fuel performance prediction model are highlighted. A program and preliminary schedule to obtain the information needed to develop model correlations are also presented.

  11. Shielding analysis for the 300 area light water reactor spent nuclear fuel within a modified multi-canister overpack canister in a modified multi-canister overpack cask

    SciTech Connect (OSTI)

    Gedeon, S.R.

    1997-04-11T23:59:59.000Z

    Spent light water reactor fuel is to be moved out of the 324 Building. It is anticipated that intact fuel assemblies will be loaded in a modified Multi-Canister Overpack Canister, which in turn will be placed in an Overpack Transportation Cask. An estimate of gamma ray dose rates from a transportation cask is desired.

  12. Spent Fuel Background Report Volume I

    SciTech Connect (OSTI)

    Abbott, D.

    1994-03-01T23:59:59.000Z

    This report is an overview of current spent nuclear fuel management in the DOE complex. Sources of information include published literature, internal DOE documents, interviews with site personnel, and information provided by individual sites. Much of the specific information on facilities and fuels was provided by the DOE sites in response to the questionnaire for data for spent fuels and facilities data bases. This information is as accurate as is currently available, but is subject to revision pending results of further data calls. Spent fuel is broadly classified into three categories: (a) production fuels, (b) special fuels, and (c) naval fuels. Production fuels, comprising about 80% of the total inventory, are those used at Hanford and Savannah River to produce nuclear materials for defense. Special fuels are those used in a wide variety of research, development, and testing activities. Special fuels include fuel from DOE and commercial reactors used in research activities at DOE sites. Naval fuels are those developed and used for nuclear-powered naval vessels and for related research and development. Given the recent DOE decision to curtail reprocessing, the topic of main concern in the management of spent fuel is its storage. Of the DOE sites that have spent nuclear fuel, the vast majority is located at three sites-Hanford, INEL, and Savannah River. Other sites with spent fuel include Oak Ridge, West Valley, Brookhaven, Argonne, Los Alamos, and Sandia. B&W NESI Lynchburg Technology Center and General Atomics are commercial facilities with DOE fuel. DOE may also receive fuel from foreign research reactors, university reactors, and other commercial and government research reactors. Most DOE spent fuel is stored in water-filled pools at the reactor facilities. Currently an engineering study is being performed to determine the feasibility of using dry storage for DOE-owned spent fuel currently stored at various facilities. Delays in opening the deep geologic repository and the decision to phase out reprocessing of production fuels are extending the need for interim storage. The report describes the basic storage conditions and the general SNF inventory at individual DOE facilities.

  13. State of Nevada comments on the OCRWM from-reactor spent fuel shipping cask preliminary design reports

    SciTech Connect (OSTI)

    Halstead, R.J.; Audin, L.; Hoskins, R.E.; Snedeker, D.F.

    1990-12-01T23:59:59.000Z

    The design of spent fuels shipping casks is described. Two casks from two different contractors are presented. The design needs are based on the OCRWM`S program specifications. (CBS)

  14. Reduction of the Radiotoxicity of Spent Nuclear Fuel Using a Two-Tiered System Comprising Light Water Reactors and Accelerator-Driven Systems

    SciTech Connect (OSTI)

    H.R. Trellue

    2003-06-01T23:59:59.000Z

    Two main issues regarding the disposal of spent nuclear fuel from nuclear reactors in the United States in the geological repository Yucca Mountain are: (1) Yucca Mountain is not designed to hold the amount of fuel that has been and is proposed to be generated in the next few decades, and (2) the radiotoxicity (i.e., biological hazard) of the waste (particularly the actinides) does not decrease below that of natural uranium ore for hundreds of thousands of years. One solution to these problems may be to use transmutation to convert the nuclides in spent nuclear fuel to ones with shorter half-lives. Both reactor and accelerator-based systems have been examined in the past for transmutation; there are advantages and disadvantages associated with each. By using existing Light Water Reactors (LWRs) to burn a majority of the plutonium in spent nuclear fuel and Accelerator-Driven Systems (ADSs) to transmute the remainder of the actinides, the benefits of each type of system can be realized. The transmutation process then becomes more efficient and less expensive. This research searched for the best combination of LWRs with multiple recycling of plutonium and ADSs to transmute spent nuclear fuel from past and projected nuclear activities (assuming little growth of nuclear energy). The neutronic design of each system is examined in detail although thermal hydraulic performance would have to be considered before a final system is designed. The results are obtained using the Monte Carlo burnup code Monteburns, which has been successfully benchmarked for MOX fuel irradiation and compared to other codes for ADS calculations. The best combination of systems found in this research includes 41 LWRs burning mixed oxide fuel with two recycles of plutonium ({approx}40 years operation each) and 53 ADSs to transmute the remainder of the actinides from spent nuclear fuel over the course of 60 years of operation.

  15. Hanford spent fuel inventory baseline

    SciTech Connect (OSTI)

    Bergsman, K.H.

    1994-07-15T23:59:59.000Z

    This document compiles technical data on irradiated fuel stored at the Hanford Site in support of the Hanford SNF Management Environmental Impact Statement. Fuel included is from the Defense Production Reactors (N Reactor and the single-pass reactors; B, C, D, DR, F, H, KE and KW), the Hanford Fast Flux Test Facility Reactor, the Shipping port Pressurized Water Reactor, and small amounts of miscellaneous fuel from several commercial, research, and experimental reactors.

  16. 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

    2008-07-01T23:59:59.000Z

    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.

  17. Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23, 2014, an OHASeptemberAssessments | Department ofSouthernofDepartmentReport6

  18. Spent nuclear fuel reprocessing modeling

    SciTech Connect (OSTI)

    Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V. [Bochvar Institute, 5 Rogova str., Moscow 123098 (Russian Federation); Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I. [Russian Federal Nuclear Center - VNIITF E.I. Zababakhin, p.o.box 245, Snezhinsk, 456770 (Russian Federation)

    2013-07-01T23:59:59.000Z

    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)

  19. Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 3, Site team reports

    SciTech Connect (OSTI)

    Not Available

    1993-11-01T23:59:59.000Z

    A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES & H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary`s request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford`s RINM storage circumstances. ES & H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks.

  20. Intermodal transportation of spent fuel

    SciTech Connect (OSTI)

    Elder, H.K.

    1983-09-01T23:59:59.000Z

    Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate.

  1. Spent-fuel-storage alternatives

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  2. US Spent (Used) Fuel Status, Management and Likely Directions- 12522

    SciTech Connect (OSTI)

    Jardine, Leslie J. [L. J. Jardine Services, Consultant, Dublin CA, 94568 (United States)

    2012-07-01T23:59:59.000Z

    As of 2010, the US has accumulated 65,200 MTU (42,300 MTU of PWR's; 23,000 MTU of BWR's) of spent (irradiated or used) fuel from 104 operating commercial nuclear power plants situated at 65 sites in 31 States and from previously shutdown commercial nuclear power plants. Further, the Department of Energy (DOE) has responsibility for an additional 2458 MTU of DOE-owned defense and non defense spent fuel from naval nuclear power reactors, various non-commercial test reactors and reactor demonstrations. The US has no centralized large spent fuel storage facility for either commercial spent fuel or DOE-owned spent fuel. The 65,200 MTU of US spent fuel is being safely stored by US utilities at numerous reactor sites in (wet) pools or (dry) metal or concrete casks. As of November 2010, the US had 63 'independent spent fuel storage installations' (or ISFSI's) licensed by the US Nuclear Regulatory Commission located at 57 sites in 33 states. Over 1400 casks loaded with spent fuel for dry storage are at these licensed ISFSI's; 47 sites are located at commercial reactor sites and 10 are located 'away' from a reactor (AFR's) site. DOE's small fraction of a 2458 MTU spent fuel inventory, which is not commercial spent fuel, is with the exception of 2 MTU, being stored at 4 sites in 4 States. The decades old US policy of a 'once through' fuel cycle with no recycle of spent fuel was set into a state of 'mass confusion or disruption' when the new US President Obama's administration started in early 2010 stopping the only US geologic disposal repository at the Yucca Mountain site in the State of Nevada from being developed and licensed. The practical result is that US nuclear power plant operators will have to continue to be responsible for managing and storing their own spent fuel for an indefinite period of time at many different sites in order to continue to generate electricity because there is no current US government plan, schedule or policy for taking possession of accumulated spent fuel from the utilities. There are technical solutions for continuing the safe storage of spent fuel for 100 years or more and these solutions will be implemented by the US utilities that need to keep their nuclear power plants operating while the unknown political events are played out to establish future US policy decisions that can remain in place long enough regarding accumulated spent fuel inventories to implement any new US spent fuel centralized storage or disposition policy by the US government. (author)

  3. Spent Nuclear Fuel project, project management plan

    SciTech Connect (OSTI)

    Fuquay, B.J.

    1995-10-25T23:59:59.000Z

    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

  4. Pyrochemical Treatment of Spent Nuclear Fuel

    SciTech Connect (OSTI)

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

    2005-10-01T23:59:59.000Z

    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 Energy’s 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.

  5. Recommended temperature limits for dry storage of spent light water reactor Zircaloy-clad fuel rods in inert gas

    SciTech Connect (OSTI)

    Levy, I.S.; Chin, B.A.; Simonen, E.P.; Beyer, C.E.; Gilbert, E.R.; Johnson, A.B. Jr.

    1987-05-01T23:59:59.000Z

    It is concluded that the recommendation of a single-valued temperature limit of 380/sup 0/C should be replaced by multiple limits to account for variations in fuel design, burnup level, spent fuel age, and storage cask design. A single-valued limit to account for these factors would, in some situations, impose unnecessary conservatisms and, potentially, economic penalties for utilities and storage cask vendors. The technical validity and conservatism of the CSFM model should assure acceptance by the NRC for utility and cask vendor use.

  6. Characterization plan for Hanford spent nuclear fuel

    SciTech Connect (OSTI)

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

    1994-12-01T23:59:59.000Z

    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.

  7. Inventory of LWR spent nuclear fuel in the 324 Building

    SciTech Connect (OSTI)

    Jenquin, U.P.

    1996-09-01T23:59:59.000Z

    This document contains the results of calculations to estimate the decay heat, neutron source term, photon source term, and radioactive inventory of light-water-reactor spent nuclear fuel in the 324 Building at Pacific Northwest National Laboratory.

  8. Investigation of the condition of spent-fuel pool components

    SciTech Connect (OSTI)

    Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

    1981-09-01T23:59:59.000Z

    It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts.

  9. Spent Fuel Transportation Risk Assessment

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently AskedEnergyIssues DOE's Nuclear EnergySmartOverviewEnergy Spelman GraduateSpent

  10. An experiment to simulate the heat transfer properties of a dry, horizontal spent nuclear fuel assembly

    E-Print Network [OSTI]

    Lovett, Phyllis Maria

    1991-01-01T23:59:59.000Z

    Nuclear power reactors generate highly radioactive spent fuel assemblies. Initially, the spent fuel assemblies are stored for a period of several years in an on-site storage facility to allow the radioactivity levels of ...

  11. Spent Nuclear Fuel Alternative Technology Risk Assessment

    SciTech Connect (OSTI)

    Perella, V.F.

    1999-11-29T23:59:59.000Z

    A Research Reactor Spent Nuclear Fuel Task Team (RRTT) was chartered by the Department of Energy (DOE) Office of Spent Fuel Management with the responsibility to recommend a course of action leading to a final technology selection for the interim management and ultimate disposition of the foreign and domestic aluminum-based research reactor spent nuclear fuel (SNF) under DOE''s jurisdiction. The RRTT evaluated eleven potential SNF management technologies and recommended that two technologies, direct co-disposal and an isotopic dilution alternative, either press and dilute or melt and dilute, be developed in parallel. Based upon that recommendation, the Westinghouse Savannah River Company (WSRC) organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and provide a WSRC recommendation to DOE for a preferred SNF alternative management technology. A technology risk assessment was conducted as a first step in this recommendation process to determine if either, or both, of the technologies posed significant risks that would make them unsuitable for further development. This report provides the results of that technology risk assessment.

  12. Arrival condition of spent fuel after storage, handling, and transportation

    SciTech Connect (OSTI)

    Bailey, W.J.; Pankaskie, P.J.; Langstaff, D.C.; Gilbert, E.R.; Rising, K.H.; Schreiber, R.E.

    1982-11-01T23:59:59.000Z

    This report presents the results of a study conducted to determine the probable arrival condition of spent light-water reactor (LWR) fuel after handling and interim storage in spent fuel storage pools and subsequent handling and accident-free transport operations under normal or slightly abnormal conditions. The objective of this study was to provide information on the expected condition of spent LWR fuel upon arrival at interim storage or fuel reprocessing facilities or at disposal facilities if the fuel is declared a waste. Results of a literature survey and data evaluation effort are discussed. Preliminary threshold limits for storing, handling, and transporting unconsolidated spent LWR fuel are presented. The difficulty in trying to anticipate the amount of corrosion products (crud) that may be on spent fuel in future shipments is also discussed, and potential areas for future work are listed. 95 references, 3 figures, 17 tables.

  13. Preliminary study on direct recycling of spent PWR fuel in PWR system

    SciTech Connect (OSTI)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesa 10 Bandung 40132 (Indonesia)

    2012-06-06T23:59:59.000Z

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  14. Spent nuclear fuel sampling strategy

    SciTech Connect (OSTI)

    Bergmann, D.W.

    1995-02-08T23:59:59.000Z

    This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation.

  15. Probability of spent fuel transportation accidents

    SciTech Connect (OSTI)

    McClure, J. D.

    1981-07-01T23:59:59.000Z

    The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10/sup -7/ spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10/sup -9//mile.

  16. Validation Work to Support the Idaho National Engineering and Environmental Laboratory Calculational Burnup Methodology Using Shippingport Light Water Breeder Reactor (LWBR) Spent Fuel Assay Data

    SciTech Connect (OSTI)

    J. W. Sterbentz

    1999-08-01T23:59:59.000Z

    Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

  17. Spent Fuel Working Group Report. Volume 1

    SciTech Connect (OSTI)

    O`Toole, T.

    1993-11-01T23:59:59.000Z

    The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety. To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary`s initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group`s Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities.

  18. Spent fuel and fuel pool component integrity. Annual report, FY 1980

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Bailey, W.J.; Bradley, E.R.; Bruemmer, S.M.; Langstaff, D.C.

    1981-09-01T23:59:59.000Z

    During program FY 1980 staff members of the Spent Fuel and Fuel Pool Component Integrity Program at Pacific Northwest Laboratory (PNL) completed the following major tasks: represented DOE on the international Behavior of Fuel Assemblies in Storage (BEFAST) Committee; the program manager, A.B. Johnson, Jr., participated in an International Survey of Water Reactor Spent Fuel Storage Experience, which was conducted jointly by the International Atomic Energy Agency (Vienna) and the Nuclear Energy Agency (Paris); provided written testimony and cross statement for the Proposed Rulemaking on Storage and Disposal of Nuclear Waste; acquired and began examination of the world's oldest pool-stored Zircaloy-clad fuel from the Shippingport reactor, stored approx. 21 years in deionized water; acquired and began examination of stainless-clad spent fuel from the Connecticut Yankee Reactor (PWR); negotiated for specimens from components stored in spent fuel pools at fuel storage facilities from the Savannah River Plant, Aiken, South Carolina, Zion (PWR) spent fuel pool, Zion, Illinois, and La Crosse (BWR) spent fuel pool, La Crosse, Wisconsin; planned for examinations in FY 81 of specimens from the three spent fuel pools; investigated a low-temperature stress corrosion cracking mechanism that developed in piping at a few PWR spent fuel pools. This report summarizes the results of these activities and investigations. Details are provided in the presentationsand publications generated under this program and summarized in Appendix A.

  19. Spent Fuel Disposal Trust Fund (Maine)

    Broader source: Energy.gov [DOE]

    Any licensee operating a nuclear power plant in this State shall establish a segregated Spent Nuclear Fuel Disposal Trust Fund in accordance with this subchapter for the eventual disposal of spent...

  20. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    SciTech Connect (OSTI)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01T23:59:59.000Z

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO/sub 2/ oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO/sub 2/ pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs.

  1. Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 2, Working Group Assessment Team reports; Vulnerability development forms; Working group documents

    SciTech Connect (OSTI)

    Not Available

    1993-11-01T23:59:59.000Z

    The Secretary of Energy`s memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability.

  2. 324 Building B-Cell Pressurized Water Reactor Spent Fuel Packaging & Shipment RL Readiness Assessment Final Report [SEC 1 Thru 3

    SciTech Connect (OSTI)

    HUMPHREYS, D C

    2002-08-01T23:59:59.000Z

    A parallel readiness assessment (RA) was conducted by independent Fluor Hanford (FH) and U. S. Department of Energy, Richland Operations Office (RL) team to verify that an adequate state of readiness had been achieved for activities associated with the packaging and shipping of pressurized water reactor fuel assemblies from B-Cell in the 324 Building to the interim storage area at the Canister Storage Building in the 200 Area. The RL review was conducted in parallel with the FH review in accordance with the Joint RL/FH Implementation Plan (Appendix B). The RL RA Team members were assigned a FH RA Team counterpart for the review. With this one-on-one approach, the RL RA Team was able to assess the FH Team's performance, competence, and adherence to the implementation plan and evaluate the level of facility readiness. The RL RA Team agrees with the FH determination that startup of the 324 Building B-Cell pressurized water reactor spent nuclear fuel packaging and shipping operations can safely proceed, pending completion of the identified pre-start items in the FH final report (see Appendix A), completion of the manageable list of open items included in the facility's declaration of readiness, and execution of the startup plan to operations.

  3. Simulated Performance of the Integrated Passive Neutron Albedo Reactivity and Self-Interrogation Neutron Resonance Densitometry Detector Designed for Spent Fuel Measurement at the Fugen Reactor in Japan

    SciTech Connect (OSTI)

    Ulrich, Timothy J. II [Los Alamos National Laboratory; Lafleur, Adrienne M. [Los Alamos National Laboratory; Menlove, Howard O. [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory; Tobin, Stephen J. [Los Alamos National Laboratory; Seya, Michio [Los Alamos National Laboratory; Bolind, Alan M. [Los Alamos National Laboratory

    2012-07-16T23:59:59.000Z

    An integrated nondestructive assay instrument, which combined the Passive Neutron Albedo Reactivity (PNAR) and the Self-Interrogation Neutron Resonance Densitometry (SINRD) techniques, is the research focus for a collaborative effort between Los Alamos National Laboratory (LANL) and the Japanese Atomic Energy Agency as part of the Next Generation Safeguard Initiative. We will quantify the anticipated performance of this experimental system in two physical environments: (1) At LANL we will measure fresh Low Enriched Uranium (LEU) assemblies for which the average enrichment can be varied from 0.2% to 3.2% and for which Gd laced rods will be included. (2) At Fugen we will measure spent Mixed Oxide (MOX-B) and LEU spent fuel assemblies from the heavy water moderated Fugen reactor. The MOX-B assemblies will vary in burnup from {approx}3 GWd/tHM to {approx}20 GWd/tHM while the LEU assemblies ({approx}1.9% initial enrichment) will vary from {approx}2 GWd/tHM to {approx}7 GWd/tHM. The estimated count rates will be calculated using MCNPX. These preliminary results will help the finalization of the hardware design and also serve a guide for the experiment. The hardware of the detector is expected to be fabricated in 2012 with measurements expected to take place in 2012 and 2013. This work is supported by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

  4. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    SciTech Connect (OSTI)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    2012-06-06T23:59:59.000Z

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  5. Recycling of nuclear spent fuel with AIROX processing

    SciTech Connect (OSTI)

    Majumdar, D. [ed.] [USDOE Idaho Field Office, Idaho Falls, ID (United States); Jahshan, S.N.; Allison, C.M.; Kuan, P. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Thomas, T.R. [Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States)

    1992-12-01T23:59:59.000Z

    This report examines the concept of recycling light water reactor (LWR) fuel through use of a dry-processing technique known as the AIROX (Atomics International Reduction Oxidation) process. In this concept, the volatiles and the cladding from spent LWR fuel are separated from the fuel by the AIROX process. The fuel is then reenriched and made into new fuel pins with new cladding. The feasibility of the concept is studied from a technical and high level waste minimization perspective.

  6. Radiological consequences of ship collisions that might occur in U.S. Ports during the shipment of foreign research reactor spent nuclear fuel to the United States in break-bulk freighters

    SciTech Connect (OSTI)

    Sprung, J.L.; Bespalko, S.J.; Massey, C.D.; Yoshimura, R. [Sandia National Laboratory, Albuquerque, NM (United States); Johnson, J.D. [GRAM Inc., Albuquerque, NM (United States); Reardon, P.C. [PCRT Technologies, Albuquerque, NM (United States); Ebert, M.W.; Gallagher D.W. [Science Applications International Corp., Reston, VA (United States)

    1996-08-01T23:59:59.000Z

    Accident source terms, source term probabilities, consequences, and risks are developed for ship collisions that might occur in U.S. ports during the shipment of spent fuel from foreign research reactors to the United States in break-bulk freighters.

  7. Electrorefining {open_quotes}N{close_quotes} reactor fuel

    SciTech Connect (OSTI)

    Gay, E.C.; Miller, W.E.

    1995-02-01T23:59:59.000Z

    Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined.

  8. Spent Nuclear Fuel (SNF) Project Execution Plan

    SciTech Connect (OSTI)

    LEROY, P.G.

    2000-11-03T23:59:59.000Z

    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.

  9. Overview of the spent nuclear fuel project at Hanford

    SciTech Connect (OSTI)

    Daily, J.L. [Dept. of Energy, Richland, WA (United States). Richland Operations Office; Fulton, J.C.; Gerber, E.W.; Culley, G.E. [Westinghouse Hanford Co., Richland, WA (United States)

    1995-02-01T23:59:59.000Z

    The Spent Nuclear Fuel Project`s mission at Hanford is to {open_quotes}Provide safe, economic and environmentally sound management of Hanford spent nuclear fuel in a manner which stages it to final disposition.{close_quotes} The inventory of spent nuclear fuel (SNF) at the Hanford Site covers a wide variety of fuel types (production reactor to space reactor) in many facilities (reactor fuel basins to hot cells) at locations all over the Site. The 2,129 metric tons of Hanford SNF represents about 80% of the total US Department of Energy (DOE) inventory. About 98.5% of the Hanford SNF is 2,100 metric tons of metallic uranium production reactor fuel currently stored in the 1950s vintage K Basins in the 100 Area. This fuel has been slowly corroding, generating sludge and contaminating the basin water. This condition, coupled with aging facilities with seismic vulnerabilities, has been identified by several groups, including stakeholders, as being one of the most urgent safety and environmental concerns at the Hanford Site. As a direct result of these concerns, the Spent Nuclear Fuel Project was recently formed to address spent fuel issues at Hanford. The Project has developed the K Basins Path Forward to remove fuel from the basins and place it in dry interim storage. Alternatives that addressed the requirements were developed and analyzed. The result is a two-phased approach allowing the early removal of fuel from the K Basins followed by its stabilization and interim storage consistent with the national program.

  10. West Valley facility spent fuel handling, storage, and shipping experience

    SciTech Connect (OSTI)

    Bailey, W.J.

    1990-11-01T23:59:59.000Z

    The result of a study on handling and shipping experience with spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory (PNL) and was jointly sponsored by the US Department of Energy (DOE) and the Electric Power Research Institute (EPRI). The purpose of the study was to document the experience with handling and shipping of relatively old light-water reactor (LWR) fuel that has been in pool storage at the West Valley facility, which is at the Western New York Nuclear Service Center at West Valley, New York and operated by DOE. A subject of particular interest in the study was the behavior of corrosion product deposits (i.e., crud) deposits on spent LWR fuel after long-term pool storage; some evidence of crud loosening has been observed with fuel that was stored for extended periods at the West Valley facility and at other sites. Conclusions associated with the experience to date with old spent fuel that has been stored at the West Valley facility are presented. The conclusions are drawn from these subject areas: a general overview of the West Valley experience, handling of spent fuel, storing of spent fuel, rod consolidation, shipping of spent fuel, crud loosening, and visual inspection. A list of recommendations is provided. 61 refs., 4 figs., 5 tabs.

  11. The united kingdom's changing requirements for spent fuel storage

    SciTech Connect (OSTI)

    Hodgson, Z.; Hambley, D.I.; Gregg, R.; Ross, D.N. [National Nuclear Laboratory, Chadwick House, Birchwood Park, Warrington, Cheshire WA3 6AE (United Kingdom)

    2013-07-01T23:59:59.000Z

    The UK is adopting an open fuel cycle, and is necessarily moving to a regime of long term storage of spent fuel, followed by geological disposal once a geological disposal facility (GDF) is available. The earliest GDF receipt date for legacy spent fuel is assumed to be 2075. The UK is set to embark on a programme of new nuclear build to maintain a nuclear energy contribution of 16 GW. Additionally, the UK are considering a significant expansion of nuclear energy in order to meet carbon reduction targets and it is plausible to foresee a scenario where up to 75 GW from nuclear power production could be deployed in the UK by the mid 21. century. Such an expansion, could lead to spent fuel storage and its disposal being a dominant issue for the UK Government, the utilities and the public. If the UK were to transition a closed fuel cycle, then spent fuel storage should become less onerous depending on the timescales. The UK has demonstrated a preference for wet storage of spent fuel on an interim basis. The UK has adopted an approach of centralised storage, but a 16 GW new build programme and any significant expansion of this may push the UK towards distributed spent fuel storage at a number of reactors station sites across the UK.

  12. Impact of PWR spent fuel variations on TRU-fueled VHTRS 

    E-Print Network [OSTI]

    Alajo, Ayodeji Babatunde

    2009-05-15T23:59:59.000Z

    The VHTR is a graphite moderated helium-cooled reactor that supplies heat with core outlet temperatures equal to or greater than 850 degree Celsius. Its basic technology has been well established in High Temperature Gas Reactors (HTGR..., representing an attractive spent fuel form. The prismatic core configuration features annular fuel compacts composed of TRISO coated fuel particles that are embedded in graphite matrix. The fuel compacts are stacked together forming the fuel element...

  13. Hanford`s spent nuclear fuel retrieval: an agressive agenda

    SciTech Connect (OSTI)

    Shen, E.J., Westinghouse Hanford

    1996-12-06T23:59:59.000Z

    Starting December 1997, spent nuclear fuel that has been stored in the K Reactor Fuel Storage Basins will be retrieved over a two year period and repackaged for long term dry storage. The aging and sometimes corroding fuel elements will be recovered and processed using log handled tools and teleoperated manipulator technology. The U.S. Department of Energy (DOE) is committed to this urgent schedule because of the environmental threats to the groundwater and nearby the Columbia River.

  14. COGEMA operating experience in the transportation of spent fuel, nuclear materials and radioactive waste

    SciTech Connect (OSTI)

    Bernard, H. [COGEMA, Velizy-Villacoublay (France)

    1993-12-31T23:59:59.000Z

    Were a spent fuel transportation accident to occur, no matter how insignificant, the public outcry could jeopardize both reprocessing operations and power plant operations for utilities that have elected to reprocess their spent fuel. Aware of this possibility, COGEMA has become deeply involved in spent fuel transportation to ensure that it is performed according to the highest standards of transportation safety. Spent fuel transportation is a vital link between the reactor site and the reprocessing plant. This paper gives an overview of COGEMA`s experience in the transportation of spent fuel.

  15. Recovery of weapon plutonium as feed material for reactor fuel

    SciTech Connect (OSTI)

    Armantrout, G.A.; Bronson, M.A.; Choi, Jor-Shan [and others

    1994-03-16T23:59:59.000Z

    This report presents preliminary considerations for recovering and converting weapon plutonium from various US weapon forms into feed material for fabrication of reactor fuel elements. An ongoing DOE study addresses the disposition of excess weapon plutonium through its use as fuel for nuclear power reactors and subsequent disposal as spent fuel. The spent fuel would have characteristics similar to those of commercial power spent fuel and could be similarly disposed of in a geologic repository.

  16. Spent fuel integrity during dry storage

    SciTech Connect (OSTI)

    McKinnon, M.A.

    1995-07-01T23:59:59.000Z

    Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at the Idaho National Engineering Laboratory (INEL) offers significant opportunities for confirmation of the benign nature of long-term dry storage. The cask performance tests conducted at INEL included visual observation and ultrasonic examination of the condition of cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of the fuel; and a qualitative determination of the effect of dry storage and fuel consolidation on fission gas release from the spent fuel rods. A variety of cover gases and cask orientations were used during the cask performance tests. Cover gases included vacuum, nitrogen, and helium. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the conclusion of each performance test, periodic gas sampling was conducted on each cask as part of a surveillance and monitoring activity. Continued surveillance and monitoring activities are being conducted for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are reported in this paper.

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

    Energy Savers [EERE]

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

  18. Determination of Plutonium Content in Spent Fuel with Nondestructive Assay

    E-Print Network [OSTI]

    Tobin, S. J.

    2010-01-01T23:59:59.000Z

    Spent Nuclear Fuel Recycling Facility – 8032,” Waste ManagementSpent Nuclear Fuel by Self-Induced X-ray,” Annual Meeting of the Institute of Nuclear Material Management,

  19. Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition...

    Office of Environmental Management (EM)

    Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy Lessons Learned Report, NNSA, Feb 2010 Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition...

  20. Annual report, FY 1979 Spent fuel and fuel pool component integrity.

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01T23:59:59.000Z

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-..mu..m) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report.

  1. Apparatus for shearing spent nuclear fuel assemblies

    DOE Patents [OSTI]

    Weil, Bradley S. (Knoxville, TN); Metz, III, Curtis F. (Knoxville, TN)

    1980-01-01T23:59:59.000Z

    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.

  2. Test plan for thermogravimetric analyses of BWR spent fuel oxidation

    SciTech Connect (OSTI)

    Einziger, R.E.

    1988-12-01T23:59:59.000Z

    Preliminary studies indicated the need for additional low-temperature spent fuel oxidation data to determine the behavior of spent fuel as a waste form for a tuffy repository. Short-term thermogravimetric analysis tests were recommended in a comprehensive technical approach as the method for providing scoping data that could be used to (1) evaluate the effects of variables such as moisture and burnup on the oxidation rate, (2) determine operative mechanisms, and (3) guide long-term, low-temperature oxidation testing. The initial test series studied the temperature and moisture effects on pressurized water reactor fuel as a function of particle and grain size. This document presents the test matrix for studying the oxidation behavior of boiling water reactor fuel in the temperature range of 140 to 225{degree}C. 17 refs., 7 figs., 3 tabs.

  3. Pyrochemical processing of DOE spent nuclear fuel

    SciTech Connect (OSTI)

    Laidler, J.J.

    1995-02-01T23:59:59.000Z

    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.

  4. Air Shipment of Spent Nuclear Fuel from Romania to Russia

    SciTech Connect (OSTI)

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01T23:59:59.000Z

    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.

  5. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    SciTech Connect (OSTI)

    P.M. O'Leary; Dr. M.L. Pitts

    2000-08-21T23:59:59.000Z

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers.

  6. Shippingport Spent Fuel Canister System Description

    SciTech Connect (OSTI)

    JOHNSON, D.M.

    2000-03-27T23:59:59.000Z

    In 1978 and 1979, a total of 72 blanket fuel assemblies (BFAs), irradiated during the operating cycles of the Shippingport Atomic Power Station's Pressurized Water Reactor (PWR) Core 2 from April 1965 to February 1974, were transferred to the Hanford Site and stored in underwater storage racks in Cell 2R at the 221-T Canyon (T-Plant). The initial objective was to recover the produced plutonium in the BFAs, but this never occurred and the fuel assemblies have remained within the water storage pool to the present time. The Shippingport Spent Fuel Canister (SSFC) is a confinement system that provides safe transport functions (in conjunction with the TN-WHC cask) and storage for the BFAs at the Canister Storage Building (CSB). The current plan is for these BFAs to be retrieved from wet storage and loaded into SSFCs for dry storage. The sealed SSFCs containing BFAs will be vacuum dried, internally backfilled with helium, and leak tested to provide suitable confinement for the BFAs during transport and storage. Following completion of the drying and inerting process, the SSFCs are to be delivered to the CSB for closure welding and long-term interim storage. The CSB will provide safe handling and dry storage for the SSFCs containing the BFAs. The purpose of this document is to describe the SSFC system and interface equipment, including the technical basis for the system, design descriptions, and operations requirements. It is intended that this document will be periodically updated as more equipment design and performance specification information becomes available.

  7. Characterization of spent fuel approved testing material---ATM-105

    SciTech Connect (OSTI)

    Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01T23:59:59.000Z

    The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

  8. National Report Joint Convention on the Safety of Spent Fuel...

    Office of Environmental Management (EM)

    National Report Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management National Report Joint Convention on the Safety of Spent...

  9. Characterization of spent fuel approved testing material--ATM-104

    SciTech Connect (OSTI)

    Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

    1991-12-01T23:59:59.000Z

    The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

  10. Direct Investigations of the Immobilization of Radionuclides in the Alteration Products of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Peter C. Burns; Robert J. Finch; David J. Wronkiewicz

    2004-12-27T23:59:59.000Z

    Safe disposal of the nation's nuclear waste in a geological repository involves unique scientific and engineering challenges owing to the very long-lived radioactivity of the waste. The repository must retain a variety of radionuclides that have vastly different chemical characters for several thousand years. Most of the radioactivity that will be housed in the proposed repository at Yucca Mountain will be associated with spent nuclear fuel, much of which is derived from commercial reactors. DOE is custodian of approximately 8000 tons of spent nuclear fuel that is also intended for eventual disposal in a geological repository. Unlike the spent fuel from commercial reactors, the DOE fuel is diverse in composition with more than 250 varieties. Safe disposal of spent fuel requires a detailed knowledge of its long-term behavior under repository conditions, as well as the fate of radionuclides released from the spent fuel as waste containers are breached.

  11. Spent Nuclear Fuel (SNF) Project Product Specification

    SciTech Connect (OSTI)

    PAJUNEN, A.L.

    2000-01-20T23:59:59.000Z

    This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

  12. Comparison of spent nuclear fuel management alternatives

    SciTech Connect (OSTI)

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

    1996-09-01T23:59:59.000Z

    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.

  13. Characteristics of fuel crud and its impact on storage, handling, and shipment of spent fuel. [Fuel crud

    SciTech Connect (OSTI)

    Hazelton, R.F.

    1987-09-01T23:59:59.000Z

    Corrosion products, called ''crud,'' form on out-of-reactor surfaces of nuclear reactor systems and are transported by reactor coolant to the core, where they deposit on external fuel-rod cladding surfaces and are activated by nuclear reactions. After discharge of spent fuel from a reactor, spallation of radioactive crud from the fuel rods could impact wet or dry storage operations, handling (including rod consolidation), and shipping. It is the purpose of this report to review earlier (1970s) and more recent (1980s) literature relating to crud, its characteristics, and any impact it has had on actual operations. Crud characteristics vary from reactor type to reactor type, reactor to reactor, fuel assembly to fuel assembly in a reactor, circumferentially and axially in an assembly, and from cycle to cycle for a specific facility. To characterize crud of pressurized-water (PWRs) and boiling-water reactors (BWRs), published information was reviewed on appearance, chemical composition, areal density and thickness, structure, adhesive strength, particle size, and radioactivity. Information was also collected on experience with crud during spent fuel wet storage, rod consolidation, transportation, and dry storage. From experience with wet storage, rod consolidation, transportation, and dry storage, it appears crud spallation can be managed effectively, posing no significant radiological problems. 44 refs., 11 figs.

  14. Heat transfer modeling of dry spent nuclear fuel storage facilities

    SciTech Connect (OSTI)

    Lee, S.Y.

    1999-07-01T23:59:59.000Z

    The present work was undertaken to provide heat transfer model that accurately predicts the thermal performance of dry spent nuclear fuel storage facilities. One of the storage configurations being considered for DOE Aluminum-clad Spent Nuclear Fuel (Al-SNF), such as the Material and Testing Reactor (MTR) fuel, is in a dry storage facility. To support design studies of storage options a computational and experimental program has been conducted at the Savannah River Site (SRS). The main objective is to develop heat transfer models including natural convection effects internal to an interim dry storage canister and to geologic codisposal Waste Package (WP). Calculated temperatures will be used to demonstrate engineering viability of a dry storage option in enclosed interim storage and geologic repository WP and to assess the chemical and physical behaviors of the Al-SNF in the dry storage facilities. The current paper describes the modeling approaches and presents the computational results along with the experimental data.

  15. Heat Transfer Modeling of Dry Spent Nuclear Fuel Storage Facilities

    SciTech Connect (OSTI)

    Lee, S.Y.

    1999-01-13T23:59:59.000Z

    The present work was undertaken to provide heat transfer model that accurately predicts the thermal performance of dry spent nuclear fuel storage facilities. One of the storage configurations being considered for DOE Aluminum-clad Spent Nuclear Fuel (Al-SNF), such as the Material and Testing Reactor (MTR) fuel, is in a dry storage facility. To support design studies of storage options a computational and experimental program has been conducted at the Savannah River Site (SRS). The main objective is to develop heat transfer models including natural convection effects internal to an interim dry storage canister and to geological codisposal Waste Package (WP). Calculated temperatures will be used to demonstrate engineering viability of a dry storage option in enclosed interim storage and geological repository WP and to assess the chemical and physical behaviors of the Al-SNF in the dry storage facilities. The current paper describes the modeling approaches and presents the computational results along with the experimental data.

  16. Spent nuclear fuel project integrated schedule plan

    SciTech Connect (OSTI)

    Squires, K.G.

    1995-03-06T23:59:59.000Z

    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.

  17. TEPP - Spent Nuclear Fuel | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium Transferon the PassingRouting TEC Working GroupEnergy- Spent

  18. Accelerator-driven transmutation of spent fuel elements

    DOE Patents [OSTI]

    Venneri, Francesco (Los Alamos, NM); Williamson, Mark A. (Los Alamos, NM); Li, Ning (Los Alamos, NM)

    2002-01-01T23:59:59.000Z

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  19. Thermomechanical modeling of the Spent Fuel Test-Climax

    SciTech Connect (OSTI)

    Butkovich, T.R.; Patrick, W.C.

    1986-02-01T23:59:59.000Z

    The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C.

  20. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01T23:59:59.000Z

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  1. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    E-Print Network [OSTI]

    Quiter, Brian

    2012-01-01T23:59:59.000Z

    Spent Fuel Assay Using Nuclear Resonance Fluo- rescence,” Annual Meeting of the Institute of Nuclear Material Management,

  2. EIS-0203F; DOE Programmatic Spent Nuclear Fuel Management and...

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

    Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final...

  3. Determination of Plutonium Content in Spent Fuel with Nondestructive Assay

    E-Print Network [OSTI]

    Tobin, S. J.

    2010-01-01T23:59:59.000Z

    for safeguards of LEU and MOX spent fuel,” Internationalsystems in use today (Safeguards Mox Python Detector, 1 Fork

  4. A Monte Carlo based spent fuel analysis safeguards strategy assessment

    SciTech Connect (OSTI)

    Fensin, Michael L [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Sandoval, Nathan P [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

    Safeguarding nuclear material involves the detection of diversions of significant quantities of nuclear materials, and the deterrence of such diversions by the risk of early detection. There are a variety of motivations for quantifying plutonium in spent fuel assemblies by means of nondestructive assay (NDA) including the following: strengthening the capabilities of the International Atomic Energy Agencies ability to safeguards nuclear facilities, shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories. Many NDA techniques exist for measuring signatures from spent fuel; however, no single NDA technique can, in isolation, quantify elemental plutonium and other actinides of interest in spent fuel. A study has been undertaken to determine the best integrated combination of cost effective techniques for quantifying plutonium mass in spent fuel for nuclear safeguards. A standardized assessment process was developed to compare the effective merits and faults of 12 different detection techniques in order to integrate a few techniques and to down-select among the techniques in preparation for experiments. The process involves generating a basis burnup/enrichment/cooling time dependent spent fuel assembly library, creating diversion scenarios, developing detector models and quantifying the capability of each NDA technique. Because hundreds of input and output files must be managed in the couplings of data transitions for the different facets of the assessment process, a graphical user interface (GUI) was development that automates the process. This GUI allows users to visually create diversion scenarios with varied replacement materials, and generate a MCNPX fixed source detector assessment input file. The end result of the assembly library assessment is to select a set of common source terms and diversion scenarios for quantifying the capability of each of the 12 NDA techniques. We present here the generalized assessment process, the techniques employed to automate the coupled facets of the assessment process, and the standard burnup/enrichment/cooling time dependent spent fuel assembly library. We also clearly define the diversion scenarios that will be analyzed during the standardized assessments. Though this study is currently limited to generic PWR assemblies, it is expected that the results of the assessment will yield an adequate spent fuel analysis strategy knowledge that will help the down-select process for other reactor types.

  5. Spent-fuel dry-storage testing at E-MAD (March 1978-March 1982)

    SciTech Connect (OSTI)

    Unterzuber, R.; Milnes, R.D.; Marinkovich, B.A.; Kubancsek, G.M.

    1982-09-01T23:59:59.000Z

    From March 1978 through March 1982, spent fuel dry storage tests were conducted at the Engine Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site to confirm that commercial reactor spent fuel could be encapsulated and passively stored in one or more interim dry storage cell concepts. These tests were: electrically heated drywell, isolated and adjacent drywell, concrete silo, fuel assembly internal temperature measurement, and air-cooled vault. This document presents the test data and results as well as results from supporting test operations (spent fuel calorimetry and canister gas sampling).

  6. Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions

    SciTech Connect (OSTI)

    Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

    1993-01-01T23:59:59.000Z

    The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO{sub 2} pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel.

  7. Nondestructive Spent Fuel Assay Using Nuclear Resonance Fluorescence

    E-Print Network [OSTI]

    Quiter, Brian

    2010-01-01T23:59:59.000Z

    Spent Fuel Library for Assessing Varied Nondestructive Assay Techniques for Nuclear Safeguards," LA-UR 09-01188, ANS Advances in Nuclear Fuel Management

  8. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect (OSTI)

    Douglas Morrell

    2011-03-01T23:59:59.000Z

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  9. Savannah River Site, Spent Nuclear Fuel Management, Draft Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    1998-12-24T23:59:59.000Z

    The proposed DOE action described in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets assigned to the Savannah River Site (SRS), including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel (20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some programmatic material stored at SRS for repackaging and dry storage pending shipment offsite).

  10. Contaminated sediment removal from a spent fuel storage canal

    SciTech Connect (OSTI)

    Geber, K R

    1993-01-01T23:59:59.000Z

    A leaking underground spent fuel transfer canal between a decommissioned reactor and a radiochemical separations building at the Oak Ridge National Laboratory (ORNL) was found to contain RCRA-hazardous and radioactive sediment. Closure of the Part B RCRA permitted facility required the use of an underwater robotic vacuum and a filtration-containment system to separate and stabilize the contaminated sediment. This paper discusses the radiological controls established to maintain contamination and exposures As Low As Reasonably Achievable (ALARA) during the sediment removal.

  11. Shipper/receiver difference verification of spent fuel by use of PDET

    SciTech Connect (OSTI)

    Ham, Y. S.; Sitaraman, S. [Global Security Directorate, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

    2011-07-01T23:59:59.000Z

    Spent fuel storage pools in most countries are rapidly approaching their design limits with the discharge of over 10,000 metric tons of heavy metal from global reactors. Countries like UK, France or Japan have adopted a closed fuel cycle by reprocessing spent fuel and recycling MOX fuel while many other countries opted for above ground interim dry storage for their spent fuel management strategy. Some countries like Finland and Sweden are already well on the way to setting up a conditioning plant and a deep geological repository for spent fuel. For all these situations, shipments of spent fuel are needed and the number of these shipments is expected to increase significantly. Although shipper/receiver difference (SRD) verification measurements are needed by IAEA when the recipient facility receives spent fuel, these are not being practiced to the level that IAEA has desired due to lack of a credible measurement methodology and instrument that can reliably perform these measurements to verify non-diversion of spent fuel during shipment and confirm facility operator declarations on the spent fuel. In this paper, we describe a new safeguards method and an associated instrument, Partial Defect Tester (PDET), which can detect pin diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies in an in-situ condition. The PDET uses multiple tiny neutron and gamma detectors in the form of a cluster and a simple, yet highly precise, gravity-driven system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly. The method takes advantage of the PWR fuel design which contains multiple guide tubes which can be accessed from the top. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. Our simulation study as well as validation measurements indicated that the ratio of the gamma signal to the thermal neutron signal at each detector location normalized to the peak ratio of all the detector locations gives a unique signature that is sensitive to missing pins. The signature is principally dependent on the geometry of the detector locations, and little sensitive to enrichment or burn-up variations. A small variation in the fuel bundle, such as a few missing pins, changes the shape of the signature to enable detection. After verification of the non-diversion of spent fuel pins, the neutron signal and gamma signal are subsequently used to verify the consistency of the operator declaration on the fuel burn-up and cooling time. (authors)

  12. Standard guide for characterization of spent nuclear fuel in support of geologic repository disposal

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2009-01-01T23:59:59.000Z

    1.1 This guide provides guidance for the types and extent of testing that would be involved in characterizing the physical and chemical nature of spent nuclear fuel (SNF) in support of its interim storage, transport, and disposal in a geologic repository. This guide applies primarily to commercial light water reactor (LWR) spent fuel and spent fuel from weapons production, although the individual tests/analyses may be used as applicable to other spent fuels such as those from research and test reactors. The testing is designed to provide information that supports the design, safety analysis, and performance assessment of a geologic repository for the ultimate disposal of the SNF. 1.2 The testing described includes characterization of such physical attributes as physical appearance, weight, density, shape/geometry, degree, and type of SNF cladding damage. The testing described also includes the measurement/examination of such chemical attributes as radionuclide content, microstructure, and corrosion product c...

  13. Pyroprocess for processing spent nuclear fuel

    DOE Patents [OSTI]

    Miller, William E. (Naperville, IL); Tomczuk, Zygmunt (Lockport, IL)

    2002-01-01T23:59:59.000Z

    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.

  14. Systems impacts of spent fuel disassembly alternatives

    SciTech Connect (OSTI)

    Not Available

    1984-07-01T23:59:59.000Z

    Three studies were completed to evaluate four alternatives to the disposal of intact spent fuel assemblies in a geologic repository. A preferred spent fuel waste form for disposal was recommended on consideration of (1) package design and fuel/package interaction, (2) long-term, in-repository performance of the waste form, and (3) overall process performance and costs for packaging, handling, and emplacement. The four basic alternative waste forms considered were (1) end fitting removal, (2) fission gas venting, (3) disassembly and close packing, and (4) shearing/immobilization. None of the findings ruled out any alternative on the basis of waste package considerations or long-term performance of the waste form. The third alternative offers flexibility in loading that may prove attractive in the various geologic media under consideration, greatly reduces the number of packages, and has the lowest unit cost. These studies were completed in October, 1981. Since then Westinghouse Electric Corporation and the Office of Nuclear Waste Isolation have completed studies in related fields. This report is now being published to provide publicly the background material that is contained within. 47 references, 28 figures, 31 tables.

  15. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect (OSTI)

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

    2014-01-01T23:59:59.000Z

    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.

  16. Potential importance of water pathways for spent fuel transportation accident risk

    SciTech Connect (OSTI)

    Ostmeyer, R.M.

    1986-01-01T23:59:59.000Z

    This paper analyzes the potential importance of water pathway contamination for spent fuel transportation accident risk using a ''worst-case'' water contamination scenario. The scenario used for the analysis involves an accident release that occurs near a reservoir. Water pathway doses are compared to doses for accident releases in urban or agricultural areas. The results of the analysis indicate that water pathways are not important for assessing the risk of transporting spent reactor fuel by truck or by rail.

  17. Cost probability analysis of reprocessing spent nuclear fuel in the US G.D. Recktenwald, M.R. Deinert

    E-Print Network [OSTI]

    Deinert, Mark

    and curium) and these would then be used to manufacture fuel for a fleet of fast burner reactors. The spent fuel from the fast reactors would itself be reprocessed, thereby creating a closed loop in which time, neither the fast reactor technologies, nor the required reprocessing technologies, are mature

  18. RADIATION ANALYSIS OF A SPENT-FUEL STORAGE CASK

    E-Print Network [OSTI]

    Shultis, J. Kenneth

    RADIATION ANALYSIS OF A SPENT-FUEL STORAGE CASK by J.K. Shultis Department of Mechanical;Radiation Analysis of a Spect-Fuel Storage Cask by J.K.Shultis Dept. Mechanical and Nuclear Engineering a single Transnuclear spent-fuel storage cask holding 68 design-basis fuel assemblies (a TN-68 cask

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

    SciTech Connect (OSTI)

    Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1994-10-01T23:59:59.000Z

    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. Spent fuel management fee methodology and computer code user's manual.

    SciTech Connect (OSTI)

    Engel, R.L.; White, M.K.

    1982-01-01T23:59:59.000Z

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively.

  1. Physical and decay characteristics of commercial LWR spent fuel

    SciTech Connect (OSTI)

    Roddy, J.W.; Claiborne, H.C.; Ashline, R.C.; Johnson, P.J.; Rhyne, B.T.

    1986-01-01T23:59:59.000Z

    Information was collected from the literature and from major manufacturers that will be useful in the design and construction of a mined geologic repository for the disposal of light-water-reactor spent fuel. Pertinent data are included on mechanical design characteristics and materials of construction for fuel assemblies and fuel rods and computed values for heat generation rates, radioactivity, and photon and neutron emission rates as a function of time for four reference cases. Calculations were made with the ORIGEN2 computer code for burnups of 27,500 and 40,000 MWd for a typical boiling-water reactor and 33,000 and 60,000 MWd for a typical pressurized-water reactor. The results are presented in figures depicting the individual contributions per metric ton of initial heavy metal for the activation products, fission products, and actinides and their daughters to the radioactivity and thermal power as a function of time. Tables are also presented that list the contribution of each major nuclide to the radioactivity, thermal power, and photons and neutrons emitted for disposal emitted for disposal periods from 1 to 100,000 years.

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

    National Nuclear Security Administration (NNSA)

    Reagan Calls for a National Spent Fuel Storage Facility | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile...

  3. Huizenga leads safety of spent fuel management, radioactive waste...

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

    Huizenga leads safety of spent fuel management, radioactive waste management meeting in Vienna | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People...

  4. Spent Nuclear Fuel project integrated safety management plan

    SciTech Connect (OSTI)

    Daschke, K.D.

    1996-09-17T23:59:59.000Z

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

  5. Assessment of the safety of spent fuel transportation in urban environs

    SciTech Connect (OSTI)

    Sandoval, R.P.; Weber, J.P.; Levine, H.S.; Romig, A.D.; Johnson, J.D.; Luna, R.E.; Newton, G.J.; Wong, B.A.; Marshall, R.W. Jr.; Alvarez, J.L.

    1983-06-01T23:59:59.000Z

    The results of a program to provide an experimental data base for estimating the radiological consequences from a hypothetical sabotage attack on a light-water-reactor spent fuel shipping cask in a densely populated area are presented. The results of subscale and full-scale experiments in conjunction with an analytical modeling study are described. The experimental data were used as input to a reactor-safety consequence model to predict radiological health consequences resulting from a hypothetical sabotage attack on a spent-fuel shipping cask in the Manhattan borough of New York City. The results of these calculations are presented.

  6. Spent fuel drying system test results (second dry-run)

    SciTech Connect (OSTI)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01T23:59:59.000Z

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks have been detected in the basins and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the second dry-run test, which was conducted without a fuel element. With the concurrence of project management, the test protocol for this run, and subsequent drying test runs, was modified. These modifications were made to allow for improved data correlation with drying procedures proposed under the IPS. Details of these modifications are discussed in Section 3.0.

  7. Spent nuclear fuel recycling with plasma reduction and etching

    DOE Patents [OSTI]

    Kim, Yong Ho

    2012-06-05T23:59:59.000Z

    A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

  8. Dry oxidation and fracture of LWR spent fuels

    SciTech Connect (OSTI)

    Ahn, T.M.

    1996-11-01T23:59:59.000Z

    This report evaluates the characteristics of oxidation and fracture of light-water reactor (LWR) spent fuel in dry air. It also discusses their effects on radionuclide releases in the anticipated high-level waste repository environment. A sphere model may describe diffusion-limited formation of lower oxides, such as U{sub 4}O{sub 9}, in the oxidation of the spent fuel (SF) matrix. Detrimental higher oxides, such as U{sub 3}O{sub 8}, may not form at temperatures below a threshold temperature. The nucleation process suggests that a threshold temperature exists. The calculated results regarding fracture properties of the SF matrix agree with experimental observations. Oxidation and fracture of Zircaloy may not be significant under anticipated conditions. Under saturated or unsaturated aqueous conditions, oxidation of the SF matrix is believed to increase the releases of Pu-(239+240), Am-(241+243), C-14, Tc-99, I-129, and Cs-135. Under dry conditions, I-129 releases are likely to be small, unlike C-14, in lower oxides; Cl-36, Tc-99, I-129, and Cs-135 may be released fast in higher oxides. 79 refs.

  9. Facts and issues of direct disposal of spent fuel; Revision 1

    SciTech Connect (OSTI)

    Parks, P.B.

    1993-10-01T23:59:59.000Z

    This report reviews those facts and issues that affect the direct disposal of spent reactor fuels. It is intended as a resource document for those impacted by the current Department of Energy (DOE) guidance that calls for the cessation of fuel reprocessing. It is not intended as a study of the specific impacts (schedules and costs) to the Savannah River Site (SRS) alone. Commercial fuels, other low enriched fuels, highly enriched defense-production, research, and naval reactor fuels are included in this survey, except as prevented by rules on classification.

  10. Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

    SciTech Connect (OSTI)

    McKinnon, M.A.; Doherty, A.L.

    1997-06-01T23:59:59.000Z

    This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs.

  11. Proliferation Resistant Nuclear Reactor Fuel

    SciTech Connect (OSTI)

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

    2011-02-18T23:59:59.000Z

    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 we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

  12. RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

    SciTech Connect (OSTI)

    Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

    2009-07-01T23:59:59.000Z

    In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

  13. Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences

    SciTech Connect (OSTI)

    Bailey, W.J.

    1990-02-01T23:59:59.000Z

    The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

  14. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J. (Los Alamos, NM)

    1987-01-01T23:59:59.000Z

    A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

  15. Managing Spent Nuclear Fuel at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Thomas Hill; Denzel L. Fillmore

    2005-10-01T23:59:59.000Z

    The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many forms—from intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

  16. Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495

    SciTech Connect (OSTI)

    Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine)] [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine); Schmieman, Eric [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)] [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)

    2013-07-01T23:59:59.000Z

    Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

  17. Safe Advantage on Dry Interim Spent Nuclear Fuel Storage

    SciTech Connect (OSTI)

    Romanato, L.S. [Centro Tecnologico da Marinha em S.Paulo, Brazilian Navy Technological Center, Sao Paulo (Brazil)

    2008-07-01T23:59:59.000Z

    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)

  18. Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

    SciTech Connect (OSTI)

    Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

    2003-02-25T23:59:59.000Z

    Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research.

  19. Thermal Cooling Limits of Sbotaged Spent Fuel Pools

    SciTech Connect (OSTI)

    Dr. Thomas G. Hughes; Dr. Thomas F. Lin

    2010-09-10T23:59:59.000Z

    To develop the understanding and predictive measures of the post “loss of water inventory” hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others

  20. Spent fuel dry storage technology development: thermal evaluation of sealed storage cask containing spent fuel

    SciTech Connect (OSTI)

    Schmitten, P.F.; Wright, J.B.

    1980-08-01T23:59:59.000Z

    A PWR spent fuel assembly was encapsulated inside the E-MAD Hot Bay and placed in a instrumented above surface storage cell during December 1978 for thermal testing. Instrumentation provided to measure canister, liner and concrete temperatures consisted of thermocouples which were inserted into tubes on the outside of the canister and liner and in three radial positions in the concrete. Temperatures from the SSC test assembly have been recorded throughout the past 16 months. Canister and liner temperatures have reached their peak values of 200{sup 0}F and 140{sup 0}F, respectively. Computer predictions of the transient and steady-state temperatures show good agreement with the test data.

  1. Spent Nuclear Fuel (SNF) Project Product Specification

    SciTech Connect (OSTI)

    PAJUNEN, A.L.

    2000-12-07T23:59:59.000Z

    The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major sub-system. Section 4.0--Specific technical basis description for each product specification. The scope of this product specification does not include data collection requirements to support accountability or environmental compliance activities.

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

    SciTech Connect (OSTI)

    Wood, Thomas W.; Rothwell, Geoffrey

    2012-01-10T23:59:59.000Z

    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.

  3. Benefits/impacts of utilizing depleted uranium silicate glass as backfill for spent fuel waste packages

    SciTech Connect (OSTI)

    Pope, R.B.; Forsberg, C.W.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

    1996-05-01T23:59:59.000Z

    An assessment has been made of the benefits and impacts which can be derived by filling a spent nuclear fuel multi-purpose canister with depleted uranium silicate (DUS) glass at a reactor site. Although the primary purpose of the DUS glass fill would be to enhance repository performance assessment and control criticality of geologic times, a number of benefits to the waste management system can be derived from adding the DUS glass prior to shipment from the reactor site.

  4. Neutron Generators for Spent Fuel Assay

    SciTech Connect (OSTI)

    Ludewigt, Bernhard A

    2010-12-30T23:59:59.000Z

    The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel (SNF) assemblies with non-destructive assay (NDA). The 14 NDA techniques being studied include several that require an external neutron source: Delayed Neutrons (DN), Differential Die-Away (DDA), Delayed Gammas (DG), and Lead Slowing-Down Spectroscopy (LSDS). This report provides a survey of currently available neutron sources and their underlying technology that may be suitable for NDA of SNF assemblies. The neutron sources considered here fall into two broad categories. The term 'neutron generator' is commonly used for sealed devices that operate at relatively low acceleration voltages of less than 150 kV. Systems that employ an acceleration structure to produce ion beam energies from hundreds of keV to several MeV, and that are pumped down to vacuum during operation, rather than being sealed units, are usually referred to as 'accelerator-driven neutron sources.' Currently available neutron sources and future options are evaluated within the parameter space of the neutron generator/source requirements as currently understood and summarized in section 2. Applicable neutron source technologies are described in section 3. Commercially available neutron generators and other source options that could be made available in the near future with some further development and customization are discussed in sections 4 and 5, respectively. The pros and cons of the various options and possible ways forward are discussed in section 6. Selection of the best approach must take a number of parameters into account including cost, size, lifetime, and power consumption, as well as neutron flux, neutron energy spectrum, and pulse structure that satisfy the requirements of the NDA instrument to be built.

  5. The effect of fuel type in unsaturated spent fuel tests

    SciTech Connect (OSTI)

    Finn, P.A.; Gong, M.; Bates, J.K.; Emery, J.W.; Hoh, J.C.

    1994-04-01T23:59:59.000Z

    Two well-characterized types of spent nuclear fuel (ATM-103 and ATM-106) were tested under simulated unsaturated conditions with simulated groundwater at 90{degree}C. The actinides present in the leachate were measured after periods of approximately 60, 120, and 275 days. The vessels were acid stripped after 120 and 275 days. Both colloidal and soluble actinide species were detected in the leachates which had pHs ranging from 4 to 7. Alpha spectroscopy studies of filtered and unfiltered leachates showed that large amounts of actinides may be bound in colloids. The uranium phases identified in the colloids were schoepite and soddyite. The actinide release behavior of the two fuels appears to be different. The ATM-106 fuel began to release actinides later than the ATM-103 fuel, but after 275 days, it had released more. The amount of americium released from the two fuels was a higher percentage of the maximum amount of americium present than was the percentage of the simultaneous amount of uranium released.

  6. Testing and analyses of the TN-24P PWR spent-fuel dry storage cask loaded with consolidated fuel

    SciTech Connect (OSTI)

    McKinnon, M A; Michener, T E; Jensen, M F; Rodman, G R

    1989-02-01T23:59:59.000Z

    A performance test of a Transnuclear, Inc. TN-24P storage cask configured for pressurized water reactor (PWR) spent fuel was performed. The work was performed by the Pacific Northwest Laboratory (PNL) and Idaho National Engineering Laboratory (INEL) for the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) and the Electric Power Research Institute. The performance test consisted of loading the TN-24P cask with 24 canisters of consolidated PWR spent fuel from Virginia Power's Surry and Florida Power and Light's Turkey Point reactors. Cask surface and fuel canister guide tube temperatures were measured, as were cask surface gamma and neutron dose rates. Testing was performed with vacuum, nitrogen, and helium backfill environments in both vertical and horizontal cask orientations. Transnuclear, Inc., arranged to have a partially insulated run added to the end of the test to simulate impact limiters. Limited spent fuel integrity data were also obtained. From both heat transfer and shielding perspectives, the TN-24P cask with minor refinements can be effectively implemented at reactor sites and central storage facilities for safe storage of unconsolidated and consolidated spent fuel. 35 refs., 93 figs., 17 tabs.

  7. Effect of Helium Accumulation on the Spent Fuel Microstructure

    SciTech Connect (OSTI)

    Ferry, Cecile [Department of Physico-chemistry, Commissariat a l'Energie Atomique, CEA-Saclay, Gif-sur-Yvette, 91191 (France); Piron, Jean-Paul [Commisariat a l'Energie Atomique, CEA - Cadarache, Saint-Paul Lez Durance, 13108 (France); Stout, Ray [Rho Beta Sigma Affaires, Livermore, CA, CA 94550 (United States)

    2007-07-01T23:59:59.000Z

    In a nuclear spent fuel repository, the aqueous rapid release of radio-activity from exposed spent fuel surfaces will depend on the pellet microstructure at the arrival time of water into the disposal container. Research performed on spent fuel evolution in a closed system has shown that the evolution of microstructure under disposal conditions should be governed by the cumulated {alpha}-decay damage and the subsequent helium behavior. The evolution of fission gas bubble characteristics under repository conditions has to be assessed. In UO{sub 2} fuels with a burnup of 47.5 GWd/t, the pressure in fission gas bubbles, including the pressure increase from {alpha}-decay helium atoms, is not expected to reach the critical bubble pressure that will cause failure, thus micro-cracking in UO{sub 2} spent fuel grains is not expected. (authors)

  8. Application of ALARA principles to shipment of spent nuclear fuel

    SciTech Connect (OSTI)

    Greenborg, J.; Brackenbush, L.W.; Murphy, D.W. Burnett, R.A.; Lewis, J.R.

    1980-05-01T23:59:59.000Z

    The public exposure from spent fuel shipment is very low. In view of this low exposure and the perfect safety record for spent fuel shipment, existing systems can be considered satisfactory. On the other hand, occupational exposure reduction merits consideration and technology improvement to decrease dose should concentrate on this exposure. Practices that affect the age of spent fuel in shipment and the number of times the fuel must be shipped prior to disposal have the largest impact. A policy to encourage a 5-year spent fuel cooling period prior to shipment coupled with appropriate cask redesign to accommodate larger loads would be consistent with ALARA and economic principles. And finally, bypassing high population density areas will not in general reduce shipment dose.

  9. DOE-EIS-0218F Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel (February 1996)

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA Approved:AdministrationAnalysisDarby/%2AO 474.2 Chg U.S. S p e chonors PECASE

  10. Spent Nuclear Fuel Project FY 1996 Multi-Year Program Plan WBS No. 1.4.1, Revision 1

    SciTech Connect (OSTI)

    NONE

    1995-09-01T23:59:59.000Z

    This document describes the Spent Nuclear Fuel (SNF) Project portion of the Hanford Strategic Plan for the Hanford Reservation in Richland, Washington. The SNF Project was established to evaluate and integrate the urgent risks associated with N-reactor fuel currently stored at the Hanford site in the K Basins, and to manage the transfer and disposition of other spent nuclear fuels currently stored on the Hanford site. An evaluation of alternatives for the expedited removal of spent fuels from the K Basin area was performed. Based on this study, a Recommended Path Forward for the K Basins was developed and proposed to the U.S. DOE.

  11. area spent fuel: Topics by E-print Network

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

    2009-05-15 8 Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography CERN Preprints Summary: This paper explores the use of cosmic ray muons to image the...

  12. atucha spent fuel: Topics by E-print Network

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

    2009-05-15 8 Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography CERN Preprints Summary: This paper explores the use of cosmic ray muons to image the...

  13. ahwr spent fuel: Topics by E-print Network

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

    2009-05-15 8 Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography CERN Preprints Summary: This paper explores the use of cosmic ray muons to image the...

  14. Risk and Responsibility Sharing in Nuclear Spent Fuel Management

    E-Print Network [OSTI]

    De Roo, Guillaume

    With the Nuclear Waste Policy Act of 1982, the responsibility of American utilities in the long-term management of spent nuclear fuel was limited to the payment of a fee. This narrow involvement did not result in faster ...

  15. Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559

    SciTech Connect (OSTI)

    Botsch, W.; Smalian, S.; Hinterding, P. [TUV NORD Nuclear c/o TUV NORD EnSys Hannover GmbH and Co.KG, Dept. Radiation Protection and Waste Disposal, Am TueV 1, 30519 Hannover (Germany)] [TUV NORD Nuclear c/o TUV NORD EnSys Hannover GmbH and Co.KG, Dept. Radiation Protection and Waste Disposal, Am TueV 1, 30519 Hannover (Germany); Voelzke, H.; Wolff, D.; Kasparek, E. [BAM Federal Institute for Materials Research and Testing Division 3.4 Safety of Storage Containers Unter den Eichen 44-46, 12203 Berlin (Germany)] [BAM Federal Institute for Materials Research and Testing Division 3.4 Safety of Storage Containers Unter den Eichen 44-46, 12203 Berlin (Germany)

    2013-07-01T23:59:59.000Z

    Dry storage systems are characterized by passive and inherent safety systems ensuring safety even in case of severe incidents or accidents. After the events of Fukushima, the advantages of such passively and inherently safe dry storage systems have become more and more obvious. As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Following safety aspects must be achieved throughout the storage period: - safe enclosure of radioactive materials, - safe removal of decay heat, - securing nuclear criticality safety, - avoidance of unnecessary radiation exposure. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. Furthermore, transport capability must be guaranteed during and after storage as well as limitation and control of radiation exposure. The safe enclosure of radioactive materials in dry storage casks can be achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat must be ensured by the design of the storage containers and the storage facility. The safe confinement of radioactive inventory has to be ensured by mechanical integrity of fuel assembly structures. This is guaranteed, e.g. by maintaining the mechanical integrity of the fuel rods or by additional safety measures for defective fuel rods. In order to ensure nuclear critically safety, possible effects of accidents have also to be taken into consideration. In case of dry storage it might be necessary to exclude the re-positioning of fissile material inside the container and/or neutron moderator exclusion might be taken into account. Unnecessary radiation exposure can be avoided by the cask or canister vault system itself. In Germany dry storage of SF in casks fulfills both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground, but due to regional constraints, one storage facility has also been built as a rock tunnel. The decay heat is always removed by natural air flow; further technical equipment is not needed. The removal of decay heat and shielding had been modeled and calculated by state-of-the-art computer codes before such a facility has been built. TueV and BAM present their long experience in the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel. Different storage systems and facilities in Germany, Europe and world-wide are compared with respect to the safety aspects mentioned above. Initial points are the safety issues of wet storage of SF, and it is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. The German storage concept for dry storage of SF and HLW is presented and discussed. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

  16. Evaluation of potential for MSRE spent fuel and flush salt storage and treatment at the INEL

    SciTech Connect (OSTI)

    Ougouag, A.M.; Ostby, P.A.; Nebeker, R.L.

    1996-09-01T23:59:59.000Z

    The potential for interim storage as well as for treatment of the Molten Salt Reactor Experiment spent fuel at INEL has been evaluated. Provided that some minimal packaging and chemical stabilization prerequisites are satisfied, safe interim storage of the spent fuel at the INEL can be achieved in a number of existing or planned facilities. Treatment by calcination in the New Waste Calcining Facility at the INEL can also be a safe, effective, and economical alternative to treatment that would require the construction of a dedicated facility. If storage at the INEL is chosen for the Molten Salt Reactor Experiment (MSRE) spent fuel salts, their transformation to the more stable calcine solid would still be desirable as it would result in a lowering of risks. Treatment in the proposed INEL Remote-Handled Immobilization Facility (RHIF) would result in a waste form that would probably be acceptable for disposal at one of the proposed national repositories. The cost increment imputable to the treatment of the MSRE salts would be a small fraction of the overall capital and operating costs of the facility or the cost of building and operating a dedicated facility. Institutional and legal issues regarding shipments of fuel and waste to the INEL are summarized. The transfer of MSRE spent fuel for interim storage or treatment at the INEL is allowed under existing agreements between the State of idaho and the Department of energy and other agencies of the Federal Government. In contrast, current agreements preclude the transfer into Idaho of any radioactive wastes for storage or disposal within the State of Idaho. This implies that wastes and residues produced from treating the MSRE spent fuel at locations outside Idaho would not be acceptable for storage in Idaho. Present agreements require that all fuel and high-level wastes stored at the INEL, including MSRE spent fuel if received at the INEL, must be moved to a location outside Idaho by the year 2035.

  17. Spent nuclear fuel Canister Storage Building CDR Review Committee report

    SciTech Connect (OSTI)

    Dana, W.P.

    1995-12-01T23:59:59.000Z

    The Canister Storage Building (CSB) is a subproject under the Spent Nuclear Fuels Major System Acquisition. This subproject is necessary to design and construct a facility capable of providing dry storage of repackaged spent fuels received from K Basins. The CSB project completed a Conceptual Design Report (CDR) implementing current project requirements. A Design Review Committee was established to review the CDR. This document is the final report summarizing that review

  18. Conditioning of spent nuclear fuel for permanent disposal

    SciTech Connect (OSTI)

    Laidler, J.J.

    1994-10-01T23:59:59.000Z

    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 preclude the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory.

  19. MELCOR model for an experimental 17x17 spent fuel PWR assembly.

    SciTech Connect (OSTI)

    Cardoni, Jeffrey

    2010-11-01T23:59:59.000Z

    A MELCOR model has been developed to simulate a pressurized water reactor (PWR) 17 x 17 assembly in a spent fuel pool rack cell undergoing severe accident conditions. To the extent possible, the MELCOR model reflects the actual geometry, materials, and masses present in the experimental arrangement for the Sandia Fuel Project (SFP). The report presents an overview of the SFP experimental arrangement, the MELCOR model specifications, demonstration calculation results, and the input model listing.

  20. Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement

    SciTech Connect (OSTI)

    N /A

    2000-04-14T23:59:59.000Z

    The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

  1. The Evolution of Dry Spent Fuel Storage in the United States

    SciTech Connect (OSTI)

    McGough, M.S. [Duratek Inc., 695 Bamesley Lane, Alpharetta, GA 30022 (United States); Bland, D.W. [TriVis, Inc., 1001 Yeager Parkway, Pelham, AL 35124 (United States)

    2006-07-01T23:59:59.000Z

    This paper reviews the evolution of Dry Spent Fuel storage technology and application in the United States. Dating back to the legislation signed by Jimmy Carter on April 7, 1977, to outlaw spent fuel reprocessing, the nations spent fuel pools are gradually becoming filled to capacity. This has necessitated the development of new technologies to store spent fuel in dry casks, predominantly at nuclear power plant sites, awaiting the availability of the federal repository at Yucca Mountain. Site-specific conditions and changes in types of fuel being discharged from reactors have driven a constant evolution of technologies to support this critical need. This paper provides an overview of those changes, which have influenced the evolution of dry storage technology. Focus is provided more towards current technology and cask loading practices, as opposed to those technologies, which are no longer in heavy use. Detailed pictorial material is presented showing the loading sequences of various systems in current use. This paper provides a critical primer on Dry Spent Fuel Storage technology. It provides anyone who is new to dry storage, or who is contemplating initiating dry storage at a nuclear plant site, with useful background and history upon which to build programmatic decisions. (authors)

  2. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01T23:59:59.000Z

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  3. Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.

    1992-04-01T23:59:59.000Z

    This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

  4. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

    The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

  5. Spent fuel drying system test results (first dry-run)

    SciTech Connect (OSTI)

    Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

    1998-07-01T23:59:59.000Z

    The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first dry-run test, which was conducted without a fuel element. The empty test apparatus was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The data from this dry-run test can serve as a baseline for the first two fuel element tests, 1990 (Run 1) and 3128W (Run 2). The purpose of this dry-run was to establish the background levels of hydrogen in the system, and the hydrogen generation and release characteristics attributable to the test system without a fuel element present. This test also serves to establish the background levels of water in the system and the water release characteristics. The system used for the drying test series was the Whole Element Furnace Testing System, described in Section 2.0, which is located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in section 3.0, and the experimental results provided in Section 4.0. These results are further discussed in Section 5.0.

  6. Carbon-14 in waste packages for spent fuel in a tuff repository

    SciTech Connect (OSTI)

    Van Konynenburg, R.A.; Smith, C.F.; Culham, H.W.; Smith, H.D.

    1986-10-14T23:59:59.000Z

    Carbon-14 is produced naturally by cosmic ray neutrons in the upper atmosphere. It is also produced in nuclear reactors, in amounts much smaller than the global inventory. About one-third of this is released directly to the atmosphere, and the other two-thirds remains in the spent fuel. Both the Environmental Protection Agency and the Nuclear Regulatory Commission have established limits on release of the {sup 14}C in spent fuel. This is of particular concern for the proposed repository in tuff, because of the unsaturated conditions and the consequent possibility of gaseous transport of {sup 14}C as CO{sub 2}. Existing measurements and calculations of the {sup 14}C inventory in spent fuel are reviewed. The physical distribution and chemical forms of the {sup 14}C are discussed. Available data on the release of {sup 14}C from spent fuel in aqueous solutions and in gaseous environments of air, nitrogen, and helium are reviewed. Projected {sup 14}C behavior in a tuff repository is described. It is concluded that {sup 14}C release measurements from spent fuel into moist air at temperatures both above and below the in situ boiling point of water as well as detailed transport calculations for the tuff geological environment will be needed to determine whether the 10CFR60 and 40CFR191 requirements can be met. 56 refs., 1 tab.

  7. Dry air oxidation kinetics of K-Basin spent nuclear fuel

    SciTech Connect (OSTI)

    Abrefah, J.; Buchanan, H.C.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.

    1998-06-01T23:59:59.000Z

    The safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor spent nuclear fuel (SNF) stored at K-Basin to an interim storage facility require information about the oxidation behavior of the metallic uranium. Limited experiments have been performed on the oxidation reaction of SNF samples taken from an N-Reactor outer fuel element in various atmospheres. This report discusses studies on the oxidation behavior of SNF using two independent experimental systems: (1) a tube furnace with a flowing gas mixture of 2% oxygen/98% argon; and (2) a thermogravimetric system for dry air oxidation.

  8. Analysis of the risk of transporting spent nuclear fuel by train

    SciTech Connect (OSTI)

    Elder, H.K.

    1981-09-01T23:59:59.000Z

    This report uses risk analyses to analyze the safety of transporting spent nuclear fuel for commercial rail shipping systems. The rail systems analyzed are those expected to be used in the United States when the total electricity-generating capacity by nuclear reactors is 100 GW in the late 1980s. Risk as used in this report is the product of the probability of a release of material to the environment and the consequences resulting from the release. The analysis includes risks in terms of expected fatalities from release of radioactive materials due to transportation accidents involving PWR spent fuel shipped in rail casks. The expected total risk from such shipments is 1.3 x 10/sup -4/ fatalities per year. Risk spectrums are developed for shipments of spent fuel that are 180 days and 4 years out-of-reactor. The risk from transporting spent fuel by train is much less (by 2 to 4 orders of magnitude) than the risk to society from other man-caused events such as dam failure.

  9. Modeling of Spent Fuel Oxidation at Low Temperature

    SciTech Connect (OSTI)

    Poulesquen, Arnaud; Ferry, Cecile [Department of Physico Chemistry, CEA, Gif sur Yvette, 91191 (France); Desgranges, Lionel [Department of Spent Fuel Studies, CEA, St Paul lez Durance, 13108 (France)

    2007-07-01T23:59:59.000Z

    During dry storage, the oxidation of the spent fuel in case of cladding and container failure (accidental scenario) could be detrimental for further handling of the spent fuel rod and for the safety of the facilities. Depending on whether the uranium dioxide is under the form of powder or pellet, irradiated or unirradiated, the weight gain curves do not present the same shape. To account for these different behaviours, two models have been developed. Firstly, the oxidation of unirradiated powders has been modelled based on the coexistence, during the oxidation, of two intermediate products, U{sub 4}O{sub 9} and U{sub 3}O{sub 7}. The comparison between the calculation and the literature data is good in terms of weight gain curves and chemical diffusion coefficient of oxygen within the two phases. Secondly, the oxidation of spent fuel fragments is approached by a convolution procedure between a grain oxidation model and an empirical parameter which represents the linear oxidation speed of grain boundary or an average distance able to cover the entire spent fuel fragment. This procedure of calculation allows in one hand to account for the incubation period noticed on unirradiated pellets or spent fuel and in another hand to link the empirical parameter to physical as porosity, cracks or linear power, or operational parameters such as fission gas release (FGR) respectively. A comparison of this new modelling with experimental data will be proposed. (authors)

  10. Shippingport Spent Fuel Canister (SSFC) Design Report Project W-518

    SciTech Connect (OSTI)

    JOHNSON, D.M.

    2000-01-27T23:59:59.000Z

    The SSFC Design Report Describes A spent fuel canister for Shippingport Core 2 blanket fuel assemblies. The design of the SSFC is a minor modification of the MCO. The modification is limited to the Shield Plug which remains unchanged with regard to interfaces with the canister shell. The performance characteristics remain those for the MCO, which bounds the payload of the SSFC.

  11. MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY

    SciTech Connect (OSTI)

    Hill, Thomas J

    2005-09-01T23:59:59.000Z

    The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to it’s mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to interim dry storage facilities, thus permitting the shutdown and decommission of the older facilities. Two wet pool facilities, one at the Idaho Nuclear Technology and Engineering Center and the other at Test Area North, were targeted for initial SNF movements since these were some of the oldest at the INL. Because of the difference in the SNF materials different types of drying processes had to be developed. Passive drying, as is done with typical commercial SNF was not an option because on the condition of some of the fuel, the materials to be dried, and the low heat generation of some of the SNF. There were also size limitations in the existing facility. Active dry stations were designed to address the specific needs of the SNF and the facilities.

  12. Nuclear Forensics Attributing the Source of Spent Fuel Used in an RDD Event

    SciTech Connect (OSTI)

    M.R. Scott

    2005-06-01T23:59:59.000Z

    An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burnup, age from discharge, reactor type, and initial fuel enrichment. It is shown that by analyzing the post-event material, these attributes can be determined with enough accuracy to be useful for investigators. The burnup can be found within a 5% accuracy, enrichment with a 2% accuracy, and age with a 10% accuracy. Reactor type can be determined if specific nuclides are measured. The methodology developed was implemented into a code call NEMASYS. NEMASYS is easy to use and it takes a minimum amount of time to learn its basic functions. It will process data within a few minutes and provide detailed information about the results and conclusions.

  13. Nonproliferation impacts assessment for the management of the Savannah River Site aluminum-based spent nuclear fuel

    SciTech Connect (OSTI)

    NONE

    1998-12-01T23:59:59.000Z

    On May 13, 1996, the US established a new, 10-year policy to accept and manage foreign research reactor spent nuclear fuel containing uranium enriched in the US. The goal of this policy is to reduce civilian commerce in weapons-usable highly enriched uranium (HEU), thereby reducing the risk of nuclear weapons proliferation. Two key disposition options under consideration for managing this fuel include conventional reprocessing and new treatment and packaging technologies. The Record of Decision specified that, while evaluating the reprocessing option, ``DOE will commission or conduct an independent study of the nonproliferation and other (e.g., cost and timing) implications of chemical separation of spent nuclear fuel from foreign research reactors.`` DOE`s Office of Arms Control and Nonproliferation conducted this study consistent with the aforementioned Record of Decision. This report addresses the nonproliferation implications of the technologies under consideration for managing aluminum-based spent nuclear fuel at the Savannah River Site. Because the same technology options are being considered for the foreign research reactor and the other aluminum-based spent nuclear fuels discussed in Section ES.1, this report addresses the nonproliferation implications of managing all the Savannah River Site aluminum-based spent nuclear fuel, not just the foreign research reactor spent nuclear fuel. The combination of the environmental impact information contained in the draft EIS, public comment in response to the draft EIS, and the nonproliferation information contained in this report will enable the Department to make a sound decision regarding how to manage all aluminum-based spent nuclear fuel at the Savannah River Site.

  14. Mission Need Statement: Idaho Spent Fuel Facility Project

    SciTech Connect (OSTI)

    Barbara Beller

    2007-09-01T23:59:59.000Z

    Approval is requested based on the information in this Mission Need Statement for The Department of Energy, Idaho Operations Office (DOE-ID) to develop a project in support of the mission established by the Office of Environmental Management to "complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research". DOE-ID requests approval to develop the Idaho Spent Fuel Facility Project that is required to implement the Department of Energy's decision for final disposition of spent nuclear fuel in the Geologic Repository at Yucca Mountain. The capability that is required to prepare Spent Nuclear Fuel for transportation and disposal outside the State of Idaho includes characterization, conditioning, packaging, onsite interim storage, and shipping cask loading to complete shipments by January 1,2035. These capabilities do not currently exist in Idaho.

  15. Information handbook on independent spent fuel storage installations

    SciTech Connect (OSTI)

    Raddatz, M.G.; Waters, M.D.

    1996-12-01T23:59:59.000Z

    In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996.

  16. Elements present in leach solutions from unsaturated spent fuel tests

    SciTech Connect (OSTI)

    Finn, P.A.; Bates, J.K.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Buck, E.C.; Gong, M.

    1993-10-01T23:59:59.000Z

    Preliminary results for the composition of the leachate from unsaturated tests at 90{degrees}C with spent fuel for 55--134 days with J-13 groundwater are reported. The pH of the leachate solutions was found to be acidic, ranging from 4 to 7. The actinide concentrations were 10{sup 5} greater than those reported for saturated spent fuel tests in which the leachate pH was 8. We also found that most species in the leachate were present as colloids containing both americium and curium. The presence of actinides in a form not currently included in repository radionuclide transport models provides information that can be used in spent fuel reaction modeling, the performance assessment of the repository and the design of the engineering barrier system. This report was prepared as part of the Yucca Mountain Site Characterization Project

  17. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement. Volume 1, Appendix D: Part A, Naval Spent Nuclear Fuel Management

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

    Volume 1 to the Department of Energy`s Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site.

  18. NUHOMS modular spent-fuel storage system: Performance testing

    SciTech Connect (OSTI)

    Strope, L.A.; McKinnon, M.A. (Pacific Northwest Lab., Richland, WA (USA)); Dyksterhouse, D.J.; McLean, J.C. (Carolina Power and Light Co., Raleigh, NC (USA))

    1990-09-01T23:59:59.000Z

    This report documents the results of a heat transfer and shielding performance evaluation of the NUTECH HOrizontal MOdular Storage (NUHOMS{reg sign}) System utilized by the Carolina Power and Light Co. (CP L) in an Independent Spent Fuel Storage Installation (ISFSI) licensed by the US Nuclear Regulatory Commission (NRC). The ISFSI is located at CP L's H. B. Robinson Nuclear Plant (HBR) near Hartsville, South Carolina. The demonstration included testing of three modules, first with electric heaters and then with spent fuel. The results indicated that the system was conservatively designed, with all heat transfer and shielding design criteria easily met. 5 refs., 45 figs., 9 tabs.

  19. CASTOR cask with high loading capacity for transport and storage of VVER 440 spent fuel

    SciTech Connect (OSTI)

    Diersch, R.; Methling, D.; Milde, G. [Gesellschaft fuer Nuklear-Behaelter mbH Essen (Germany)

    1993-12-31T23:59:59.000Z

    GNB has developed a CASTOR transport and storage cask with a capacity of 84 spent fuel assemblies from reactors of the type VVER 440. The safety analyses are performed with the help of modern, benchmarked calculation programs. The results show that the cask design is able to fulfill both the Type B test conditions on basis of IAEA Regulations-1985 edition and the requirements for interim storage sites in Germany.

  20. Spent Fuel Test-Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Final report

    SciTech Connect (OSTI)

    Patrick, W.C. (comp.)

    1986-03-30T23:59:59.000Z

    In the Climax stock granite on the Nevada Test Site, eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized. When test data indicated that the test objectives were met during the 3-year storage phase, the spent-fuel canisters were retrieved and the thermal sources were de-energized. The project demonstrated the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner. In addition to emplacement and retrieval operations, three exchanges of spent-fuel assemblies between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. The test led to development of a technical measurements program. To meet these objectives, nearly 1000 instruments and a computer-based data acquisition system were deployed. Geotechnical, seismological, and test status data were recorded on a continuing basis for the three-year storage phase and six-month monitored cool-down of the test. This report summarizes the engineering and scientific endeavors which led to successful design and execution of the test. The design, fabrication, and construction of all facilities and handling systems are discussed, in the context of test objectives and a safety assessment. The discussion progresses from site characterization and experiment design through data acquisition and analysis of test data in the context of design calculations. 117 refs., 52 figs., 81 tabs.

  1. Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary.

    SciTech Connect (OSTI)

    Gregson, Michael Warren; Brockmann, John E.; Nolte, O. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Loiseau, O. (Institut de radioprotection et de Surete Nucleaire, France); Koch, W. (Fraunhofer institut fur toxikologie und experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno (Institut de radioprotection et de Surete Nucleaire, France); Pretzsch, Gunter Guido (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Billone, M. C. (Argonne National Laboratory, USA); Lucero, Daniel A.; Burtseva, T. (Argonne National Laboratory, USA); Brucher, W (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2006-10-01T23:59:59.000Z

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO{sub 2}, test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage--aerosol test program, performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission, had significant inputs from, and is strongly supported and coordinated by both the U.S. and international program participants in Germany, France, and the U.K., as part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC.

  2. Foreign travel report: Visits to UK, Belgium, Germany, and France to benchmark European spent fuel and waste management technology

    SciTech Connect (OSTI)

    Ermold, L.F.; Knecht, D.A.

    1993-08-01T23:59:59.000Z

    The ICPP WINCO Spent Fuel and Waste Management Development Program recently was funded by DOE-EM to develop new technologies for immobilizing ICPP spent fuels, sodium-bearing liquid waste, and calcine to a form suitable for disposal. European organizations are heavily involved, in some cases on an industrial scale in areas of waste management, including spent fuel disposal and HLW vitrification. The purpose of this trip was to acquire first-hand European efforts in handling of spent reactor fuel and nuclear waste management, including their processing and technical capabilities as well as their future planning. Even though some differences exist in European and U.S. DOE waste compositions and regulations, many aspects of the European technologies may be applicable to the U.S. efforts, and several areas offer potential for technical collaboration.

  3. Spent Fuel Analyses for Nuclear Safeguards | ornl.gov

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of ScienceandMesa del Sol HomeFacebookScholarship Fund3Biology|SolarSpeakersSpectroscopy|Spent Fuel

  4. 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

    2003-07-01T23:59:59.000Z

    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.

  5. Spent nuclear fuel project design basis capacity study

    SciTech Connect (OSTI)

    Cleveland, K.J.

    1996-09-09T23:59:59.000Z

    A parametric study of the Spent Nuclear Fuel Project system capacity is presented. The study was completed using a commercially available software package to develop a summary level model of the major project systems. Alternative configurations, sub-system cycle times, and operating scenarios were tested to identify their impact on total project duration and equipment requirements.

  6. Review of Drying Methods for Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Large, W.S.

    1999-10-21T23:59:59.000Z

    SRTC is developing technology for direct disposal of aluminum spent nuclear fuel (SNF). The development program includes analyses and tests to support design and safe operation of a facility for ''road ready'' dry storage of SNF-filled canisters. The current technology development plan includes review of available SNF drying methods and recommendation of a drying method for aluminum SNF.

  7. Proliferation resistance of small modular reactors fuels

    SciTech Connect (OSTI)

    Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

    2013-07-01T23:59:59.000Z

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

  8. FIELD-DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE

    SciTech Connect (OSTI)

    Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

    2012-09-12T23:59:59.000Z

    Methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of aqueous spent fuel storage basins and determine the oxide thickness on the spent fuel basin materials were developed to assess the corrosion potential of a basin. this assessment can then be used to determine the amount of time fuel has spent in a storage basin to ascertain if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations and assist in evaluating general storage basin operations. The test kit was developed based on the identification of key physical, chemical and microbiological parameters identified using a review of the scientific and basin operations literature. The parameters were used to design bench scale test cells for additional corrosion analyses, and then tools were purchased to analyze the key parameters. The tools were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The sampling kit consisted of a total organic carbon analyzer, an YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization confirmed that the L Area basin is a well operated facility with low corrosion potential.

  9. Radiation measurements of uranium ingots from the electrometallurgical treatment of spent fuel.

    SciTech Connect (OSTI)

    Westphal, B. R.; Liaw, J. R.; Krsul, J. R.; Maddison, D. W.; Jensen, B. A.

    2003-03-24T23:59:59.000Z

    Radiation measurements and gamma spectroscopy analyses were made on numerous uranium ingots produced during the treatment of Experimental Breeder Reactor-II (EBR-II) spent nuclear fuel. The objective of these measurements was to provide background data for shielding concerns and potential process optimization. The uranium ingots resulted from the processing of both driver and blanket fuel by the electrometallurgical treatment process. The observed variation in the measurements was traced to the levels of certain fission product residues that remained in the uranium ingots produced during spent fuel treatment. A minor process change to hold the material at an elevated temperature for a specified length of time was found to significantly reduce concentrations of high-activity fission products and, thus the radiation field.

  10. The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life

    SciTech Connect (OSTI)

    J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

    1996-10-01T23:59:59.000Z

    A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

  11. Spent nuclear fuel project design basis capacity study

    SciTech Connect (OSTI)

    Cleveland, K.J.

    1998-07-22T23:59:59.000Z

    A parametric study of the Spent Nuclear Fuel Project system capacity is presented. The study was completed using a commercially available software package to develop a summary level model of the major project systems. A base case, reflecting the Fiscal Year 1998 process configuration, is evaluated. Parametric evaluations are also considered, investigating the impact of higher fuel retrieval system productivity and reduced shift operations at the canister storage building on total project duration.

  12. NAC-1 cask dose rate calculations for LWR spent fuel

    SciTech Connect (OSTI)

    CARLSON, A.B.

    1999-02-24T23:59:59.000Z

    A Nuclear Assurance Corporation nuclear fuel transport cask, NAC-1, is being considered as a transport and storage option for spent nuclear fuel located in the B-Cell of the 324 Building. The loaded casks will be shipped to the 200 East Area Interim Storage Area for dry interim storage. Several calculations were performed to assess the photon and neutron dose rates. This report describes the analytical methods, models, and results of this investigation.

  13. Thermal analysis for fuel handling system for sodium cooled reactor considering minor actinide-bearing metal fuel.

    SciTech Connect (OSTI)

    Chikazawa, Y.; Grandy, C.; Nuclear Engineering Division

    2009-03-01T23:59:59.000Z

    The Advanced Burner Reactor (ABR) is one of the components of the Global Nuclear Energy Partnership (GNEP) used to close the fuel cycle. ABR is a sodium-cooled fast reactor that is used to consume transuranic elements resulting from the reprocessing of light water reactor spent nuclear fuel. ABR-1000 [1000 MW(thermal)] is a fast reactor concept created at Argonne National Laboratory to be used as a reference concept for various future trade-offs. ABR-1000 meets the GNEP goals although it uses what is considered base sodium fast reactor technology for its systems and components. One of the considerations of any fast reactor plant concept is the ability to perform fuel-handling operations with new and spent fast reactor fuel. The transmutation fuel proposed as the ABR fuel has a very little experience base, and thus, this paper investigates a fuel-handling concept and potential issues of handling fast reactor fuel containing minor actinides. In this study, two thermal analyses supporting a conceptual design study on the ABR-1000 fuel-handling system were carried out. One analysis investigated passive dry spent fuel storage, and the other analysis investigated a fresh fuel shipping cask. Passive dry storage can be made suitable for the ABR-1000 spent fuel storage with sodium-bonded metal fuel. The thermal analysis shows that spent fast reactor fuel with a decay heat of 2 kW or less can be stored passively in a helium atmosphere. The 2-kW value seems to be a reasonable and practical level, and a combination of reasonably-sized in-sodium storage followed by passive dry storage could be a candidate for spent fuel storage for the next-generation sodium-cooled reactor with sodium-bonded metal fuel. Requirements for the shipping casks for minor actinide-bearing fuel with a high decay heat level are also discussed in this paper. The shipping cask for fresh sodium-cooled-reactor fuel should be a dry type to reduce the reaction between residual moisture on fresh fuel and the sodium coolant. The cladding temperature requirement is maintained below the creep temperature limit to avoid any damage before core installation. The thermal analysis shows that a helium gas-filled cask can accommodate ABR-1000 fresh minor actinide-bearing fuel with 700-W decay heat. The above analysis results revealed the overall requirement for minor actinide-bearing metal fuel handling. The information is thought to be helpful in the design of the ABR-1000 and future sodium-cooled-reactor fuel-handling system.

  14. Behavior of Spent Nuclear Fuel in Water Pool Storage

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office511041clothAdvanced Materials Advanced Materials Find More Like3.3 PrintVultureBehavior of Spent Nuclear

  15. Spent Fuel Transportation Risk Assessment | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium Transfer toSensorSoftware HelpsSouthAgendaSpectrum Policy SeminarSpent

  16. Storage of LWR spent fuel in air. Volume 3, Results from exposure of spent fuel to fluorine-contaminated air

    SciTech Connect (OSTI)

    Cunningham, M.E.; Thomas, L.E.

    1995-06-01T23:59:59.000Z

    The Behavior of Spent Fuel in Storage (BSFS) Project has conducted research to develop data on spent nuclear fuel (irradiated U0{sub 2}) that could be used to support design, licensing, and operation of dry storage installations. Test Series B conducted by the BSFS Project was designed as a long-term study of the oxidation of spent fuel exposed to air. It was discovered after the exposures were completed in September 1990 that the test specimens had been exposed to an atmosphere of bottled air contaminated with an unknown quantity of fluorine. This exposure resulted in the test specimens reacting with both the oxygen and the fluorine in the oven atmospheres. The apparent source of the fluorine was gamma radiation-induced chemical decomposition of the fluoro-elastomer gaskets used to seal the oven doors. This chemical decomposition apparently released hydrofluoric acid (HF) vapor into the oven atmospheres. Because the Test Series B specimens were exposed to a fluorine-contaminated oven atmosphere and reacted with the fluorine, it is recommended that the Test Series B data not be used to develop time-temperature limits for exposure of spent nuclear fuel to air. This report has been prepared to document Test Series B and present the collected data and observations.

  17. Initial measurements of BN-350 spent fuel in dry storage casks using the dual slab verification detonator

    SciTech Connect (OSTI)

    Santi, Peter Angelo [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Freeman, Corey R [Los Alamos National Laboratory; Parker, Robert F [Los Alamos National Laboratory; Williams, Richard B [Los Alamos National Laboratory

    2010-01-01T23:59:59.000Z

    The Dual Slab Verification Detector (DSVD) has been developed, built, and characterized by Los Alamos National Laboratory in cooperation with the International Atomic Energy Agency (IAEA) as part of the dry storage safeguards system for the spent fuel from the BN-350 fast reactor. The detector consists of two rows of 3He tubes embedded in a slab of polyethylene which has been designed to be placed on the outer surface of the dry storage cask. By performing DSVD measurements at several different locations around the outer surface of the DUC, a signature 'fingerprint' can be established for each DUC based on the neutron flux emanating from inside the dry storage cask. The neutron fingerprint for each individual DUC will be dependent upon the spatial distribution of nuclear material within the cask, thus making it sensitive to the removal of a certain amount of material from the cask. An initial set of DSVD measurements have been performed on the first set of dry storage casks that have been loaded with canisters of spent fuel and moved onto the dry storage pad to both establish an initial fingerprint for these casks as well as to quantify systematic uncertainties associated with these measurements. The results from these measurements will be presented and compared with the expected results that were determined based on MCNPX simulations of the dry storage facility. The ability to safeguard spent nuclear fuel is strongly dependent on the technical capabilities of establishing and maintaining continuity of knowledge (COK) of the spent fuel as it is released from the reactor core and either reprocessed or packaged and stored at a storage facility. While the maintenance of COK is often done using continuous containment and surveillance (C/S) on the spent fuel, it is important that the measurement capabilities exist to re-establish the COK in the event of a significant gap in the continuous CIS by performing measurements that independently confirm the presence and content of Plutonium (Pu) in the spent fuel. The types of non-destructive assay (NDA) measurements that can be performed on the spent fuel are strongly dependent on the type of spent fuel that is being safeguarded as well as the location in which the spent fuel is being stored. The BN-350 Spent Fuel Disposition Project was initiated to improve the safeguards and security of the spent nuclear fuel from the BN-350 fast-breeder reactor and was developed cooperatively to meet the requirements of the International Atomic Energy Agency (IAEA) as well as the terms of the 1993 CTR and MPC&A Implementing Agreements. The unique characteristics of fuel from the BN-350 fast-breeder reactor have allowed for the development of an integrated safeguards measurement program to inventory, monitor, and if necessary, re-verify Pu content of the spent fuel throughout the lifetime of the project. This approach includes the development of a safeguards measurement program to establish and maintain the COK on the spent fuel during the repackaging and eventual relocation of the spent-fuel assemblies to a long-term storage site. As part of the safeguards measurement program, the Pu content of every spent-fuel assembly from the BN-350 reactor was directly measured and characterized while the spent-fuel assemblies were being stored in the spent-fuel pond at the BN-350 facility using the Spent Fuel Coincidence Counter (SFCC). Upon completion of the initial inventory of the Pu content of the individual spent-fuel assemblies, the assemblies were repackaged into welded steel canisters that were filled with inert argon gas and held either four or six individual spent-fuel assemblies depending on the type of assembly that was being packaged. This repackaging of the spent-fuel assemblies was performed in order to improve the stability of the spent-fuel assemblies for long-term storage and increase the proliferation resistance of the spent fuel. To maintain the capability of verifying the presence of the spent-fuel assemblies inside the welded steel canisters, measurements were performed on the canis

  18. Technical strategy for the management of INEEL spent nuclear fuel

    SciTech Connect (OSTI)

    NONE

    1997-03-01T23:59:59.000Z

    This report presents evaluations, findings, and recommendations of the Idaho National Engineering and Environmental Laboratory (INEEL) Spent Nuclear Fuel Task Team. The technical strategy developed by the Task Team includes stabilization, near term storage, packaging, transport, and ultimate disposal. Key issues identified and discussed include waste characterization, criticality, packaging, waste form performance, and special fuels. Current plans focus on onsite needs, and include three central elements: (1) resolution of near-term vulnerabilities, (2) consolidation of storage locations, and (3) achieving dry storage in transportable packages. In addition to the Task Team report, appendices contain information on the INEEL spent fuel inventory; regulatory decisions and agreements; and analyses of criticality, packaging, storage, transportation, and system performance of a geological repository. 16 refs., 6 figs., 4 tabs.

  19. Spent fuel dry storage technology development: fuel temperature measurements under imposed dry storage conditions (I kW PWR spent fuel assembly)

    SciTech Connect (OSTI)

    Unterzuber, R.; Wright, J.B.

    1980-09-01T23:59:59.000Z

    A spent fuel assembly temperature test under imposed dry storage conditions was conducted at the Engine Maintenance Assembly and Disassembly (E-MAD) facility on the Nevada Test Site in support of spent fuel dry storage technology development. This document presents the test data and results obtained from an approximately 1.0 kW decay heat level PWR spent fuel assembly. A spent fuel test apparatus was designed to utilize a representative stainless steel spent fuel canister, a canister lid containing internal temperature instrumentation to measure fuel cladding temperatures, and a carbon steel liner that encloses the canister and lid. Electrical heaters along the liner length, on the lid, and below the canister are used to impose dry storage canister temperature profiles. Temperature instrumentation is provided on the liner and canister. The liner and canister are supported by a test stand in one of the large hot cells (West Process Cell) inside E-MAD. Fuel temperature measurements have been performed using imposed canister temperature profiles from the electrically heated and spent fuel drywell tests being conducted at E-MAD as well as for four constant canister temperature profiles, each with a vacuum, helium and air backfill. Computer models have been utilized in conjunction with the test to predict the thermal response of the fuel cladding. Computer predictions are presented, and they show good agreement with the test data.

  20. Alternate-fuel reactor studies

    SciTech Connect (OSTI)

    Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

    1983-02-01T23:59:59.000Z

    A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

  1. Feasibility of x ray fluorescence for spent fuel safeguards

    SciTech Connect (OSTI)

    Freeman, Corey Ross [Los Alamos National Laboratory; Mozin, Vladimir [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Fensin, Michael L [Los Alamos National Laboratory; White, Julia M [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Stafford, Alissa [TAMU; Charlton, William [TAMU

    2010-01-01T23:59:59.000Z

    Quantifying the Pu content in spent nuclear fuel is necessary for many reasons, in particular to verify that diversion or other illicit activities have not occurred. Therefore, safeguarding the world's nuclear fuel is paramount to responsible nuclear regulation and public acceptance, but achieving this goal presents many difficulties from both a technical and economic perspective. The Next Generation Safeguards Initiative (NGSI) of NA-24 is funding a large collaborative effort between multiple laboratories and universities to improve spent nuclear fuel safeguards methods and equipment. This effort involves the current work of modeling several different nondestructive assay (NDA) techniques. Several are being researched, because no single NDA technique, in isolation, has the potential to properly characterize fuel assemblies and offer a robust safeguards measure. The insights gained from this research, will be used to down-select from the original set a few of the most promising techniques that complement each other. The goal is to integrate the selected instruments to create an accurate measurement system for fuel verification that is also robust enough to detect diversions. These instruments will be fabricated and tested under realistic conditions. This work examines one of the NDA techniques; the feasibility of using x ray emission peaks from Pu and U to gather information about their relative quantities in the spent fuel. X Ray Fluorescence (XRF), is unique compared to the investigated techniques in that it is the only one able to give the elemental ratio of Pu to U, allowing the possibility of a Pu gram quantity for the assembly to be calculated. XRF also presents many challenges, mainly its low penetration, since the low energy x rays of interest are effectively shielded by the first few millimeters of a fuel pin. This paper will explore the results of Monte Carlo N-Particle eXtended (MCNPX) transport code calculations of spent fuel x ray peaks. The MCNPX simulations will be benchmarked against measurements taken at Oak Ridge. Analysis of the feasibility of XRFs role in spent nuclear fuel safeguards efforts, particularly in the context of the overall NGSI effort will be discussed.

  2. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    SciTech Connect (OSTI)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2008-03-01T23:59:59.000Z

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and coordinated by both the U.S. and international program participants in Germany, France, and others, as part of the International Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC).

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

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

    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...

  4. FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

    SciTech Connect (OSTI)

    Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.; Livingston, James V.

    2014-03-31T23:59:59.000Z

    This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset will be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.

  5. An Empirical Approach to Bounding the Axial Reactivity Effects of PWR Spent Nuclear Fuel

    SciTech Connect (OSTI)

    P. M. O'Leary; J. M. Scaglione

    2001-04-04T23:59:59.000Z

    One of the significant issues yet to be resolved for using burnup credit (BUC) for spent nuclear fuel (SNF) is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters (such as local power, fuel temperature, moderator temperature, burnable poison rod history, and soluble boron concentration) affect the isotopic inventory of fuel that is depleted in a pressurized water reactor (PWR). However, obtaining the detailed operating histories needed to model all PWR fuel assemblies to which BUC would be applied is an onerous and costly task. Simplifications therefore have been suggested that could lead to using ''bounding'' depletion parameters that could be broadly applied to different fuel assemblies. This paper presents a method for determining a set of bounding depletion parameters for use in criticality analyses for SNF.

  6. Examining 239Pu and 240Pu Nuclear Resonance Fluorescence Measurements on Spent Fuel for Nuclear Safeguards

    E-Print Network [OSTI]

    Quiter, Brian

    2013-01-01T23:59:59.000Z

    Spent Fuel Assay Using Nuclear Resonance Fluo- rescence,” Annual Meeting of the Institute of Nuclear Material Management,

  7. Extending dry storage of spent LWR fuel for 100 years.

    SciTech Connect (OSTI)

    Einziger, R. E.

    1998-12-16T23:59:59.000Z

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and excessive creep and mechanical property changes. Postulated accident scenarios would be the same for 20-year or 100-year storage, because they are mostly governed by operational or outside events, and not by the cask or fuel. Analyses of accident scenarios during extended dry storage could be impacted by fuel and cask changes that would result from the extended period of storage. Overall, the results of this work indicate that, based on fuel behavior, spent fuel at burnups below {approximately}45 GWd/MTU can be dry stored for 100 years. Long-term storage of higher burnup fuel or fuels with newer cladding will require the determination of temperature limits based on evaluation of stress-driven degradation mechanisms of the cladding.

  8. Impact of PWR spent fuel variations on TRU-fueled VHTRS

    E-Print Network [OSTI]

    Alajo, Ayodeji Babatunde

    2009-05-15T23:59:59.000Z

    Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the first 500 years. One of the most viable approaches suggests...

  9. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    E-Print Network [OSTI]

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01T23:59:59.000Z

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

  10. 3DD - Three Dimensional Disposal of Spent Nuclear Fuel - 12449

    SciTech Connect (OSTI)

    Dvorakova, Marketa; Slovak, Jiri [Radioactive Waste Repository Authority (RAWRA), Prague (Czech Republic)

    2012-07-01T23:59:59.000Z

    Three dimensional disposal is being considered as a way in which to store long-term spent nuclear fuel in underground disposal facilities in the Czech Republic. This method involves a combination of the two most common internationally recognised disposal methods in order to practically apply the advantages of both whilst, at the same time, eliminating their weaknesses; the method also allows easy removal in case of potential re-use. The proposed method for the disposal of spent nuclear fuel will reduce the areal requirements of future deep geological repositories by more than 30%. It will also simplify the container handling process by using gravitational forces in order to meet requirements concerning the controllability of processes and ensuring operational and nuclear safety. With regard to the issue of the efficient potential removal of waste containers, this project offers an ingenious solution which does not disrupt the overall stability of the original disposal complex. (authors)

  11. Molten tin reprocessing of spent nuclear fuel elements

    DOE Patents [OSTI]

    Heckman, Richard A. (Castro Valley, CA)

    1983-01-01T23:59:59.000Z

    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.

  12. Operation of N Reactor and Fuels Fabrication Facilities, Hanford Reservation, Richland, Benton County, Washington: Environmental assessment

    SciTech Connect (OSTI)

    Not Available

    1980-08-01T23:59:59.000Z

    Environmental data, calculations and analyses show no significant adverse radiological or nonradiological impacts from current or projected future operations resulting from N Reactor, Fuels Fabrication and Spent Fuel Storage Facilities. Nonoccupational radiation exposures resulting from 1978 N Reactor operations are summarized and compared to allowable exposure limits.

  13. Method For Processing Spent (Trn,Zr)N Fuel

    DOE Patents [OSTI]

    Miller, William E. (Naperville, IL); Richmann, Michael K. (Woodridge, IL)

    2004-07-27T23:59:59.000Z

    A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

  14. Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements

    SciTech Connect (OSTI)

    KLEM, M.J.

    2000-10-18T23:59:59.000Z

    In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the references used for this document.

  15. Machine Vision Tests for Spent Fuel Scrap Characteristics

    SciTech Connect (OSTI)

    BERGER, W.W.

    2000-04-27T23:59:59.000Z

    The purpose of this work is to perform a feasibility test of a Machine Vision system for potential use at the Hanford K basins during spent nuclear fuel (SNF) operations. This report documents the testing performed to establish functionality of the system including quantitative assessment of results. Fauske and Associates, Inc., which has been intimately involved in development of the SNF safety basis, has teamed with Agris-Schoen Vision Systems, experts in robotics, tele-robotics, and Machine Vision, for this work.

  16. Study on concrete cask storage of spent fuel in Japan

    SciTech Connect (OSTI)

    Itoh, C. [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Nuclear Fuel Cycle Dept.; Onodera, A.; Yamada, N. [Hitachi Zosen Corp., Tokyo (Japan). Nuclear Div.

    1993-12-31T23:59:59.000Z

    The present report describes the status of the first year`s work of a five-year-long study on concrete cask storage of spent fuel in Japan. Firstly, the proposed study program is elaborated to clarify the position of the present work. Then, the results of the study which have been obtained so far are described and the technical issues are addressed to make the concrete cask storage viable in Japan.

  17. Thermoelectric powered wireless sensors for spent fuel monitoring

    SciTech Connect (OSTI)

    Carstens, T.; Corradini, M.; Blanchard, J. [Dept. of Engineering Physics, Univ. of Wisconsin-Madison, Madison, WI 53706 (United States); Ma, Z. [Dept. of Electrical and Computer Engineering, Univ. of Wisconsin-Madison, Madison, WI 53706 (United States)

    2011-07-01T23:59:59.000Z

    This paper describes using thermoelectric generators to power wireless sensors to monitor spent nuclear fuel during dry-cask storage. OrigenArp was used to determine the decay heat of the spent fuel at different times during the service life of the dry-cask. The Engineering Equation Solver computer program modeled the temperatures inside the spent fuel storage facility during its service life. The temperature distribution in a thermoelectric generator and heat sink was calculated using the computer program Finite Element Heat Transfer. From these temperature distributions the power produced by the thermoelectric generator was determined as a function of the service life of the dry-cask. In addition, an estimation of the path loss experienced by the wireless signal can be made based on materials and thickness of the structure. Once the path loss is known, the transmission power and thermoelectric generator power requirements can be determined. This analysis estimates that a thermoelectric generator can produce enough power for a sensor to function and transmit data from inside the dry-cask throughout its service life. (authors)

  18. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    NONE

    1995-09-01T23:59:59.000Z

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  19. Opportunities to increase the productivity of spent fuel shipping casks in the United States

    SciTech Connect (OSTI)

    Winsor, G.H.; Faletti, D.W.; DeSteese, J.G.

    1980-03-01T23:59:59.000Z

    Trends indicate that future transportation requirements for spent fuel will be different from those anticipated when the current generation of casks and vehicles was designed. Increased storage capacity at most reactors will increase the average post irradiation age of the spent fuel to be transported. A scenario is presented which shows the 18 casks currently available should be sufficient until approximately 1983. Beyond this time, it appears that an adequate transportation system can be maintained by acquiring, as needed, casks of current designs and new casks currently under development. Spent fuel transportation requirements in the post-1990 period can be met by a new generation of casks specifically designed to transport long-cooled fuel. In terms of the number of casks needed, productivity may be increased by 19% if rail cask turnaround time is reduced to 4 days from the current range of 6.5 to 8.5 days. Productivity defined as payloads per cask year could be increased 62% if the turnaround time for legal weight truck casks were reduced from 12 hours to 4 hours. On a similar basis, overweight truck casks show a 28% increase in productivity.

  20. Evaluation of Effect of Fuel Assembly Loading Patterns on Thermal and Shielding Performance of a Spent Fuel Storage/Transportation Cask

    SciTech Connect (OSTI)

    Cuta, Judith M.; Jenquin, Urban P.; McKinnon, Mikal A.

    2001-11-20T23:59:59.000Z

    The licensing of spent fuel storage casks is generally based on conservative analyses that assume a storage system being uniformly loaded with design basis fuel. The design basis fuel typically assumes a maximum assembly enrichment, maximum burn up, and minimum cooling time. These conditions set the maximum decay heat loads and radioactive source terms for the design. Recognizing that reactor spent fuel pools hold spent fuel with an array of initial enrichments, burners, and cooling times, this study was performed to evaluate the effect of load pattern on peak cladding temperature and cask surface dose rate. Based on the analysis, the authors concluded that load patterns could be used to reduce peak cladding temperatures in a cask without adversely impacting the surface dose rates.

  1. An approach to determine a defensible spent fuel ratio.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Lindgren, Eric Richard

    2014-03-01T23:59:59.000Z

    Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO2), have been conducted in the interim to more definitively determine the source term from these postulated events. In all the previous studies, the postulated attack of greatest interest was by a conical shape charge (CSC) that focuses the explosive energy much more efficiently than bulk explosives. However, the validity of these large-scale results remain in question due to the lack of a defensible Spent Fuel Ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical DUO2 surrogate. Previous attempts to define the SFR have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Different researchers have suggested using SFR values of 3 to 5.6. Sound technical arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and dry storage of spent nuclear fuel. Currently, Oak Ridge National Laboratory (ORNL) is in possession of several samples of spent nuclear fuel (SNF) that were used in the original SFR studies in the 1980's and were intended for use in a modern effort at Sandia National Laboratories (SNL) in the 2000's. A portion of these samples are being used for a variety of research efforts. However, the entirety of SNF samples at ORNL is scheduled for disposition at the Waste Isolation Pilot Plant (WIPP) by approximately the end of 2015. If a defensible SFR is to be determined for use in storage and transportation security analyses, the need to begin this effort is urgent in order to secure the only known available SNF samples with a clearly defined path to disposal.

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

    DOE Patents [OSTI]

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

    1982-01-19T23:59:59.000Z

    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.

  3. Method for reprocessing and separating spent nuclear fuels

    DOE Patents [OSTI]

    Krikorian, Oscar H. (Danville, CA); Grens, John Z. (Livermore, CA); Parrish, Sr., William H. (Walnut Creek, CA)

    1983-01-01T23:59:59.000Z

    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.

  4. Spent nuclear fuel storage -- Performance tests and demonstrations

    SciTech Connect (OSTI)

    McKinnon, M.A.; DeLoach, V.A.

    1993-04-01T23:59:59.000Z

    This report summarizes the results of heat transfer and shielding performance tests and demonstrations conducted from 1983 through 1992 by or in cooperation with the US Department of Energy (DOE), Office of Commercial Radioactive Waste Management (OCRWM). The performance tests consisted of 6 to 14 runs involving one or two loadings, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. A description of the test plan, spent fuel load patterns, results from temperature and dose rate measurements, and fuel integrity evaluations are contained within the report.

  5. Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; D. Vaden; S.X. Li; G.L. Fredrickson; R.D. Mariani

    2009-09-01T23:59:59.000Z

    During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontally displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.

  6. Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests

    SciTech Connect (OSTI)

    Wilson, C.N.

    1990-06-01T23:59:59.000Z

    The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85{degree}C and 25{degree}C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs.

  7. Reactor Physics Assessment of the Inclusion of Unseparated Neptunium in MOX Reactor Fuel

    SciTech Connect (OSTI)

    Ellis, Ronald James [ORNL

    2009-01-01T23:59:59.000Z

    Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.

  8. Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R. [and others

    1996-07-01T23:59:59.000Z

    The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

  9. Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel

    E-Print Network [OSTI]

    Ludewigt, Bernhard A

    2011-01-01T23:59:59.000Z

    Spent Fuel Library for Assessing Varied Nondestructive Assay Techniques for Nuclear Safeguards,” LA-UR 09-01188, ANS Advances in Nuclear Fuel Management

  10. Dose Rates for Various Loading Patterns of Spent Fuel Assemblies in a Dry Cask

    SciTech Connect (OSTI)

    Jenquin, Urban P. (BATTELLE (PACIFIC NW LAB))

    2001-01-01T23:59:59.000Z

    Shielding calculations were performed to assess the impact of loading various combinations of spent fuel on dose rates and fuel temperature in a dry storage cask.

  11. High Energy Delayed Gamma Spectroscopy for Plutonium Assay of Spent Fuel

    SciTech Connect (OSTI)

    Campbell, Luke W.; Misner, Alex C.; Smith, Leon E.; Reese, Steve; Robinson, Joshua

    2010-11-05T23:59:59.000Z

    The direct measurement of plutonium in spent reactor fuel is an unmet challenge in international safeguards. In this simulation study, we investigate the use of the delayed gamma rays from fission product nuclei to determine the amount of fissile isotopes (Pu-239, Pu-241, and U-235) in irradiated light water reactor fuel assemblies. Fission is stimulated with an interrogating neutron source, and the radiation from the short lived fission products is measured. This measured gamma spectrum is then fit to a linear combination of spectra from pure Pu-239, Pu-241, and U-235 to determine the proportion of fissile isotopes present. In this paper, we describe the modelling and analysis methods used to represent the background of radioemissions from long-lived isotopes originally present in the spent fuel and the short time scale delayed gamma signal. Results are presented for simulations using a nominal instrument design on a library of fuel assemblies with burnups ranging from 0 to 60 GWd/MTU.

  12. Nuclear reactor composite fuel assembly

    DOE Patents [OSTI]

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01T23:59:59.000Z

    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.

  13. APPLICATIONS OF CURRENT TECHNOLOGY FOR CONTINUOUS MONITORING OF SPENT FUEL

    SciTech Connect (OSTI)

    Drayer, R.

    2013-06-09T23:59:59.000Z

    Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each specific need. There are products on the market for smart meters, industrial lighting control and home automation that can be applied to the Back End Fuel Cycle. With a little integration and innovation a cost effective solution is achievable.

  14. Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories

    Office of Legacy Management (LM)

    Radiological Condition of the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories Cheswick, Pennsylvania -. -, -- AGENCY: Office of Operational Safety, Department...

  15. Corrosion of Spent Nuclear Fuel: The Long-Term Assessment

    SciTech Connect (OSTI)

    Rodney C. Ewing

    2004-10-07T23:59:59.000Z

    Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

  16. Closure mechanism and method for spent nuclear fuel canisters

    DOE Patents [OSTI]

    Doman, Marvin J. (Monroeville, PA)

    2004-11-23T23:59:59.000Z

    A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

  17. Interface agreement for the management of FFTF Spent Nuclear Fuel

    SciTech Connect (OSTI)

    McCormack, R.L.

    1995-02-02T23:59:59.000Z

    The Hanford Site Spent Nuclear Fuel (SNF) Project was formed to manage the SNF at Hanford. The mission of the Fast Flux Test Facility (FFTF) Transition Project is to place the facility in a radiologically and industrially safe shutdown condition for turnover to the Environmental Restoration Contractor (ERC) for subsequent D&D. To satisfy both project missions, FFTF SNF must be removed from the FFTF and subsequently dispositioned. This documented provides the interface agreement between FFTF Transition Project and SNF Project for management of the FFTF SNF.

  18. Smelting Associated with the Advanced Spent Fuel Conditioning Process

    SciTech Connect (OSTI)

    Hur, J-M.; Jeong, M-S.; Lee, W-K.; Cho, S-H.; Seo, C-S.; Park, S-W.

    2004-10-03T23:59:59.000Z

    The smelting process associated with the advanced spent fuel conditioning process (ACP) of Korea Atomic Energy Research Institute was studied by using surrogate materials. Considering the vaporization behaviors of input materials, the operation procedure of smelting was set up as (1) removal of residual salts, (2) melting of metal powder, and (3) removal of dross from a metal ingot. The behaviors of porous MgO crucible during smelting were tested and the chemical stability of MgO in the salt-being atmosphere was confirmed.

  19. Production of metal waste forms from spent fuel treatment

    SciTech Connect (OSTI)

    Westphal, B.R.; Keiser, D.D.; Rigg, R.H.; Laug, D.V.

    1995-02-01T23:59:59.000Z

    Treatment of spent nuclear fuel at Argonne National Laboratory consists of a pyroprocessing scheme in which the development of suitable waste forms is being advanced. Of the two waste forms being proposed, metal and mineral, the production of the metal waste form utilizes induction melting to stabilize the waste product. Alloying of metallic nuclear materials by induction melting has long been an Argonne strength and thus, the transition to metallic waste processing seems compatible. A test program is being initiated to coalesce the production of the metal waste forms with current induction melting capabilities.

  20. Technology status in support of refined technical baseline for the Spent Nuclear Fuel project. Revision 1

    SciTech Connect (OSTI)

    Puigh, R.J.; Toffer, H.; Heard, F.J.; Irvin, J.J.; Cooper, T.D.

    1995-10-20T23:59:59.000Z

    The Spent Nuclear Fuel Project (SNFP) has undertaken technology acquisition activities focused on supporting the technical basis for the removal of the N Reactor fuel from the K Basins to an interim storage facility. The purpose of these technology acquisition activities has been to identify technology issues impacting design or safety approval, to establish the strategy for obtaining the necessary information through either existing project activities, or the assignment of new work. A set of specific path options has been identified for each major action proposed for placing the N Reactor fuel into a ``stabilized`` form for interim storage as part of this refined technical basis. This report summarizes the status of technology information acquisition as it relates to key decisions impacting the selection of specific path options. The following specific categories were chosen to characterize and partition the technology information status: hydride issues and ignition, corrosion, hydrogen generation, drying and conditioning, thermal performance, criticality and materials accountability, canister/fuel particulate behavior, and MCO integrity. This report represents a preliminary assessment of the technology information supporting the SNFP. As our understanding of the N Reactor fuel performance develops the technology information supporting the SNFP will be updated and documented in later revisions to this report. Revision 1 represents the incorporation of peer review comments into the original document. The substantive evolution in our understanding of the technical status for the SNFP (except section 3) since July 1995 have not been incorporated into this revision.

  1. Spent Nuclear Fuel Dry Transfer System Cold Demonstration Project Final Report

    SciTech Connect (OSTI)

    Christensen, Max R; McKinnon, M. A.

    1999-12-01T23:59:59.000Z

    The spent nuclear fuel dry transfer system (DTS) provides an interface between large and small casks and between storage-only and transportation casks. It permits decommissioning of reactor pools after shutdown and allows the use of large storage-only casks for temporary onsite storage of spent nuclear fuel irrespective of reactor or fuel handling limitations at a reactor site. A cold demonstration of the DTS prototype was initiated in August 1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). The major components demonstrated included the fuel assembly handling subsystem, the shield plug/lid handling subsystem, the cask interface subsystem, the demonstration control subsystem, a support frame, and a closed circuit television and lighting system. The demonstration included a complete series of DTS operations from source cask receipt and opening through fuel transfer and closure of the receiving cask. The demonstration included both normal operations and recovery from off-normal events. It was designed to challenge the system to determine whether there were any activities that could be made to jeopardize the activities of another function or its safety. All known interlocks were challenged. The equipment ran smoothly and functioned as designed. A few "bugs" were corrected. Prior to completion of the demonstration testing, a number of DTS prototype systems were modified to apply lessons learned to date. Additional testing was performed to validate the modifications. In general, all the equipment worked exceptionally well. The demonstration also helped confirm cost estimates that had been made at several points in the development of the system.

  2. Application of Neutron-Absorbing Structural-Amorphous metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls

    E-Print Network [OSTI]

    2006-01-01T23:59:59.000Z

    241 Pu, etc. ). To prevent nuclear criticality in spent fuelto enhance criticality safety for spent nuclear fuel inSpent Nuclear Fuel (SNF) Container to Enhance Criticality

  3. Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study

    SciTech Connect (OSTI)

    CLEVELAND, K.J.

    2000-08-17T23:59:59.000Z

    This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage.

  4. Sensitivity Analysis of Reprocessing Cooling Times on Light Water Reactor and Sodium Fast Reactor Fuel Cycles

    SciTech Connect (OSTI)

    R. M. Ferrer; S. Bays; M. Pope

    2008-04-01T23:59:59.000Z

    The purpose of this study is to quantify the effects of variations of the Light Water Reactor (LWR) Spent Nuclear Fuel (SNF) and fast reactor reprocessing cooling time on a Sodium Fast Reactor (SFR) assuming a single-tier fuel cycle scenario. The results from this study show the effects of different cooling times on the SFR’s transuranic (TRU) conversion ratio (CR) and transuranic fuel enrichment. Also, the decay heat, gamma heat and neutron emission of the SFR’s fresh fuel charge were evaluated. A 1000 MWth commercial-scale SFR design was selected as the baseline in this study. Both metal and oxide CR=0.50 SFR designs are investigated.

  5. Test plan for long-term, low-temperature oxidation of BWR spent fuel

    SciTech Connect (OSTI)

    Einziger, R.E.

    1988-12-01T23:59:59.000Z

    Preliminary studies indicated the need for more spent fuel oxidation data in order to determine the probable behavior of spent fuel in a tuff repository. Long-term, low-temperature testing was recommended in a comprehensive technical approach to (1) confirm the findings of the short-term thermogravimetric analysis tests; (2) evaluate the effects of variables such as burnup, atmospheric moisture,and fuel type on the oxidation rate; and (3) extend the oxidation data base to representative repository temperatures and better define the temperature dependence of the operative oxidation mechanisms. This document presents the test plan to study the effects of atmospheric moisture and temperature on oxidation rate and phase formation using a large number of boiling-water reactor fuel samples. Tests will run for up to two years, use characterized fragmented and pulverized fuel samples, cover a temperature range of 110{degree}C to 175{degree}C, and be conducted with an atmospheric moisture content ranging from <{minus}55{degree}C to {approximately}80{degree}C dew point. After testing, the samples will be examined and made available for leaching testing. 15 refs., 2 figs., 2 tabs.

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

    SciTech Connect (OSTI)

    Parker, Frank L. [Vanderbilt University (United States)

    2012-07-01T23:59:59.000Z

    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 term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)

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

    Broader source: Energy.gov [DOE]

    Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs

  8. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    SciTech Connect (OSTI)

    Canaan, R.E.

    1995-12-01T23:59:59.000Z

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

  9. Delayed Gamma-Ray Spectroscopy for Spent Nuclear Fuel Assay

    SciTech Connect (OSTI)

    Campbell, Luke W.; Hunt, Alan W.; Ludewigt, Bernhard A.; Mozin, Vladimir V.

    2012-04-01T23:59:59.000Z

    High-energy, beta-delayed gamma-ray spectroscopy is investigated as a non-destructive assay technique for the determination of plutonium mass in spent nuclear fuel. This approach exploits the unique isotope-specific signatures contained in the delayed gamma-ray emission spectra detected following active interrogation with an external neutron source. A high fidelity modeling approach is described that couples radiation transport, analytical decay/depletion, and a newly developed gamma-ray emission source reconstruction code. Initially simulated and analyzed was a “one-pass” delayed gamma-ray assay that focused on the long-lived signatures. Also presented are the results of an independent study that investigated “pulsed mode” measurements, to capture the more isotope-specific, short-lived signatures. Initial modeling results outlined in this paper suggest that delayed gamma-ray assay of spent nuclear fuel assemblies can be accomplished with a neutron generator of sufficient strength and currently available gamma-ray detectors.

  10. Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment

    SciTech Connect (OSTI)

    NONE

    1996-02-01T23:59:59.000Z

    On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operation of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts.

  11. DEVELOPMENT OF METHODOLOGY AND FIELD DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE

    SciTech Connect (OSTI)

    Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

    2012-06-04T23:59:59.000Z

    This project developed methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of the fuel storage medium and determine the oxide thickness on the spent fuel basin materials. The overall objective of this project was to determine the amount of time fuel has spent in a storage basin to determine if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations. This project developed and validated forensic tools that can be used to predict the age and condition of spent nuclear fuels stored in liquid basins based on key physical, chemical and microbiological basin characteristics. Key parameters were identified based on a literature review, the parameters were used to design test cells for corrosion analyses, tools were purchased to analyze the key parameters, and these were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The key parameters identified in the literature review included chloride concentration, conductivity, and total organic carbon level. Focus was also placed on aluminum based cladding because of their application to weapons production. The literature review was helpful in identifying important parameters, but relationships between these parameters and corrosion rates were not available. Bench scale test systems were designed, operated, harvested, and analyzed to determine corrosion relationships between water parameters and water conditions, chemistry and microbiological conditions. The data from the bench scale system indicated that corrosion rates were dependent on total organic carbon levels and chloride concentrations. The highest corrosion rates were observed in test cells amended with sediment, a large microbial inoculum and an organic carbon source. A complete characterization test kit was field tested to characterize the SRS L-Area spent fuel basin. The sampling kit consisted of a TOC analyzer, a YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization was done over a two day period in June 2011, and confirmed that the L Area basin is a well operated facility with low corrosion potential.

  12. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01T23:59:59.000Z

    Oxford ; New York ; Oxford University Press. Fuel- Trac,Spent Fuel / Reprocessing, in Nuclear Industry Statusto Burn Non-Fissile Fuels. 2008. GA. Energy Multiplier

  13. Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques NGSI Research Overview and Update on NDA Techniques

    E-Print Network [OSTI]

    A., V. Mozin, S.J. Tobin, L.W. Cambell, J.R. Cheatham, C.R. Freeman, C.J. Gesh,

    2012-01-01T23:59:59.000Z

    Spent Fuel Library for Assessing Varied Nondestructive Assay Techniques for Nuclear Safeguards,” Advances in Nuclear Fuel Management

  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. [RPA - V.G.Khlopin Radium Institute, St-Petersburg (Russian Federation); Baryshnikov, M.V.; Kryukov, O.V.; Khaperskaya, A.V. [State Corporation ROSATOM, Moscow (Russian Federation)

    2013-07-01T23:59:59.000Z

    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. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

  16. Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content 

    E-Print Network [OSTI]

    Stafford, Alissa Sarah

    2010-10-12T23:59:59.000Z

    The quantification of plutonium (Pu) in spent nuclear fuel is an increasingly important safeguards issue. There exists an estimated worldwide 980 metric tons of Pu in the nuclear fuel cycle and the majority is in spent nuclear fuel waiting for long...

  17. EM Prepares Report for Convention on Safety of Spent Fuel and...

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

    agency to produce a report recently for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The Convention was...

  18. EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    DOE prepared a EIS that evaluated the potential environmental impacts of treatment and management of DOE-owned sodium bonded spent nuclear fuel.

  19. Brittle Failure Design Criteria for Ductile Cast Iron Spent-Fuel

    Office of Scientific and Technical Information (OSTI)

    for Ductile and Brittle Failure Design Criteria for Ductile Cast Iron Spent-Fuel Shipping Containers This report was prepared as an account of work sponsored by an...

  20. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

    SciTech Connect (OSTI)

    Wagner, J.C.; Parks, C.V.

    2000-09-01T23:59:59.000Z

    This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub inf} estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k{sub inf} estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration ({approximately}2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion ({le} 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE. Consequently, the findings presented here do not represent a significant safety concern unless/until the subcritical margin associated with the soluble boron (that is not currently explicitly credited) is offset by the uncertainties associated with burnup credit and/or the expanded allowance of credit for the soluble boron.

  1. Radioactive Waste Management: Study of Spent Fuel Dissolution Rates in Geological Storage Using Dosimetry Modeling and Experimental Verification

    SciTech Connect (OSTI)

    Hansen, Brady; Miller, William

    2011-10-28T23:59:59.000Z

    This research will provide improved predictions into the mechanisms and effects of radiolysis on spent nuclear fuel dissolution in a geological respository through accurate dosimetry modeling of the dose to water, mechanistic chemistry modeling of the resulting radiolytic reactions and confirmatory experimental measurements. This work will combine effort by the Nuclear Science and Engineering Institute (NSEI) and the Missouri University Research Reactor (MURR) at the University of Missouri-Columbia, and the expertise and facilities at the Pacific Northwest National Laboratory (PNNL).

  2. HOW MANY DID YOU SAY? HISTORICAL AND PROJECTED SPENT NUCLEAR FUEL SHIPMENTS IN THE UNITED STATES, 1964 - 2048

    SciTech Connect (OSTI)

    Halstead, Robert J.; Dilger, Fred

    2003-02-27T23:59:59.000Z

    No comprehensive, up-to-date, official database exists for spent nuclear fuel shipments in the United States. The authors review the available data sources, and conclude that the absence of such a database can only be rectified by a major research effort, similar to that carried out by Oak Ridge National Laboratory (ORNL) in the early 1990s. Based on a variety of published references, and unpublished data from the U.S. Nuclear Regulatory Commission (NRC), the authors estimate cumulative U.S. shipments of commercial spent fuel for the period 1964-2001. The cumulative estimates include quantity shipped, number of cask-shipments, and shipment-miles, by truck and by rail. The authors review previous estimates of future spent fuel shipments, including contractor reports prepared for the U.S. Department of Energy (DOE), NRC, and the State of Nevada. The DOE Final Environmental Impact Statement (FEIS) for Yucca Mountain includes projections of spent nuclear fuel and high-level radioactive was te shipments for two inventory disposal scenarios (24 years and 38 years) and two national transportation modal scenarios (''mostly legal-weight truck'' and ''mostly rail''). Commercial spent fuel would compromise about 90 percent of the wastes shipped to the repository. The authors estimate potential shipments to Yucca Mountain over 38 years (2010-2048) for the DOE ''mostly legal-weight truck'' and ''mostly rail'' scenarios, and for an alternative modal mix scenario based on current shipping capabilities of the 72 commercial reactor sites. The cumulative estimates of future spent fuel shipments include quantity shipped, number of cask-shipments, and shipment-miles, by legal-weight truck, heavy-haul truck, rail and barge.

  3. Utilizing Divers in Support of Spent Fuel Basin Closure Subproject

    SciTech Connect (OSTI)

    Allen Nellesen

    2005-01-01T23:59:59.000Z

    A number of nuclear facilities in the world are aging and with this comes the fact that we have to either keep repairing them or decommission them. At the Department of Energy Idaho Site (DOEID) there are a number of facilities that are being decommissioned, but the facilities that pose the highest risk to the large aquifer that flows under the site are given highest priorities. Aging spent nuclear fuel pools at DOE-ID are among the facilities that pose the highest risk, therefore four pools were targeted for decommissioning in Fiscal Year 2004. To accomplish this task the Idaho Completion Project (ICP) of Bechtel BWXT Idaho, LLC, put together an integrated Basin Closure Subproject team. The team was assigned a goal to look beyond traditional practices at the Idaho National Engineering and Environmental Laboratory (INEEL) to find ways to get the basin closure work done safer and more efficiently. The Idaho Completion Project (ICP) was faced with a major challenge – cleaning and preparing aging spent nuclear fuel basins for closure by removing sludge and debris, as necessary, and removing water to eliminate a potential risk to the Snake River Plain Aquifer. The project included cleaning and removing water from four basins. Two of the main challenges to a project like this is the risk of contamination from the basin walls and floors becoming airborne as the water is removed and keeping personnel exposures ALARA. ICP’s baseline plan had workers standing at the edges of the basins and on rafts or bridge cranes and then using long-handled tools to manually scrub the walls of basin surfaces. This plan had significant risk of skin contamination events, workers falling into the water, or workers sustaining injuries from the awkward working position. Analysis of the safety and radiation dose risks presented by this approach drove the team to look for smarter ways to get the work done.

  4. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    SciTech Connect (OSTI)

    Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

    2011-06-30T23:59:59.000Z

    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 determining Pu mass in spent fuel assemblies.

  5. Fort Calhoun Station disposal of spent fuel pool racks

    SciTech Connect (OSTI)

    Jamieson, T.W. [Omaha Public Power District, Fort Calhoun Station, NE (United States)

    1995-09-01T23:59:59.000Z

    The original plan was to have the racks pulled out of the pool, washed down and wrapped and placed in Sea/Lands to be sent to a vendor for free release and disposal. In the winter of 93 the proposed quotations on the Spent Fuel Rerack Processing were all rejected. With the rerack job starting in March of 94 and the closing of Barnwell in July we were faced with what to do with the racks. Processing of the existing racks were required since if the racks were sent to Barnwell for burial intact the cost would be prohibitive, that is, if Barnwell would have stayed open. If the racks were sent to a smelter, such as Scientific Ecology Group (SEG), there are restrictions on the length of the components that can go through the smelter. If SEG were to do the rack processing (sectioning) at their facility, the cost would also be prohibitive and they would not be in a position to receive the racks until June, 1995. Therefore, bid specifications were requested for on-site volume reduction processing of the existing spent fuel storage racks, with further ultimate disposal to be performed by SEG. The processing of the racks included piping and supports. Volume reduction (VR) was an issue in the evaluation since after this process the racks were to be shipped to SEG. If a low VR ratio option was chosen, OPPD would need a significant number of shipping containers and required more radwaste shipments versus if a high VR ratio option were chosen.

  6. Management Of Hanford KW Basin Knockout Pot Sludge As Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Raymond, R. E. [CH2M HIll Plateau Remediation Company, Richland, WA (United States); Evans, K. M. [AREVA, Avignon (France)

    2012-10-22T23:59:59.000Z

    CH2M HILL Plateau Remediation Company (CHPRC) and AREVA Federal Services, LLC (AFS) have been working collaboratively to develop and deploy technologies to remove, transport, and interim store remote-handled sludge from the 10S-K West Reactor Fuel Storage Basin on the U.S. Department of Energy (DOE) Hanford Site near Richland, WA, USA. Two disposal paths exist for the different types of sludge found in the K West (KW) Basin. One path is to be managed as Spent Nuclear Fuel (SNF) with eventual disposal at an SNF at a yet to be licensed repository. The second path will be disposed as remote-handled transuranic (RH-TRU) waste at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. This paper describes the systems developed and executed by the Knockout Pot (KOP) Disposition Subproject for processing and interim storage of the sludge managed as SNF, (i.e., KOP material).

  7. Spent fuel test - Climax: technical measurements. Interim report, fiscal year 1981

    SciTech Connect (OSTI)

    Patrick, W.C.; Ballou, L.B.; Butkovich, T.R.

    1982-04-30T23:59:59.000Z

    The Spent Fuel Test-Climax (SFT-C) is located 420 m below surface in the Climax granite stock on the Nevada Test Site. Eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized from April to May 1980, initiating the 3- to 5-year-duration test. The SFT-C operational objective of demonstrating the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner has been met. Technical objectives of the test led to development of a technical measurements program, which is the subject of this report. Geotechnical, seismological, and test status data have been recorded on a continuing basis for the first 1-1/2 years of the test on more than 900 channels. Much of the acquired data are now available for analysis and are presented here. Highlights of activities this year include completion of site characterization field work, major modifications to the data acquisition and the management systems, and the addition of instrument evaluation as an explicit objective of the test.

  8. Radionuclide release rates from spent fuel for performance assessment modeling

    SciTech Connect (OSTI)

    Curtis, D.B.

    1994-11-01T23:59:59.000Z

    In a scenario of aqueous transport from a high-level radioactive waste repository, the concentration of radionuclides in water in contact with the waste constitutes the source term for transport models, and as such represents a fundamental component of all performance assessment models. Many laboratory experiments have been done to characterize release rates and understand processes influencing radionuclide release rates from irradiated nuclear fuel. Natural analogues of these waste forms have been studied to obtain information regarding the long-term stability of potential waste forms in complex natural systems. This information from diverse sources must be brought together to develop and defend methods used to define source terms for performance assessment models. In this manuscript examples of measures of radionuclide release rates from spent nuclear fuel or analogues of nuclear fuel are presented. Each example represents a very different approach to obtaining a numerical measure and each has its limitations. There is no way to obtain an unambiguous measure of this or any parameter used in performance assessment codes for evaluating the effects of processes operative over many millennia. The examples are intended to suggest by example that in the absence of the ability to evaluate accuracy and precision, consistency of a broadly based set of data can be used as circumstantial evidence to defend the choice of parameters used in performance assessments.

  9. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, Victor T. (Idaho Falls, ID)

    1993-01-01T23:59:59.000Z

    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.

  10. Simulated nuclear reactor fuel assembly

    DOE Patents [OSTI]

    Berta, V.T.

    1993-04-06T23:59:59.000Z

    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.

  11. Outlooks of HLW Partitioning Technologies Usage for Recovering of Platinum Metals from Spent Fuel

    SciTech Connect (OSTI)

    Pokhitonov, Y. A.; Estimantovskiy, V.; Romanovski, v.; Zatsev, B.; Todd, T.

    2003-02-24T23:59:59.000Z

    The existing practice of management of high level waste (HLW) generated by NPPs, call for a task of selective separation of the most dangerous long-lived radionuclides with the purpose of their subsequent immobilization and disposal. HLW partitioning allows to reduce substantially the cost of vitrified product storage owing to isolation of the most dangerous radionuclides, such as transplutonium elements (TPE) into separate fractions of small volumes, intended for ultimate storage. By now numerous investigations on partitioning of HLW of various composition have been carried out in many countries and a lot of processes permitting to recover cesium, strontium, TPE and rare earth elements (REE) have been already tested. Apart from enumerated radionuclides, a fair quantity of palladium and rhodium presents in spent fuel, but the problem of these elements recovery has not yet been decided at the operating radiochemical plants. A negative effect of platinum group metals (PGM) occurrence is determined by the formation of separate metal phase, which not only worsens the conditions of glass-melting but also shortens considerably the service life of the equipment. At the same time, the exhaustion of PGMs natural resources may finally lead to such a growth of their costs that the spent nuclear fuel would became a substituting source of these elements industrial production. Allowing above mentioned, it is of interest to develop the technique for ''reactor'' palladium and rhodium recovery process which would be compatible with HLW partitioning and could be realized using the same facilities. In the report the data on platinum metals distribution in spent fuel reprocessing products and the several flowsheets for palladium separation from HLW are presented.

  12. Spent Fuel Test-Climax: technical measurements data management system description and data presentation

    SciTech Connect (OSTI)

    Carlson, R.C.

    1985-08-01T23:59:59.000Z

    The Spent Fuel Test-Climax (SFT-C) was located 420 m below surface in the Climax Stock granite on the Nevada Test Site. The test was conducted under the technical direction of the Lawrence Livermore National Laboratory (LLNL) as part of the Nevada Nuclear Waste Storage Investigations (NNWSI) for the US Department of Energy. Eleven canisters of spent nuclear reactor fuel were emplaced, along with six electrical simulators, in April-May 1980. The spent fuel canisters were retrieved and the electrical simulators de-energized in March-April 1983. During the test, just over 1000 MW-hr of thermal energy was deposited in the site, causing temperature changes 100{sup 0}C near the canisters, and about 5{sup 0} in the tunnels. More than 900 channels of geotechnical, seismological, and test status data were recorded on nearly continuous basis for about 3-1/2 years, ending in September 1983. Most geotechnical instrumentation was known to be temperature sensitive, and thus would require temperature compensation before interpretation. Accordingly, a 10-in. reel of digital tape was off-loaded and shipped to Livermore every 4 to 8 weeks, where the data were verified, organized into 45 one-million-word files, and temperature corrected. The purpose of this report is to document the receipt and processing of the data by LLNL Livermore personnel, present facts about the history of the instruments which may be important to the interpretation of the data, present the data themselves in graphical form for each instrument over its operating lifetime, document the forms and locations in which the data will be archived, and offer the data to the geotechnical community for future use in understanding and predicting the effects of the storage of heat-generating waste in hard rocks such as granite.

  13. Electrolysis cell for reprocessing plutonium reactor fuel

    DOE Patents [OSTI]

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

    1985-01-04T23:59:59.000Z

    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.

  14. Electrolysis cell for reprocessing plutonium reactor fuel

    DOE Patents [OSTI]

    Miller, William E. (Naperville, IL); Steindler, Martin J. (Park Forest, IL); Burris, Leslie (Naperville, IL)

    1986-01-01T23:59:59.000Z

    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.

  15. Rethinking the light water reactor fuel cycle

    E-Print Network [OSTI]

    Shwageraus, Evgeni, 1973-

    2004-01-01T23:59:59.000Z

    The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

  16. Dose reduction improvements in storage basins of spent nuclear fuel

    SciTech Connect (OSTI)

    Huang, Fan-Hsiung F.

    1997-08-13T23:59:59.000Z

    Spent nuclear fuel in storage basins at the Hanford Site has corroded and contaminated basin water, which has leaked into the soil; the fuel also had deposited a layer of radioactive sludge on basin floors. The SNF is to be removed from the basins to protect the nearby Columbia River. Because the radiation level is high, measures have been taken to reduce the background dose rate to as low as reasonably achievable (ALARA) to prevent radiation doses from becoming the limiting factor for removal of the SW in the basins to long-term dry storage. All activities of the SNF Project require application of ALARA principles for the workers. On the basis of these principles dose reduction improvements have been made by first identifying radiological sources. Principal radiological sources in the basin are basin walls, basin water, recirculation piping and equipment. Dose reduction activities focus on cleaning and coating basin walls to permit raising the water level, hydrolasing piping, and placing lead plates. In addition, the transfer bay floor will be refinished to make decontamination easier and reduce worker exposures in the radiation field. The background dose rates in the basin will be estimated before each task commences and after it is completed; these dose reduction data will provide the basis for cost benefit analysis.

  17. Fusion option to dispose of spent nuclear fuel and transuranic elements

    SciTech Connect (OSTI)

    Gohar, Y.

    2000-02-10T23:59:59.000Z

    The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

  18. Report on the Savannah River Site aluminum-based spent nuclear fuel alternatives cost study

    SciTech Connect (OSTI)

    NONE

    1998-12-01T23:59:59.000Z

    Initial estimates of costs for the interim management and disposal of aluminum-based spent nuclear fuel (SNF) were developed during preparation of the Environmental Impact Statement (EIS) on the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. The Task Team evaluated multiple alternatives, assessing programmatic, technical, and schedule risks, and generated life-cycle cost projections for each alternative. The eight technology alternatives evaluated were: direct co-disposal; melt and dilute; reprocessing; press and dilute; glass material oxidation dissolution system (GMODS); electrometallurgical treatment; dissolve and vitrify; and plasma arc. In followup to the Business Plan that was developed to look at SNF dry storage, WSRC prepared an addendum to the cost study. This addendum estimated the costs for the modification and use of an existing (105L) reactor facility versus a greenfield approach for new facilities (for the Direct Co-Disposal and Melt and Dilute alternatives). WSRC assessed the impacts of a delay in reprocessing due to the potential reservation of H-Canyon for other missions (i.e., down blending HEU for commercial use or the conversion of plutonium to either MOX fuel or an immobilized repository disposal form). This report presents the relevant results from these WSRC cost studies, consistent with the most recent project policy, technology implementation, canyon utilization, and inventory assumptions. As this is a summary report, detailed information on the technical alternatives or the cost assumptions raised in each of the above-mentioned cost studies is not provided. A comparison table that briefly describes the bases used for the WSRC analyses is included as Appendix A.

  19. Development of Gd-Enriched Alloys for Spent Nuclear Fuel Applications--Part 1: Preliminary Characterization

    E-Print Network [OSTI]

    DuPont, John N.

    composition for any Gd level. Keywords gadolinium, neutron absorbing material, nuclear criticality safety support, (2) spent nuclear fuel geometry control, and (3) nuclear criticality safety. In additionDevelopment of Gd-Enriched Alloys for Spent Nuclear Fuel Applications--Part 1: Preliminary

  20. BWR spent fuel storage cask performance test. Volume 1. Cask handling experience and decay heat, heat transfer, and shielding data

    SciTech Connect (OSTI)

    McKinnon, M.A.; Doman, J.W.; Tanner, J.E.; Guenther, R.J.; Creer, J.M.; King, C.E.

    1986-02-01T23:59:59.000Z

    This report documents a heat transfer and shielding performance test conducted on a Ridihalgh, Eggers and Associates REA 2023 boiling water reactor (BWR) spent fuel storage cask. The testing effort consisted of three parts: pretest preparations, performance testing, and post-test activities. Pretest preparations included conducting cask handling dry runs and characterizing BWR spent fuel assemblies from Nebraska Public Power District's Cooper Nuclear Station. The performance test matrix included 14 runs consisting of two loadings, two cask orientations, and three backfill environments. Post-test activities included calorimetry and axial radiation scans of selected fuel assemblies, in-basin sipping of each assembly, crud collection, video and photographic scans, and decontamination of the cask interior and exterior.

  1. Deep-Burn Modular Helium Reactor Fuel Development Plan

    SciTech Connect (OSTI)

    McEachern, D

    2002-12-02T23:59:59.000Z

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes the Design Data Needs to: (1) fabricate the coated particle fuel, (2) predict its performance in the reactor core, (3) predict the radionuclide release rates from the reactor core, and (4) predict the performance of spent fuel in a geological repository. The heart of this fuel development plan is Section 6, which describes the development activities proposed to satisfy the DDNs presented in Section 5. The development scope is divided into Fuel Process Development, Fuel Materials Development, Fission Product Transport, and Spent Fuel Disposal. Section 7 describes the facilities to be used. Generally, this program will utilize existing facilities. While some facilities will need to be modified, there is no requirement for major new facilities. Section 8 states the Quality Assurance requirements that will be applied to the development activities. Section 9 presents detailed costs organized by WBS and spread over time. Section 10 presents a list of the types of deliverables that will be prepared in each of the WBS elements. Four Appendices contain supplementary information on: (a) design data needs, (b) the interface with the separations plant, (c) the detailed development schedule, and (d) the detailed cost estimate.

  2. Qualitative human reliability analysis for spent fuel handling

    SciTech Connect (OSTI)

    Brewer, J. D. [Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185-0748 (United States); Amico, P. [Science Applications International Corporation (United States); Cooper, S. E. [United Stated Nuclear Regulatory Commission (United States)

    2006-07-01T23:59:59.000Z

    Human reliability analysis (HRA) methods have been developed primarily to provide information for use in probabilistic risk assessments (PRAs) that analyze nuclear power plant (NPP) operations. Given the original emphasis of these methods, it is understandable that many HRAs have not ventured far from NPP control room applications. Despite this historical focus on the control room, there has been growing interest and discussion regarding the application of HRA methods to other NPP activities such as spent fuel handling (SFH) or operations in different types of facilities. One recently developed HRA method, 'A Technique for Human Event Analysis' (ATHEANA) has been proposed as a promising candidate for diverse applications due to its particular approach for systematically uncovering the dynamic, contextual conditions influencing human performance. This paper describes one successful test of this proposition by presenting portions of a recently completed project in which a scoping study was performed to accomplish the following goals: (1) investigate what should be included in a qualitative HRA for spent fuel and cask handling operations; and (2) demonstrate that the ATHEANA HRA technique can be usefully applied to these operations. The preliminary, scoping qualitative HRA examined, in a generic manner, how human performance of SFH and dry cask storage operations (DCSOs) can plausibly lead to radiological consequences that impact the public and the environment. The study involved the performance of typical, qualitative HRA tasks such as collecting relevant information and the preliminary identification of human failure events or unsafe actions, relevant influences (e.g., performance shaping factors, other contextual factors), event scenario development and categorization of human failure event (HFE) scenario groupings. Information from relevant literature sources was augmented with subject matter expert interviews and analysis of an edited video of selected operations. Elements of NUREG-1792, Good Practices for Implementing Human Reliability Analyses (HRA) and NUREG-1624, Rev. 1, Technical Basis and Implementation Guidelines for A Technique for Human Event Analysis (ATHEANA) formed critical parts of the technical basis for the preliminary analysis. Mis-loading of spent fuel into a cask and dropping of a loaded cask were the two human failure event groupings of primary interest, although all human performance aspects of DCSOs were considered to some extent. Of important note is that HRA is typically performed in the context of a plant-specific PRA study. This analysis was performed without the benefit of the context provided by a larger PRA study, nor was it plant specific, and so it investigated only generic HRA issues relevant to SFH. However, the improved understanding of human performance issues provided by the study will likely enhance the ability to carry out a detailed qualitative HRA for a specific NPP at some point in the future. Furthermore, support was obtained regarding the potential for applying ATHEANA beyond NPP settings. This paper provides a description of the process followed during the analysis, a description of the HFE scenario groupings, discussion regarding general human performance vulnerabilities, and a detailed examination of one HFE scenario developed in the study. (authors)

  3. Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

    SciTech Connect (OSTI)

    Chung, Kiwhan [Los Alamos National Laboratory; Beddingfield, David H. [Los Alamos National Laboratory; Geist, William H. [Los Alamos National Laboratory; Lee, Sang-Yoon [unaffiliated

    2012-07-03T23:59:59.000Z

    A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

  4. US Department of Energy Storage of Spent Fuel and High Level Waste

    SciTech Connect (OSTI)

    Sandra M Birk

    2010-10-01T23:59:59.000Z

    ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and high level waste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

  5. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    SciTech Connect (OSTI)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01T23:59:59.000Z

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems` Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment.

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

    SciTech Connect (OSTI)

    Takeuchi, M.; Koizumi, T. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Inoue, M.; Koyama, S.I. [Japan Atomic Energy Agency, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1393 (Japan)

    2013-07-01T23:59:59.000Z

    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)

  7. Cermet Spent Nuclear Fuel Casks and Waste Packages

    SciTech Connect (OSTI)

    Forsberg, Charles W.; Dole, Leslie R. [Nuclear Science and Technology Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN, 37831-6165 (United States)

    2007-07-01T23:59:59.000Z

    Multipurpose transport, aging, and disposal casks are needed for the management of spent nuclear fuel (SNF). Self-shielded cermet casks can out-perform current SNF casks because of the superior properties of cermets, which consist of encapsulated hard ceramic particulates dispersed in a continuous ductile metal matrix to produce a strong high-integrity, high-thermal conductivity cask. A multi-year, multinational development and testing program has been developing cermet SNF casks made of steel, depleted uranium dioxide, and other materials. Because cermets are the traditional material of construction for armor, cermet casks can provide superior protection against assault. For disposal, cermet waste packages (WPs) with appropriate metals and ceramics can buffer the local geochemical environment to (1) slow degradation of SNF, (2) reduce water flow though the degraded WP, (3) sorb neptunium and other radionuclides that determine the ultimate radiation dose to the public from the repository, and (4) contribute to long-term nuclear criticality control. Finally, new cermet cask fabrication methods have been partly developed to manufacture the casks with the appropriate properties. The results of this work are summarized with references to the detailed reports. (authors)

  8. Packaging and transportation system for K-Basin spent fuel

    SciTech Connect (OSTI)

    Kee, A.T.

    1998-03-03T23:59:59.000Z

    This paper describes the cask/transportation system that was designed, procured and delivered to the Hanford K-Basin site at Richland, Washington. The performance requirements and design of the various components -- cask, trailer with cask tie-down system, and the cask operation equipment for the load-out pit -- will be discussed. The presentation will include the details of the factory acceptance testing and its results. The performance requirements for the cask/transportation system was dictated by the constraints imposed by the large number of high priority shipments and the spent fuel pool environment, and the complex interface requirements with other equipment and facility designs. The results of the testing form the basis for the conclusion that the system satisfies the site performance requirements. The cask/transportation system design was driven by the existing facility constraints and the limitations imposed by the large number of shipments over a short two-year period. This system may be useful information for other DOE facilities that may be or will be in a similar situation.

  9. A NOVEL APPROACH TO SPENT FUEL POOL DECOMMISSIONING

    SciTech Connect (OSTI)

    R. L. Demmer

    2011-04-01T23:59:59.000Z

    The Idaho National Laboratory (INL) has been at the forefront of developing methods to reduce the cost and schedule of deactivating spent fuel pools (SFP). Several pools have been deactivated at the INL using an underwater approach with divers. These projects provided a basis for the INL cooperation with the Dresden Nuclear Power Station Unit 1 SFP (Exelon Generation Company) deactivation. It represents the first time that a commercial nuclear power plant (NPP) SFP was decommissioned using this underwater coating process. This approach has advantages in many aspects, particularly in reducing airborne contamination and allowing safer, more cost effective deactivation. The INL pioneered underwater coating process was used to decommission three SFPs with a total combined pool volume of over 900,000 gallons. INL provided engineering support and shared project plans to successfully initiate the Dresden project. This report outlines the steps taken by INL and Exelon to decommission SFPs using the underwater coating process. The rationale used to select the underwater coating process and the advantages and disadvantages are described. Special circumstances are also discussed, such as the use of a remotely-operated underwater vehicle to visually and radiologically map the pool areas that were not readily accessible. A larger project, the INTEC-603 SFP in-situ (grouting) deactivation, is reviewed. Several specific areas where special equipment was employed are discussed and a Lessons Learned evaluation is included.

  10. Seismic Behavior of Spent Fuel Dry Cask Storage Systems

    SciTech Connect (OSTI)

    Shaukat, Syed K. [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States); Luk, Vincent K. [Sandia National Laboratories, PO Box 5800. Albuquerque, New Mexico 87185-0744 (United States)

    2002-07-01T23:59:59.000Z

    The U. S. Nuclear Regulatory Commission (NRC) is conducting a research program to investigate technical issues concerning the dry cask storage systems of spent nuclear fuel by conducting confirmatory research for establishing criteria and review guidelines for the seismic behavior of these systems. The program focuses on developing 3-D finite element analysis models that address the dynamic coupling of a module/cask, a flexible concrete pad, and an underlying soil/rock foundation, in particular, the soil-structure-interaction. Parametric analyses of the coupled models are performed to include variations in module/cask geometry, site seismicity, underlying soil properties, and cask/pad interface friction. The analyses performed include: 1) a rectangular dry cask module typical of Transnuclear West design at a site in Western USA where high seismicity is expected; 2) a cylindrical dry cask typical of Holtec design at a site in Eastern USA where low seismicity is expected; and 3) a cylindrical dry cask typical of Holtec design at a site in Western USA with medium high seismicity. The paper includes assumptions made in seismic analyses performed, results, and conclusions. (authors)

  11. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect (OSTI)

    Samuel Bays; Ayodeji Alajo

    2010-05-01T23:59:59.000Z

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  12. Analysis of radiation doses from operation of postulated commercial spent fuel transportation systems: Main report

    SciTech Connect (OSTI)

    Schneider, K.J.; Hostick, C.J.; Ross, W.A.; Peterson, R.W.; Smith, R.I.; Stiles, D.L.; Daling, P.M.; Weakley, S.A.; Grinde, R.B.; Young, J.R.

    1987-11-01T23:59:59.000Z

    This report contains a system study of estimated radiation doses to the public and workers resulting from the transport of spent fuel from commercial nuclear power reactors to a geologic repository. The report contains a detailed breakdown of activities and a description of time/distance/dose-rate estimates for each activity within the system. Collective doses are estimated for each of the major activities at the reactor site, in transit, and at the repository receiving facility. Annual individual doses to the maximally exposed individuals or groups of individuals are also estimated. A total of 17 alternatives and subalternatives to the postulated reference transportation system are identified, conceptualized, and their dose-reduction potentials and costs estimated. Resulting ratios of ..delta..cost/..delta..collective system dose for each alternative relative to the postulated reference transportation system are given. Most of the alternatives evaluated are estimated to provide both cost and dose reductions. Major reductions in transportation system dose and cost are estimated to result from using higher-capacity rail and truck casks, and particularly when replacing legalweight truck casks with ''advanced design'' overweight truck casks. The greatest annual dose reduction to the highest exposed individual workers (i.e., at the repository) is estimated to be achieved by using remote handling equipment for the cask handling operations at the repository. Additional shielding is also effective in reducing doses to both radiation workers at the reactor and repository and to transport workers. 69 refs., 36 figs., 156 tabs.

  13. Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor

    SciTech Connect (OSTI)

    Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Boer, B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); SCK-CEN, Boertang 200, BE-2400 Mol (Belgium)

    2012-07-01T23:59:59.000Z

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

  14. Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor

    SciTech Connect (OSTI)

    Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

    2012-04-01T23:59:59.000Z

    Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

  15. Spent fuel disassembly hardware and other non-fuel bearing components: characterization, disposal cost estimates, and proposed repository acceptance requirements

    SciTech Connect (OSTI)

    Luksic, A.T.; McKee, R.W.; Daling, P.M.; Konzek, G.J.; Ludwick, J.D.; Purcell, W.L.

    1986-10-01T23:59:59.000Z

    There are two categories of waste considered in this report. The first is the spent fuel disassembly (SFD) hardware. This consists of the hardware remaining after the fuel pins have been removed from the fuel assembly. This includes end fittings, spacer grids, water rods (BWR) or guide tubes (PWR) as appropriate, and assorted springs, fasteners, etc. The second category is other non-fuel-bearing (NFB) components the DOE has agreed to accept for disposal, such as control rods, fuel channels, etc., under Appendix E of the standard utiltiy contract (10 CFR 961). It is estimated that there will be approximately 150 kg of SFD and NFB waste per average metric ton of uranium (MTU) of spent uranium. PWR fuel accounts for approximately two-thirds of the average spent-fuel mass but only 50 kg of the SFD and NFB waste, with most of that being spent fuel disassembly hardware. BWR fuel accounts for one-third of the average spent-fuel mass and the remaining 100 kg of the waste. The relatively large contribution of waste hardware in BWR fuel, will be non-fuel-bearing components, primarily consisting of the fuel channels. Chapters are devoted to a description of spent fuel disassembly hardware and non-fuel assembly components, characterization of activated components, disposal considerations (regulatory requirements, economic analysis, and projected annual waste quantities), and proposed acceptance requirements for spent fuel disassembly hardware and other non-fuel assembly components at a geologic repository. The economic analysis indicates that there is a large incentive for volume reduction.

  16. Irradiation behavior of metallic fast reactor fuels

    SciTech Connect (OSTI)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01T23:59:59.000Z

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

  17. Expert System analysis of non-fuel assembly hardware and spent fuel disassembly hardware: Its generation and recommended disposal

    SciTech Connect (OSTI)

    Williamson, D.A.

    1991-12-31T23:59:59.000Z

    Almost all of the effort being expended on radioactive waste disposal in the United States is being focused on the disposal of spent Nuclear Fuel, with little consideration for other areas that will have to be disposed of in the same facilities. one area of radioactive waste that has not been addressed adequately because it is considered a secondary part of the waste issue is the disposal of the various Non-Fuel Bearing Components of the reactor core. These hardware components fall somewhat arbitrarily into two categories: Non-Fuel Assembly (NFA) hardware and Spent Fuel Disassembly (SFD) hardware. This work provides a detailed examination of the generation and disposal of NFA hardware and SFD hardware by the nuclear utilities of the United States as it relates to the Civilian Radioactive Waste Management Program. All available sources of data on NFA and SFD hardware are analyzed with particular emphasis given to the Characteristics Data Base developed by Oak Ridge National Laboratory and the characterization work performed by Pacific Northwest Laboratories and Rochester Gas & Electric. An Expert System developed as a portion of this work is used to assist in the prediction of quantities of NFA hardware and SFD hardware that will be generated by the United States` utilities. Finally, the hardware waste management practices of the United Kingdom, France, Germany, Sweden, and Japan are studied for possible application to the disposal of domestic hardware wastes. As a result of this work, a general classification scheme for NFA and SFD hardware was developed. Only NFA and SFD hardware constructed of zircaloy and experiencing a burnup of less than 70,000 MWD/MTIHM and PWR control rods constructed of stainless steel are considered Low-Level Waste. All other hardware is classified as Greater-ThanClass-C waste.

  18. Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques NGSI Research Overview and Update on NDA Techniques

    E-Print Network [OSTI]

    A., V. Mozin, S.J. Tobin, L.W. Cambell, J.R. Cheatham, C.R. Freeman, C.J. Gesh,

    2012-01-01T23:59:59.000Z

    considered one of the 17x17 PWR assemblies from the NGSIplutonium signal because in a PWR spent fuel its content isspectra for a single PWR fuel pin with fresh and spent UO 2

  19. Measurement of Spent Fuel Assemblies - Overview of the Status of the Technology for Initiating Discussion at NATIONAL RESEARCH CENTRE KURCHATOV INSTITUTE June 2013

    SciTech Connect (OSTI)

    SISKIND B.; N /A

    2013-06-03T23:59:59.000Z

    This presentation provides an overview of the status of the technology for the measurement of the fissile material content of spent nuclear reactor fuel. The emphasis is on how the needs of the U.S. Nuclear Regulatory Commission and the International Atomic Energy Agency are met by the available technology and what more needs to be done in this area.

  20. DEMONSTRATION OF LONG-TERM STORAGE CAPABILITY FOR SPENT NUCLEAR FUEL IN L BASIN

    SciTech Connect (OSTI)

    Sindelar, R.; Deible, R.

    2011-04-27T23:59:59.000Z

    The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report by the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The overall conclusion is that the fuel can be stored in L Basin, meeting general safety functions for fuel storage, for an additional 50 years and possibly beyond contingent upon continuation of existing fuel management activities and several augmented program activities. It is concluded that the technical bases and well-founded technologies have been established to store spent nuclear fuel in the L Basin. Methodologies to evaluate the fuel condition and characteristics, and systems to prepare fuel, isolate damaged fuel, and maintain water quality storage conditions have been established. Basin structural analyses have been performed against present NPH criteria. The aluminum fuel storage experience to date, supported by the understanding of the effects of environmental variables on materials performance, demonstrates that storage systems that minimize degradation and provide full retrievability of the fuel up to and greater than 50 additional years will require maintaining the present management programs, and with the recommended augmented/additional activities in this report.

  1. Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

    DOE Patents [OSTI]

    Reichner, P.; Dollard, W.J.

    1991-01-08T23:59:59.000Z

    An electrochemical apparatus is made having a generator section containing axially elongated electrochemical cells, a fresh gaseous feed fuel inlet, a gaseous feed oxidant inlet, and at least one gaseous spent fuel exit channel, where the spent fuel exit channel passes from the generator chamber to combine with the fresh feed fuel inlet at a mixing apparatus, reformable fuel mixture channel passes through the length of the generator chamber and connects with the mixing apparatus, that channel containing entry ports within the generator chamber, where the axis of the ports is transverse to the fuel electrode surfaces, where a catalytic reforming material is distributed near the reformable fuel mixture entry ports. 2 figures.

  2. Synergistic Smart Fuel For In-pile Nuclear Reactor Measurements

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Randall A. Ali; Steven L . Garrett

    2013-10-01T23:59:59.000Z

    In March 2011, an earthquake of magnitude 9.0 on the Richter scale struck Japan with its epicenter on the northeast coast, near the Tohoku region. In addition to the immense physical destruction and casualties across the country, several nuclear power plants (NPP) were affected. It was the Fukushima Daiichi NPP that experienced the most severe and irreversible damage. The earthquake brought the reactors at Fukushima to an automatic shutdown and because the power transmission lines were damaged, emergency diesel generators (EDGs) were activated to ensure that there was continued cooling of the reactors and spent fuel pools. The situation was being successfully managed until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to the reactors.2 At this point, the situation became critical. There was a loss of the sensors and instrumentation within the reactor that could have provided valuable information to guide the operators to make informed decisions and avoid the unfortunate events that followed. In the light of these events, we have developed and tested a potential self-powered thermoacoustic system, which will have the ability to serve as a temperature sensor and can transmit data independently of electronic networks. Such a device is synergistic with the harsh environment of the nuclear reactor as it utilizes the heat from the nuclear fuel to provide the input power.

  3. Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques -- Preliminary Modeling Results Emphasizing Integration among Techniques

    E-Print Network [OSTI]

    Tobin, S. J.

    2010-01-01T23:59:59.000Z

    Spent Nuclear Fuel by Self-Induced X- ray,” Annual Meeting of the Institute of Nuclear Material Management,

  4. Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques -- Preliminary Modeling Results Emphasizing Integration among Techniques

    E-Print Network [OSTI]

    Tobin, S. J.

    2010-01-01T23:59:59.000Z

    for safeguards of LEU and MOX spent fuel,” Internationalsystems in use today (Safeguards Mox Python Detector, 1 Fork

  5. The characteristic assessment of spent ion exchange resin from PUSPATI TRIGA REACTOR (RTP) for immobilization process

    SciTech Connect (OSTI)

    Wahida, Nurul [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor, Malaysia and Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi; Majid, Amran Ab; Irwan, M. N. [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia); Wahab, Mohd Abd; Marzukee, Nik; Paulus, Wilfred; Phillip, Esther; Thanaletchumy [Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor (Malaysia)

    2014-09-03T23:59:59.000Z

    In this paper, spent ion exchange resin generated from PUSPATI TRIGA reactor (RTP) in Malaysian Nuclear Agency were characterized based on the water content, radionuclide content and radionuclide leachability. The result revealed that the water content in the spent resin is 48%. Gamma spectrometry analysis indicated the presence of {sup 134}Cs, {sup 137}Cs, {sup 152}Eu, {sup 54}Mn, {sup 58}Co, {sup 60}Co and {sup 65}Zn. The leachability test shows a small concentrations (<1 Bq/l) of {sup 152}Eu and {sup 134}Cs were leached out from the spent resin while {sup 60}Co activity concentrations slightly exceeded the limit generally used for industrial wastewater i.e. 1 Bq/l. Characterization of spent ion exchange resin sampled from RTP show that this characterization is important as a basis to immobilize this radioactive waste using geopolymer technology.

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

    SciTech Connect (OSTI)

    Cipiti, Benjamin B.; Smith, James D.

    2008-09-01T23:59:59.000Z

    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 the fuel cycle. AcknowledgementThe authors would like to acknowledge the support of Laboratory Directed Research and Development Project 125862 for funding this research.

  7. A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544

    E-Print Network [OSTI]

    Croft, S.

    2012-01-01T23:59:59.000Z

    for Safeguards of LEU and MOX Spent Fuel,” Internationalfuel rods (including some MOX fuel) followed by confirmatoryfresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent

  8. Development of Light Water Reactor Fuels with Enhanced Accident...

    Energy Savers [EERE]

    Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to Congress Development of Light Water Reactor Fuels with Enhanced Accident Tolerance - Report to...

  9. SciTech Connect: Fundamental aspects of nuclear reactor fuel...

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

    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...

  10. Spent fuel storage and waste management fuel cycle optimization using CAFCA

    SciTech Connect (OSTI)

    Brinton, S.; Kazimi, M. [Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge MA 02139 (United States)

    2013-07-01T23:59:59.000Z

    Spent fuel storage modeling is at the intersection of nuclear fuel cycle system dynamics and waste management policy. A model that captures the economic parameters affecting used nuclear fuel storage location options, which complements fuel cycle economic assessment has been created using CAFCA (Code for Advanced Fuel Cycles Assessment) of MIT. Research has also expanded to the study on dependency of used nuclear fuel storage economics, environmental impact, and proliferation risk. Three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module which provides an easy to use interface for education on fuel cycle waste management economic impacts. Storage options costs can be compared to literature values with simple variation available for sensitivity study. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (authors)

  11. 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.

    1997-06-01T23:59:59.000Z

    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.

  12. The effect of the composition of plutonium loaded on the reactivity change and the isotopic composition of fuel produced in a fast reactor

    SciTech Connect (OSTI)

    Blandinskiy, V. Yu., E-mail: blandinsky@mail.ru [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15T23:59:59.000Z

    This paper presents the results of a numerical investigation into burnup and breeding of nuclides in metallic fuel consisting of a mixture of plutonium and depleted uranium in a fast reactor with sodium coolant. The feasibility of using plutonium contained in spent nuclear fuel from domestic thermal reactors and weapons-grade plutonium is discussed. It is shown that the largest production of secondary fuel and the least change in the reactivity over the reactor lifetime can be achieved when employing plutonium contained in spent nuclear fuel from a reactor of the RBMK-1000 type.

  13. Water Chemistry Control System for Recovery of Damaged and Degraded Spent Fuel

    SciTech Connect (OSTI)

    Sindelar, R.; Fisher, D.; Thomas, J.

    2011-02-18T23:59:59.000Z

    The International Atomic Energy Agency (IAEA) and the government of Serbia have led the project cosponsored by the U.S, Russia, European Commission, and others to repackage and repatriate approximately 8000 spent fuel elements from the RA reactor fuel storage basins at the VIN?A Institute of Nuclear Sciences to Russia for reprocessing. The repackaging and transportation activities were implemented by a Russian consortium which includes the Sosny Company, Tekhsnabeksport (TENEX) and Mayak Production Association. High activity of the water of the fuel storage basin posed serious risk and challenges to the fuel removal from storage containers and repackaging for transportation. The risk centered on personnel exposure, even above the basin water, due to the high water activity levels caused by Cs-137 leached from fuel elements with failed cladding. A team of engineers from the U.S. DOE-NNSA's Global Threat Reduction Initiative, the Vinca Institute, and the IAEA performed the design, development, and deployment of a compact underwater water chemistry control system (WCCS) to remove the Cs-137 from the basin water and enable personnel safety above the basin water for repackaging operations. Key elements of the WCCS system included filters, multiple columns containing an inorganic sorbent, submersible pumps and flow meters. All system components were designed to be remotely serviceable and replaceable. The system was assembled and successfully deployed at the Vinca basin to support the fuel removal and repackaging activities. Following the successful operations, the Cs-137 is now safely contained and consolidated on the zeolite sorbent used in the columns of the WCCS, and the fuel has been removed from the basins. This paper reviews the functional requirements, design, and deployment of the WCCS.

  14. Deployment evaluation methodology for the electrometallurgical treatment of DOE-EM spent nuclear fuel

    SciTech Connect (OSTI)

    Dahl, C.A.; Adams, J.P.; Ramer, R.J.

    1998-07-01T23:59:59.000Z

    Part of the Department of Energy (DOE) spent nuclear fuel (SNF) inventory may require some type of treatment to meet acceptance criteria at various disposition sites. The current focus for much of this spent nuclear fuel is the electrometallurgical treatment process under development at Argonne National Laboratory. Potential flowsheets for this treatment process are presented. Deployment of the process for the treatment of the spent nuclear fuel requires evaluation to determine the spent nuclear fuel program need for treatment and compatibility of the spent nuclear fuel with the process. The evaluation of need includes considerations of cost, technical feasibility, process material disposition, and schedule to treat a proposed fuel. A siting evaluation methodology has been developed to account for these variables. A work breakdown structure is proposed to gather life-cycle cost information to allow evaluation of alternative siting strategies on a similar basis. The evaluation methodology, while created specifically for the electrometallurgical evaluation, has been written such that it could be applied to any potential treatment process that is a disposition option for spent nuclear fuel. Future work to complete the evaluation of the process for electrometallurgical treatment is discussed.

  15. FATE Unified Modeling Method for Spent Nuclear Fuel and Sludge Processing, Shipping and Storage - 13405

    SciTech Connect (OSTI)

    Plys, Martin; Burelbach, James; Lee, Sung Jin; Apthorpe, Robert [Fauske and Associates, LLC, 16W070 83rd St., Burr Ridge, IL, 60527 (United States)] [Fauske and Associates, LLC, 16W070 83rd St., Burr Ridge, IL, 60527 (United States)

    2013-07-01T23:59:59.000Z

    A unified modeling method applicable to the processing, shipping, and storage of spent nuclear fuel and sludge has been incrementally developed, validated, and applied over a period of about 15 years at the US DOE Hanford site. The software, FATE{sup TM}, provides a consistent framework for a wide dynamic range of common DOE and commercial fuel and waste applications. It has been used during the design phase, for safety and licensing calculations, and offers a graded approach to complex modeling problems encountered at DOE facilities and abroad (e.g., Sellafield). FATE has also been used for commercial power plant evaluations including reactor building fire modeling for fire PRA, evaluation of hydrogen release, transport, and flammability for post-Fukushima vulnerability assessment, and drying of commercial oxide fuel. FATE comprises an integrated set of models for fluid flow, aerosol and contamination release, transport, and deposition, thermal response including chemical reactions, and evaluation of fire and explosion hazards. It is one of few software tools that combine both source term and thermal-hydraulic capability. Practical examples are described below, with consideration of appropriate model complexity and validation. (authors)

  16. 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.

    2003-02-27T23:59:59.000Z

    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.

  17. Fast facility spent-fuel and waste assay instrument. [Fluorinel Dissolution and Fuel Storage (FAST) Facility

    SciTech Connect (OSTI)

    Eccleston, G.W.; Johnson, S.S.; Menlove, H.O.; Van Lyssel, T.; Black, D.; Carlson, B.; Decker, L.; Echo, M.W.

    1983-01-01T23:59:59.000Z

    A delayed-neutron assay instrument was installed in the Fluorinel Dissolution and Fuel Storage Facility at Idaho National Engineering Laboratory. The dual-assay instrument is designed to measure both spent fuel and waste solids that are produced from fuel processing. A set of waste standards, fabricated by Los Alamos using uranium supplied by Exxon Nuclear Idaho Company, was used to calibrate the small-sample assay region of the instrument. Performance testing was completed before installation of the instrument to determine the effects of uranium enrichment, hydrogenous materials, and neutron poisons on assays. The unit was designed to measure high-enriched uranium samples in the presence of large neutron backgrounds. Measurements indicate that the system can assay low-enriched uranium samples with moderate backgrounds if calibrated with proper standards.

  18. aluminium-clad spent fuel: Topics by E-print Network

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

    2009-05-15 8 Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography CERN Preprints Summary: This paper explores the use of cosmic ray muons to image the...

  19. EA-1117: Management of Spent Nuclear Fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of the proposal for the management of spent nuclear fuel on the U.S. Department of Energy's Oak Ridge Reservation to implement the preferred alternative...

  20. 324 Building spent fuel segments pieces and fragments removal summary report

    SciTech Connect (OSTI)

    SMITH, C L

    2003-01-09T23:59:59.000Z

    As part of the 324 Building Deactivation Project, all Spent Nuclear Fuel (SNF) and Special Nuclear Material were removed. The removal entailed packaging the material into a GNS-12 cask and shipping it to the Central Waste Complex (CWC).

  1. Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

    SciTech Connect (OSTI)

    Koji Shirai

    2006-04-01T23:59:59.000Z

    The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

  2. International management of spent fuel storage : technical alternatives and constraints, topical report

    E-Print Network [OSTI]

    Miller, Marvin M.

    1978-01-01T23:59:59.000Z

    Some of the important technical issues involved in the implementation of a spent fuel storage regime under international auspices are discussed. In particular, we consider: the state of the art as far as the different ...

  3. Conceptual design report for the ICPP spent nuclear fuel dry storage project

    SciTech Connect (OSTI)

    NONE

    1996-07-01T23:59:59.000Z

    The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.

  4. The French national program for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Giraud, J.P.; Demontalembert, J.A. [COGEMA, Velizy-Villacoublay (France)

    1993-12-31T23:59:59.000Z

    From its very beginning, the French national program for spent fuel and HLW management is aimed at the recycling of energetic materials and the safe disposal of nuclear waste. Spent fuel reprocessing is the cornerstone of this program, since it directly opens the way to energetic material recycling, waste minimization and safe conditioning. It is complemented by the HLW management program which is defined by the HLW disposal regulation and the Waste Act issued in 1991.

  5. Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal

    SciTech Connect (OSTI)

    Not Available

    1984-07-01T23:59:59.000Z

    The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

  6. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

    This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

  7. A dynamic fuel cycle analysis for a heterogeneous thorium-DUPIC recycle in CANDU reactors

    SciTech Connect (OSTI)

    Jeong, C. J.; Park, C. J.; Choi, H. [Korea Atomic Energy Research Inst., P.O. Box 150, Yuseong, Daejeon, 305-600 (Korea, Republic of)

    2006-07-01T23:59:59.000Z

    A heterogeneous thorium fuel recycle scenario in a Canada deuterium uranium (CANDU) reactor has been analyzed by the dynamic analysis method. The thorium recycling is performed through a dry process which has a strong proliferation resistance. In the fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides, and fission products of a multiple thorium recycling fuel cycle were estimated and compared to those of the once-through fuel cycle. The analysis results have shown that the heterogeneous thorium fuel cycle can be constructed through the dry process technology. It is also shown that the heterogeneous thorium fuel cycle can reduce the spent fuel inventory and save on the natural uranium resources when compared with the once-through cycle. (authors)

  8. Market driven strategy for acquisition of waste acceptance and transportation services for commercial spent fuel in the United States

    SciTech Connect (OSTI)

    Lemeshewky, W.; Macaluso, C.; Smith, P. [Dept. of Energy, Washington, DC (United States); Teer, B. [JAI Corp., Fairfax, VA (United States)

    1998-05-01T23:59:59.000Z

    The Department of Energy has the responsibility for the shipment of spent nuclear fuel (SNF) from commercial reactors to a Federal facility for storage and/or disposal. DOE has developed a strategy for a market driven approach for the acquisition of transportation services and equipment which will maximize the participation of private industry. To implement this strategy, DOE is planning to issue a Request for Proposal (RFP) for the provision of the required services and equipment to accept SNF from the utilities and transport the SNF to a Federal facility. The paper discusses this strategy and describes the RFP.

  9. Electrolytic recovery of reactor metal fuel

    DOE Patents [OSTI]

    Miller, W.E.; Tomczuk, Z.

    1994-09-20T23:59:59.000Z

    A new electrolytic process and apparatus are provided using sodium, cerium or a similar metal in alloy or within a sodium beta or beta[double prime]-alumina sodium ion conductor to electrolytically displace each of the spent fuel metals except for cesium and strontium on a selective basis from the electrolyte to an inert metal cathode. Each of the metals can be deposited separately. An electrolytic transfer of spent fuel into the electrolyte includes a sodium or cerium salt in the electrolyte with sodium or cerium alloy being deposited on the cathode during the transfer of the metals from the spent fuel. The cathode with the deposit of sodium or cerium alloy is then shunted to an anode and the reverse transfer is carried out on a selective basis with each metal being deposited separately at the cathode. The result is that the sodium or cerium needed for the process is regenerated in the first step and no additional source of these reactants is required. 2 figs.

  10. Electrolytic recovery of reactor metal fuel

    DOE Patents [OSTI]

    Miller, W.E.; Tomczuk, Z.

    1993-02-03T23:59:59.000Z

    This invention is comprised of a new electrolytic process and apparatus using sodium, cerium or a similar metal in an alloy or within a sodium beta or beta-alumina sodium ion conductor to electrolytically displace each of the spent fuel metals except for Cesium and strontium on a selective basis from the electrolyte to an inert metal cathode. Each of the metals can be deposited separately. An electrolytic transfer of spent fuel into the electrolyte includes a sodium or cerium salt in the electrolyte with sodium or cerium alloy being deposited on the cathode during the transfer of the metals from the spent fuel. The cathode with the deposit of sodium or cerium alloy is then changed to an anode and the reverse transfer is carried out on a selective basis with each metal being deposited separately at the cathode. The result is that the sodium or cerium needed for the process is regenerated in the first step and no additional source of these reactants is required.

  11. A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel

    E-Print Network [OSTI]

    Haviland, David

    A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel S. Jacobsson.g. in medicine. This thesis describes a tomographic method developed for measurements on nuclear fuel assemblies of the integrity of the assemblies, i.e. for controlling that all fuel rods are present. The application has been

  12. Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel

    SciTech Connect (OSTI)

    N /A

    2000-08-04T23:59:59.000Z

    DOE is responsible for the safe and efficient management of its sodium-bonded spent nuclear fuel. This fuel contains metallic sodium, a highly reactive material; metallic uranium, which is also reactive; and in some cases, highly enriched uranium. The presence of reactive materials could complicate the process of qualifying and licensing DOE's sodium-bonded spent nuclear fuel inventory for disposal in a geologic repository. Currently, more than 98 percent of this inventory is located at the Idaho National Engineering and Environmental Laboratory (INEEL), near Idaho Falls, Idaho. In addition, in a 1995 agreement with the State of Idaho, DOE committed to remove all spent nuclear fuel from Idaho by 2035. This EIS evaluates the potential environmental impacts associated with the treatment and management of sodium-bonded spent nuclear fuel in one or more facilities located at Argonne National Laboratory-West (ANL-W) at INEEL and either the F-Canyon or Building 105-L at the Savannah River Site (SRS) near Aiken, South Carolina. DOE has identified and assessed six proposed action alternatives in this EIS. These are: (1) electrometallurgical treatment of all fuel at ANL-W, (2) direct disposal of blanket fuel in high-integrity cans with the sodium removed at ANL-W, (3) plutonium-uranium extraction (PUREX) processing of blanket fuel at SRS, (4) melt and dilute processing of blanket fuel at ANL-W, (5) melt and dilute processing of blanket fuel at SRS, and (6) melt and dilute processing of all fuel at ANL-W. In addition, Alternatives 2 through 5 include the electrometallurgical treatment of driver fuel at ANL-W. Under the No Action Alternative, the EIS evaluates both the continued storage of sodium-bonded spent nuclear fuel until the development of a new treatment technology or direct disposal without treatment. Under all of the alternatives, the affected environment is primarily within 80 kilometers (50 miles) of spent nuclear fuel treatment facilities. Analyses indicate little difference in the environmental impacts among alternatives. DOE has identified electrometallurgical treatment as its Preferred Alternative for the treatment and management of all sodium-bonded spent nuclear fuel, except for the Fermi-1 blanket fuel. The No Action Alternative is preferred for the Fermi-1 blanket spent nuclear fuel.

  13. Pyrometallurgical processing of Integral Fast Reactor metal fuels

    SciTech Connect (OSTI)

    Battles, J.E.; Miller, W.E.; Gay, E.C.

    1991-01-01T23:59:59.000Z

    The pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor is now in an advanced state of development. This process involves electrorefining spent fuel with a cadmium anode, solid and liquid cathodes, and a molten salt electrolyte (LiCl-KCl) at 500{degrees}C. The initial process feasibility and flowsheet verification studies have been conducted in a laboratory-scale electrorefiner. Based on these studies, a dual cathode approach has been adopted, where uranium is recovered on a solid cathode mandrel and uranium-plutonium is recovered in a liquid cadmium cathode. Consolidation and purification (salt and cadmium removal) of uranium and uranium-plutonium products from the electrorefiner have been successful. The process is being developed with the aid of an engineering-scale electrorefiner, which has been successfully operated for more than three years. In this electrorefiner, uranium has been electrotransported from the cadmium anode to a solid cathode in 10 kg quantities. Also, anodic dissolution of 10 kg batches of chopped, simulated fuel (U--10% Zr) has been demonstrated. Development of the liquid cadmium cathode for recovering uranium-plutonium is under way.

  14. The benefits of a fast reactor closed fuel cycle in the UK

    SciTech Connect (OSTI)

    Gregg, R.; Hesketh, K. [United Kingdom National Nuclear Laboratory, Springfields Works, Building 709, Salwick, Preston, PR4 0XJ (United Kingdom)

    2013-07-01T23:59:59.000Z

    The work has shown that starting a fast reactor closed fuel cycle in the UK, requires virtually all of Britain's existing and future PWR spent fuel to be reprocessed, in order to obtain the plutonium needed. The existing UK Pu stockpile is sufficient to initially support only a modest SFR 'closed' fleet assuming spent fuel can be reprocessed shortly after discharge (i.e. after two years cooling). For a substantial fast reactor fleet, most Pu will have to originate from reprocessing future spent PWR fuel. Therefore, the maximum fast reactor fleet size will be limited by the preceding PWR fleet size, so scenarios involving fast reactors still require significant quantities of uranium ore indirectly. However, once a fast reactor fuel cycle has been established, the very substantial quantities of uranium tails in the UK would ensure there is sufficient material for several centuries. Both the short and long term impacts on a repository have been considered in this work. Over the short term, the decay heat emanating from the HLW and spent fuel will limit the density of waste within a repository. For scenarios involving fast reactors, the only significant heat bearing actinide content will be present in the final cores, resulting in a 50% overall reduction in decay energy deposited within the repository when compared with an equivalent open fuel cycle. Over the longer term, radiological dose becomes more important. Total radiotoxicity (normalised by electricity generated) is lower for scenarios with Pu recycle after 2000 years. Scenarios involving fast reactors have the lowest radiotoxicity since the quantities of certain actinides (Np, Pu and Am) eventually stabilise. However, total radiotoxicity as a measure of radiological risk does not account for differences in radionuclide mobility once in repository. Radiological dose is dominated by a small number of fission products so is therefore not affected significantly by reactor type or recycling strategy (since the fission product will primarily be a function of nuclear energy generated). However, by reprocessing spent fuel, it is possible to immobilise the fission product in a more suitable waste form that has far more superior in-repository performance. (authors)

  15. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    SciTech Connect (OSTI)

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01T23:59:59.000Z

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions.

  16. Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges

    SciTech Connect (OSTI)

    B. McLeod

    2002-02-28T23:59:59.000Z

    This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M&O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M&O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report summarizes the results of the 2002 Reference SNF Discharge Projection.

  17. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary.

    SciTech Connect (OSTI)

    Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Mo, Tin (U.S. Nuclear Regulatory Commission, Washington, DC); Billone, Michael C. (Argonne National Laboratory, Argonne, IL); Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission, Washington, DC); Coats, Richard Lee; Burtseva, Tatiana (Argonne National Laboratory, Argonne, IL); Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Thompson, Nancy Slater (U.S. Department of Energy, Washington, DC); Hibbs, Russell S. (U.S. Department of Energy, Washington, DC); Gregson, Michael Warren; Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Molecke, Martin Alan; Tsai, Han-Chung (Argonne National Laboratory, Argonne, IL)

    2005-07-01T23:59:59.000Z

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) and supported by both the U.S. Department of Energy and the Nuclear Regulatory Commission.

  18. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Timothy A. Hyde

    2012-06-01T23:59:59.000Z

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  19. Assessing the Feasibility of Using Neutron Resonance Transmission Analysis (NRTA) for Assaying Plutonium in Spent Fuel Assemblies

    SciTech Connect (OSTI)

    D. L. Chichester; J. W. Sterbentz

    2012-07-01T23:59:59.000Z

    Neutron resonance transmission analysis (NRTA) is an active-interrogation nondestructive assay (NDA) technique capable of assaying spent nuclear fuel to determine plutonium content. Prior experimental work has definitively shown the technique capable of assaying plutonium isotope composition in spent-fuel pins to a precision of approximately 3%, with a spatial resolution of a few millimeters. As a Grand Challenge to investigate NDA options for assaying spent fuel assemblies (SFAs) in the commercial fuel cycle, Idaho National Laboratory has explored the feasibility of using NRTA to assay plutonium in a whole SFA. The goal is to achieve a Pu assay precision of 1%. The NRTA technique uses low-energy neutrons from 0.1-40 eV, at the bottom end of the actinide-resonance range, in a time-of-flight arrangement. Isotopic composition is determined by relating absorption of the incident neutrons to the macroscopic cross-section of the actinides of interest in the material, and then using this information to determine the areal density of the isotopes in the SFA. The neutrons used for NRTA are produced using a pulsed, accelerator-based neutron source. Distinguishable resonances exist for both the plutonium (239,240,241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Based on extensive modeling of the problem using Monte Carlo-based simulation codes, our preliminary results suggest that by rotating an SFA to acquire four symmetric views, sufficient neutron transmission can be achieved to assay a SFA. In this approach multiple scan information for the same pins may also be unfolded to potentially allow the determination of plutonium for sub-regions of the assembly. For a 17 ? 17 pressurized water reactor SFA, a simplistic preliminary analysis indicates the mass of 239Pu may be determined with a precision on the order of 5%, without the need for operator-supplied fuel information or operational histories. This paper will present our work to date on this topic, indicate our preliminary findings for a conceptual assay approach, discuss resilience against spoofing, and outline our future plans for evaluating the NRTA technique for SFA plutonium determination.

  20. A source-term method for determining spent-fuel transport cask containment requirements: Executive summary

    SciTech Connect (OSTI)

    Sanders, T.L.; Seager, K.D. (Sandia National Labs., Albuquerque, NM (United States)); Reardon, P.C. (GRAM, Inc., Albuquerque, NM (United States))

    1993-02-01T23:59:59.000Z

    This Executive Summary presents the methodology for determining containment requirements for spent-fuel transport casks under normal and hypothetical accident conditions. Three sources of radioactive material are considered: (1) the spent fuel itself, (2) radioactive material, referred to as CRUD, attached to the outside surfaces of fuel rod cladding, and (3) residual contamination adhering to interior surfaces of the cask cavity. The methodologies for determining the concentrations of freely suspended radioactive materials within a spent-fuel transport cask for these sources are discussed in much greater detail in three companion reports: A Method for Determining the Spent-Fuel Contribution to Transport Cask Containment Requirements,'' Estimate of CRUD Contribution to Shipping Cask Containment Requirements,'' and A Methodology for Estimating the Residual Contamination Contribution to the Source Term in a Spent-Fuel Transport Cask.'' Examples of cask containment requirements that combine the individually determined containment requirements for the three sources are provided, and conclusions from the three companion reports to this Executive Summary are presented.

  1. The EU Approach for Responsible and Safe Management of Spent Fuel and Radioactive Waste - 12118

    SciTech Connect (OSTI)

    Blohm-Hieber, Ute; Necheva, Christina [European Commission, Directorate-General for Energy, Luxembourg L-2920 (Luxembourg)

    2012-07-01T23:59:59.000Z

    In July 2011 legislation on responsible and safe management of spent fuel and radioactive waste was adopted in the European Union (EU). It aims at ensuring a high level of safety, avoiding undue burdens on future generations and enhancing transparency. EU Member States are responsible for the management of their spent fuel and/or radioactive waste. Each Member State remains free to define its fuel cycle policy. The spent fuel can be regarded either as a valuable resource that may be reprocessed or as radioactive waste that is destined for direct disposal. Whatever option is chosen, the disposal of high level waste, separated at reprocessing, or of spent fuel regarded as waste should be considered. The storage of radioactive waste, including long-term storage, is an interim solution, but not an alternative to disposal. To this end, each Member State has to establish, maintain and implement national policy, framework and programme for management of spent fuel and/or radioactive waste in the long term. Member States will invite international peer reviews to ensure that high safety standards are achieved. The EU approach is anchored in internationally endorsed principles and requirements of the IAEA safety standards and the Joint Convention and in this context makes them legally binding and enforceable in the EU. The EU approach of regulating the management of spent fuel and radioactive waste is anchored in the competence of the national regulatory authorities and in the internationally endorsed principles and requirements of the IAEA Safety Standards and the Joint Convention. Member States have to report to the Commission on the implementation of Directive 2011/70/Euratom for the first time by 23 August 2015, and every 3 years thereafter, taking advantage of the review and reporting under the Joint Convention. On the basis of the Member States' reports, the Commission will submit to the European Parliament and the Council a report on progress made and an inventory of radioactive waste and spent fuel present in the EU territory and the future prospects. Directive 2011/70/Euratom is a logical next step after the Council Directive 2009/71/Euratom on the nuclear safety of nuclear installations. The EU is the first major regional actor providing a binding legal framework on nuclear safety and on responsible and safe management of spent fuel and radioactive waste, and thus is a real model to progress spent fuel and waste management in a safe and responsible manner. (authors)

  2. Threat of Hydride Re-orientation to Spent Fuel Integrity During Transportation Accidents: Myth or Reality?

    SciTech Connect (OSTI)

    Rashid, Joe [ANATECH, 5435 Oberlin Drive, San Diego, CA 92121 (United States); Machiels, Albert [EPRI, 3420 Hillview Avenue, Palo Alto, CA 94304 (United States)

    2007-07-01T23:59:59.000Z

    The source-term study conducted by Sandia National Laboratories nearly two decades ago for the spent fuel inventory known at the time, which was in the low-to-medium burnup range ({approx}35 GWd/MTU), showed that the effects of transportation accidents on spent fuel failures, and consequential radioactivity release to the environment, were relatively benign. However, with today's discharged fuel burnups routinely greater than 45 GWd/MTU, potential hydride reorientation during interim dry storage, and its effects on cladding properties, has become one of the primary concerns for spent fuel transportation. Laboratory tests of un-irradiated cladding specimens subjected to heat treatments promoting hydride dissolution followed by re-precipitation in the radial direction have shown that relatively moderate concentrations ({approx}70 ppm) of radial hydrides can significantly degrade cladding ductility, at least at room temperature. The absence of specific data that are relevant to high-burnup spent fuel under dry storage conditions have led to the conjecture, deduced from those tests, that massive cladding failures, possibly resulting in fuel reconfiguration, can be expected during cask drop events. Such conclusions are not borne out by the findings in this paper. The analysis results indicate that cladding failure is bi-modal: a state of failure initiation at the cladding ID remaining as part-wall damage with less than 2% probability of occurrence, and a through-wall failure at a probability of 1 E-5. These results indicate that spent fuel conditions that could promote the formation of radial hydrides during dry storage are not sufficient to produce radial hydrides concentrations of significant levels to cause major threat to spent fuel integrity. It is important to note in this regard that the through-wall cladding failure probability of 1 E-5 is of the same order of magnitude as calculated in the cited Sandia study for low burnup fuel. (authors)

  3. Spectroscopic Monitoring of Spent Nuclear Fuel Reprocessing Streams: An Evaluation of Spent Fuel Solutions via Raman, Visible, and Near-Infrared Spectroscopy

    SciTech Connect (OSTI)

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Johnsen, Amanda M.; Orton, Christopher R.; Peterson, James M.

    2011-09-01T23:59:59.000Z

    The potential of using optical spectroscopic techniques, such as Raman and Visible/Near Infrared (Vis/NIR), for on-line process control and special nuclear materials accountability applications at a spent nuclear fuel reprocessing facility was evaluated. Availability of on-line real-time techniques that directly measure process concentrations of nuclear materials will enhance performance and proliferation resistance of the solvent extraction processes. Further, on-line monitoring of radiochemical streams will also improve reprocessing plant operation and safety. This report reviews current state of development of the spectroscopic on-line monitoring techniques for such solutions. To further examine applicability of optical spectroscopy for monitoring reprocessing solutions, segments of a spent nuclear fuel, with approximate burn-up values of 70 MWd/kgM, were dissolved in concentrated nitric acid and adjusted to varying final concentrations of HNO3. The resulting spent fuel solutions were batch-contacted with tributyl phosphate/dodecane organic solvent. The feed and equilibrium aqueous and loaded organic solutions were subjected to optical measurements. The obtained spectra showed the presence of the quantifiable Raman bands due to NO3- and UO22+ and Vis/NIR bands due to multiple species of Pu(IV), Pu(VI), Np(V), the Np(V)-U(VI) cation-cation complex, and Nd(III) in fuel solutions, justifying spectroscopic techniques as a promising methodology for monitoring spent fuel processing solutions in real-time. Quantitative evaluation of the fuel solution was performed based on spectroscopic measurements and compared to ICP-MS analysis.

  4. CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11T23:59:59.000Z

    Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and at cask-loading-specific conditions could be performed to demonstrate that release is within the allowable leak rates of the cask.

  5. Comparison of selected foreign plans and practices for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

    1990-04-01T23:59:59.000Z

    This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible forPortsmouth/Paducah ProjectPRE-AWARDenergy use isDiscusses Solar

  7. Electrochemical cell apparatus having axially distributed entry of a fuel-spent fuel mixture transverse to the cell lengths

    DOE Patents [OSTI]

    Reichner, Philip (Plum Borough, PA); Dollard, Walter J. (Churchill Borough, PA)

    1991-01-01T23:59:59.000Z

    An electrochemical apparatus (10) is made having a generator section (22) containing axially elongated electrochemical cells (16), a fresh gaseous feed fuel inlet (28), a gaseous feed oxidant inlet (30), and at least one gaseous spent fuel exit channel (46), where the spent fuel exit channel (46) passes from the generator chamber (22) to combine with the fresh feed fuel inlet (28) at a mixing apparatus (50), reformable fuel mixture channel (52) passes through the length of the generator chamber (22) and connects with the mixing apparatus (50), that channel containing entry ports (54) within the generator chamber (22), where the axis of the ports is transverse to the fuel electrode surfaces (18), where a catalytic reforming material is distributed near the reformable fuel mixture entry ports (54).

  8. Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay

    E-Print Network [OSTI]

    Quiter, Brian

    2012-01-01T23:59:59.000Z

    the mass of 239 Pu in a 17x17 PWR fuel assembly with 45 GWd/center of 40 GWd/MTU burn-up PWR fuel assembly with coolingrate for the 11 y cooled PWR fuel was used as a source term

  9. EIS-0279: Spent Nuclear Fuel Management, Aiken, South Carolina | Department

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't Your Destiny:Revised Finding of98-F, Western22,EERE:8: Supplement4: RecordFinalRecord ofof

  10. Performance assessment for the geological disposal of Deep Burn spent fuel using TTBX

    SciTech Connect (OSTI)

    Van den Akker, B.P.; Ahn, J. [Department of Nuclear Engineering, University of California, Berkeley, CA 94720 (United States)

    2013-07-01T23:59:59.000Z

    The behavior of Deep Burn Modular High Temperature Reactor Spent Fuel (DBSF) is investigated in the Yucca Mountain geological repository (YMR) with respect to the annual dose (Sv/yr) delivered to the Reasonably Maximally Exposed Individual (RMEI) from the transport of radionuclides released from the graphite waste matrix. Transport calculations are performed with a novel computer code, TTBX which is capable of modeling transport pathways that pass through heterogeneous geological formations. TTBX is a multi-region extension of the existing single region TTB transport code. Overall the peak annual dose received by the RMEI is seen to be four orders of magnitude lower than the regulatory threshold for exposure, even under pessimistic scenarios. A number of factors contribute to the favorable performance of DBSF. A reduction of one order of magnitude in the peak annual dose received by the RMEI is observed for every order of magnitude increase in the waste matrix lifetime, highlighting the importance of the waste matrix durability and suggesting graphite's utility as a potential waste matrix for the disposal of high-level waste. Furthermore, we see that by incorporating a higher fidelity far-field model the peak annual dose calculated to be received by the RMEI is reduced by two orders of magnitude. By accounting for the heterogeneities of the far field we have simultaneously removed unnecessary conservatisms and improved the fidelity of the transport model. (authors)

  11. The Universal Canister Strategy in Spent Fuel Reprocessing: UC-C a Real Industrial Improvement

    SciTech Connect (OSTI)

    Thomasson, J.; Barithel, S.; Cocaud, A.; Derycke, P.; Pierre, P.

    2003-02-25T23:59:59.000Z

    In commercial nuclear activities, spent fuel back end management is a key issue for nuclear countries as spent fuel represent most of national civil nuclear waste legacy. Ensuring public safety and protection of the environment, now and in the future has been and still remains a major commitment, it is still the subject of thorough development efforts and active public debates. Considerable benefits can be obtained from the Universal Canister strategy as implemented in France in spent fuel treatment and waste conditioning based on reprocessing. COGEMA developed sophisticated waste conditioning processes to simplify High Level and Long Lived Intermediate Level Waste storage and final disposal. Main benefits are: waste stabilization by immobilization and encapsulation; ultimate waste toxicity reduction; drastic ultimate waste volume reduction; and ultimate waste packages standardization.

  12. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect (OSTI)

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01T23:59:59.000Z

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  13. Public information circular for shipments of irradiated reactor fuel. Report for 16 Jul 79-1 May 82

    SciTech Connect (OSTI)

    Not Available

    1982-06-01T23:59:59.000Z

    This circular has been prepared in response to numerous requests for information regarding routes used for the shipment of irradiated reactor (spent) fuel subject to regulation by the Nuclear Regulatory Commission (NRC), and to meet the requirements of Public Law 96-295. The NRC staff must approve such routes prior to their first use. Spent fuel shipment routes, primarily for road transportation, but also including one rail route, are indicated on reproductions of DOT road maps. Also included are the amounts of material shipped during the approximate three year period that safeguards regulations for spent fuel shipments have been effective. In addition, the Commission provided information in this document regarding the NRC's safety and safeguards regulations for spent fuel shipments as well as safeguards incidents regarding same.

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

    SciTech Connect (OSTI)

    Chen, J.; Wen, M. [Institute of Nuclear and New Energy Technology, Tsinghua University, Bejing 10084 (China)

    2013-07-01T23:59:59.000Z

    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)

  15. advanced spent fuel: Topics by E-print Network

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

    Technologies and Information Sciences Websites Summary: gasoline internal combustion engine vehicles to hydrogen fuel cell electric vehicles (FCEVs) is likely, * a Advanced...

  16. Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently20,000 RussianBy: Thomas P.Department of EnergyLessons Learned Report, NNSA, Feb

  17. Huizenga leads safety of spent fuel management, radioactive waste

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA groupTubahq.na.gov Office ofDepartmentAdministrationto Submit a FOIA

  18. Spent Fuel Transportation Risk Assessment | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently AskedEnergyIssues DOE's Nuclear EnergySmartOverviewEnergy Spelman

  19. Safeguards Guidance for Independent Spent Fuel Storage Installations (ISFSI)

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA Approved:AdministrationAnalysis andBHoneywell9/%2ARequest forMod0/%2A21) May

  20. Naval Spent Fuel Rail Shipment Accident Exercise Objectives | Department of

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_CostNSAR - T en Y ear R enewable E nergyEnergy November

  1. Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_CostNSAR - T en Y earEnergy Research andSite TritiumManagement |

  2. Naval Spent Fuel Rail Shipment Accident Exercise Objectives

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently Asked QuestionsDepartment of EnergyofPROTECTINGof NovemberSTATEMENT.|Department

  3. Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently Asked QuestionsDepartment ofDepartment640 FederalDepartment ofNRC's Integrated Strategy

  4. LWR fuel assembly designs for the transmutation of LWR Spent Fuel TRU with FCM and UO{sub 2}-ThO{sub 2} Fuels

    SciTech Connect (OSTI)

    Bae, G.; Hong, S. G. [Department of Nuclear Engineering, KyungHee University, 1732 Deokyoungdaero, Giheung-gu, Yongin, Gyeonggi-do, 446-701 (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    In this paper, transmutation of transuranic (TRU) nuclides from LWR spent fuels is studied by using LWR fuel assemblies which consist of UO{sub 2}-ThO{sub 2} fuel pins and FCM (Fully Ceramic Microencapsulated) fuel pins. TRU from LWR spent fuel is loaded in the kernels of the TRISO particle fuels of FCM fuel pins. In the FCM fuel pins, the TRISO particle fuels are distributed in SiC matrix having high thermal conductivity. The loading patterns of fuel pins and the fuel compositions are searched to have high transmutation rate and feasible neutronic parameters including pin power peaking, temperature reactivity coefficients, and cycle length. All studies are done only in fuel assembly calculation level. The results show that our fuel assembly designs have good transmutation performances without multi-recycling and without degradation of the safety-related neutronic parameters. (authors)

  5. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01T23:59:59.000Z

    Fuels for sodium-cooled fast reactors: US perspective.Pitch to Diameter Sodium-cooled Fast Reactor Simple Movingreactor (GFR), sodium-cooled fast reactor (SFR) and lead-

  6. Draft Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel

    SciTech Connect (OSTI)

    N /A

    1999-07-30T23:59:59.000Z

    This document summarizes the U.S. Department of Energy's ''Draft Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel''. In addition to information concerning the background, purpose and need for the proposed action, and the National Environmental Policy Act process, this summary describes the characteristics of sodium-bonded spent nuclear fuel, the proposed treatment methods, the proposed facilities, the alternatives considered, and the environmental consequences of these alternatives. A glossary is included at the end to assist the reader with some of the technical terms used in this document.

  7. Nuclear forensics: attributing the source of spent fuel used in an RDD event

    E-Print Network [OSTI]

    Scott, Mark Robert

    2005-08-29T23:59:59.000Z

    accurate for measuring ratios of isotopes from the samples. 7 Fig. 2. Overview of RDD material attribution project. 8 The ability to measure trace actinides in spent fuel with an accuracy better than 1% is necessary. In a 1 gram sample of spent fuel..., some of the measured actinides will be present in the microgram range. Mass spectrometry has advanced to the point that isotopic ratios can be measured in the nanogram range for fission products and picogram range for actinides [8,9]. Isotopic...

  8. TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

    SciTech Connect (OSTI)

    DOE

    1997-04-01T23:59:59.000Z

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

  9. PLUTONIUM METALLIC FUELS FOR FAST REACTORS

    SciTech Connect (OSTI)

    STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

    2007-02-07T23:59:59.000Z

    Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

  10. U.S. Plans for the Next Fast Reactor Transmutation Fuels Irradiation Test

    SciTech Connect (OSTI)

    B. A. Hilton

    2007-09-01T23:59:59.000Z

    The U.S. Advanced Fuel Cycle Initiative (AFCI) seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposal and the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is actinide-bearing transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. Metallic alloy and oxide fuel forms are being developed as the near term options for fast reactor implementation.

  11. Optimally moderated nuclear fission reactor and fuel source therefor

    DOE Patents [OSTI]

    Ougouag, Abderrafi M. (Idaho Falls, ID); Terry, William K. (Shelley, ID); Gougar, Hans D. (Idaho Falls, ID)

    2008-07-22T23:59:59.000Z

    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.

  12. Integrated data base for 1993: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 9

    SciTech Connect (OSTI)

    Klein, J.A.; Storch, S.N.; Ashline, R.C. [and others

    1994-03-01T23:59:59.000Z

    The Integrated Data Base (IDB) Program has compiled historic data on inventories and characteristics of both commercial and DOE spent fuel; also, commercial and U.S. government-owned radioactive wastes through December 31, 1992. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest U.S. Department of Energy/Energy Information Administration (DOE/EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste (HLW), transuranic (TRU), waste, low-level waste (LLW), commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) LLW. For most of these categories, current and projected inventories are given through the calendar-year (CY) 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  13. Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10

    SciTech Connect (OSTI)

    Not Available

    1994-12-01T23:59:59.000Z

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

  14. Integrated Data Base for 1992: US spent fuel and radioactive waste inventories, projections, and characteristics. Revision 8

    SciTech Connect (OSTI)

    Payton, M. L.; Williams, J. T.; Tolbert-Smith, M.; Klein, J. A.

    1992-10-01T23:59:59.000Z

    The Integrated Data Base (IDB) Program has compiled current data on inventories and characteristics of commercial spent fuel and both commercial and US government-owned radioactive wastes through December 31, 1991. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration (DOE/EIA) projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional (I/I) activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, environmental restoration wastes, commercial reactor and fuel cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal.

  15. Standard guide for evaluation of materials used in extended service of interim spent nuclear fuel dry storage systems

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of t...

  16. N-Reactor (U-metal) Fuel Characteristics for Disposal Criticality Analysis

    SciTech Connect (OSTI)

    Taylor, Larry Lorin

    2000-05-01T23:59:59.000Z

    DOE-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into nine characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. Additionally, the criticality analysis will also require data to support design of the canister internals, thermal, and radiation shielding. The purpose of this report is to consolidate and provide in a concise format, material and information/data needed to perform supporting analyses to qualify N-Reactor fuels for acceptance into the designated repository. The N Reactor fuels incorporate zirconium cladding and uranium metal with unique fabrication details in terms of physical size, and method of construction. The fuel construction and post-irradiation handling have created attendant issues relative to cladding failure in the underwater storage environment. These fuels were comprised of low-enriched metal (0.947 to 1.25 wt% 235U) that were originally intended to generate weapons-grade plutonium for national defense. Modifications in subsequent fuel design and changes in the mode of reactor operation in later years were focused more toward power production.

  17. PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

    SciTech Connect (OSTI)

    S. T. Khericha

    2007-04-01T23:59:59.000Z

    The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

  18. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

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

    12. Brazil and Venezuela 13. Canada 14. Italy and Germany 15. Japan 16. Chile and Argentina 17. Austria, Germany, and Netherlands 18. Germany, Sweden, and Japan 19. Denmark 20....

  19. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, J.P.

    1992-03-17T23:59:59.000Z

    A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  20. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, John P. (Downers Grove, IL)

    1992-01-01T23:59:59.000Z

    A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

  1. Natural Fueling of a Tokamak Fusion Reactor

    E-Print Network [OSTI]

    Wan, Weigang; Chen, Yang; Perkins, Francis W

    2009-01-01T23:59:59.000Z

    A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

  2. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    SciTech Connect (OSTI)

    Not Available

    1990-03-01T23:59:59.000Z

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs.

  3. A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544

    E-Print Network [OSTI]

    Croft, S.

    2012-01-01T23:59:59.000Z

    to the safe management of spent nuclear fuel (SNF) and theSpent Fuel Library for Assessing Varied Nondestructive Assay Techniques for Nuclear Safeguards,” American Nuclear Society’s Advances in Nuclear Fuel Management

  4. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix C, Savannah River Site Spent Nuclear Fuel Mangement Program

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

    The US Department of Energy (DOE) is engaged in two related decision making processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad programmatic decisions that will have applicability across the DOE complex and describes in detail the purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document, which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program, supports Volume 1 of the EIS. Following the introduction, Chapter 2 contains background information related to the SRS and the framework of environmental regulations pertinent to spent nuclear fuel management. Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement at the SRS, and summarizes their potential environmental consequences. Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel activities could affect. Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel management alternative and describes cumulative impacts. The chapter also contains information on unavoidable adverse impacts, commitment of resources, short-term use of the environment and mitigation measures.

  5. Comparison of REMIX vs. MOX fuel characteristics in multiple recycling in VVER reactor

    SciTech Connect (OSTI)

    Dekusar, V.M.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Puzakov, A.Y. [State Scientific Centre of Russian Federation, Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2013-07-01T23:59:59.000Z

    Multiple recycling of regenerated uranium-plutonium fuel in thermal reactors of VVER-1000 type with high enriched uranium feeding (REMIX-fuel) gives a possibility to terminate the accumulation of spent nuclear fuels (SNF) and Pu and decrease the accumulation of irradiated uranium by an order of magnitude. Results of comparison of VVER-1000 nuclear fuel cycle characteristics vs different fuel types such as UOX, MOX and REMIX-fuel have been presented. REMIX fuel (Regenerated Mixture of U-, Pu oxides) is the mixture of plutonium and uranium extracted from SNF and refined from other actinides and fission products with the addition of enriched uranium to provide the power potential necessary. The savings in terms of uranium quantities and separation works in the nuclear energy system (NES) with reactors using REMIX-fuel compared to the NES with uranium-fuelled reactors are shown to be of about 30% and 8%, respectively. For the NES with thermal reactors partially loaded with MOX-fuel, the uranium and separation works saving of about 14% would be obtained. Production of neptunium and americium in reactors with REMIX-fuel in steady state increases by a factor 3, and production of curium - by 10 compared to the reactors with UOX-fuel. This increase of minor actinide buildup is owed to the multiple recycling of plutonium. It should be noted that in this case all fuel assemblies contain high-background plutonium, and their manufacturing involves an expensive technology. Besides, management of REMIX-fuel will require special protection measures even during the fresh fuel manufacturing phase. The above-said gives ground to state that the use of REMIX fuel would be questionable in economic aspect.

  6. Fuel handling system for a nuclear reactor

    DOE Patents [OSTI]

    Saiveau, James G. (Hickory Hills, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

    1986-01-01T23:59:59.000Z

    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.

  7. Spent fuel management in France: Reprocessing, conditioning, recycling

    SciTech Connect (OSTI)

    Giraud, J.P.; Montalembert, J.A. de [COGEMA, Cedex (France)

    1994-12-31T23:59:59.000Z

    The French energy policy has been based for 20 years on the development of nuclear power. The some 75% share of nuclear in the total electricity generation, representing an annual production of 317 TWh requires full fuel cycle control from the head-end to the waste management. This paper presents the RCR concept (Reprocessing, Conditioning, Recycling) with its industrial implementation. The long lasting experience acquired in reprocessing and MOX fuel fabrication leads to a comprehensive industrial organization with minimized impact on the environment and waste generation. Each 900 MWe PWR loaded with MOX fuel avoids piling up 2,500 m{sup 3} per year of mine tailings. By the year 2000, less than 500 m{sup 3} of high-level and long-lived waste will be annually produced at La Hague for the French program. The fuel cycle facilities and the associated MOX loading programs are ramping-up according to schedule. Thus, the RCR concept is a reality as well as a policy adopted in several countries. Last but not least, RCR represents a strong commitment to non-proliferation as it is the way to fully control and master the plutonium inventory.

  8. Differential die-away technique for determination of the fissile contents in spent fuel assembly

    SciTech Connect (OSTI)

    Lee, Tachoon [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Nartyn T [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory

    2010-01-01T23:59:59.000Z

    Monte Carlo simulations were performed for the differential die-away (DDA) technique to quantify its capability to measure the fissile contents in spent fuel assemblies of 64 different cases in terms of initial enrichment, burnup, and cooling time. The DDA count rate varies according to the contents of fissile isotopes such as {sup 235}U, {sup 239}Pu, and {sup 241}Pu contained in the spent fuel assembly. The effective {sup 239}Pu concept was introduced to quantify the total fissile mass of spent fuel by weighting the relative signal contributions of {sup 235}U and {sup 241}Pu compared to that of {sup 239}Pu. The Monte Carlo simulation results show that the count rate of the DDA instrument for a spent fuel assembly of 4% initial enrichment, 45 GWD/MTU burnup, and 5 year cooling time is {approx} 9.8 x 10{sup 4} counts per second (c/s) with the 100-Hz repeated interrogation pattern of 0 to 10 {micro}s interrogation, 0.2 ms to 1 ms counting time, and 1 x 10{sup 9} n/s neutron source. The {sup 244}Cm neutron background count rate for this counting time scheme is {approx} 1 x 10{sup 4} c/s, and thus the signal to background ratio is {approx}10.

  9. Preliminary waste acceptance criteria for the ICPP spent fuel and waste management technology development program

    SciTech Connect (OSTI)

    Taylor, L.L.; Shikashio, R.

    1993-09-01T23:59:59.000Z

    The purpose of this document is to identify requirements to be met by the Producer/Shipper of Spent Nuclear Fuel/High-LeveL Waste SNF/HLW in order for DOE to be able to accept the packaged materials. This includes defining both standard and nonstandard waste forms.

  10. Operating experience feedback report: Assessment of spent fuel cooling. Volume 12

    SciTech Connect (OSTI)

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F.; Ornstein, H.L.; Pullani, S.V.

    1997-02-01T23:59:59.000Z

    This report documents the results of an independent assessment by a team from the Office of Analysis and Evaluation of Operational Data of spent-fuel-pool (SFP) cooling in operating nuclear power plants. The team assessed the likelihood and consequences of an extended loss of SFP cooling and suggested corrective actions, based on their findings.

  11. Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content

    E-Print Network [OSTI]

    Stafford, Alissa Sarah

    2010-10-12T23:59:59.000Z

    ) in July 2008 and January 2009. These measurements successfully showed that it is possible to measure the Pu x-ray peak at 103.7 keV in PWR spent fuel (~1 percent Pu) using a planar HPGe detector. Prior to these measurement campaigns, the Pu peak has only...

  12. Characterization program management plan for Hanford K basin spent nuclear fuel

    SciTech Connect (OSTI)

    TRIMBLE, D.J.

    1999-07-19T23:59:59.000Z

    The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng.

  13. 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. [JSC - VNIINM Bochvar, Moscow (Russian Federation); Alekseenko, V. [FSUE - Federal Nuclear and Radiation Safety Center, Moscow (Russian Federation); Krivitsky, Y.; Alekseenko, S.; Bondin, V. [FSUE - Mining and Chemical Combine, Zheleznogorsk (Russian Federation)

    2013-07-01T23:59:59.000Z

    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.

  14. Chemical Effects at the Reaction Front in Corroding Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Fortner, Jeffrey A.; Kropf, A. Jeremy; Jerden, James L.; Cunnane, James C. [Chemical Engineering, Argonne National Laboratory, CMT/205, 9700 S. Cass Avenue, Argonne, IL, 60439 (United States)

    2007-07-01T23:59:59.000Z

    Performance assessment models of the U. S. repository at Yucca Mountain, Nevada suggest that neptunium from spent nuclear fuel is a potentially important dose contributor. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of Np is needed to predict the behavior of Np at the fuel surface during aqueous corrosion. Neptunium would most likely be transported as aqueous Np(V) species, but for this to occur it must first be oxidized from the Np(IV) state found within the parent spent nuclear fuel. In this paper we present synchrotron X-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within uranium oxide spent nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the Pu and Np in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the UO{sub 2} matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl (UO{sub 2}{sup 2+}) mineral assemblage that is depleted in Np and Pu relative to the parent fuel. The transition from U(IV) in the fuel to a fully U(VI) character across the corrosion front is not sharp, but occurs over a transition zone of {approx} 50 micrometers. We find evidence of a thin ({approx} 20 micrometer) layer that is enriched in Pu and Np within a predominantly U(IV) environment on the fuel side of the transition zone. These experimental observations are consistent with available data for the standard reduction potentials for NpO{sub 2}{sup +}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples, which indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potential of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. (authors)

  15. A method for determining the spent-fuel contribution to transport cask containment requirements

    SciTech Connect (OSTI)

    Sanders, T.L.; Seager, K.D. [Sandia National Labs., Albuquerque, NM (United States); Rashid, Y.R.; Barrett, P.R. [ANATECH Research Corp., La Jolla, CA (United States); Malinauskas, A.P. [Oak Ridge National Lab., TN (United States); Einziger, R.E. [Pacific Northwest Lab., Richland, WA (United States); Jordan, H. [EG and G Rocky Flats, Inc., Golden, CO (United States). Rocky Flats Plant; Duffey, T.A.; Sutherland, S.H. [APTEK, Inc., Colorado Springs, CO (United States); Reardon, P.C. [GRAM, Inc., Albuquerque, NM (United States)

    1992-11-01T23:59:59.000Z

    This report examines containment requirements for spent-fuel transport containers that are transported under normal and hypothetical accident conditions. A methodology is described that estimates the probability of rod failure and the quantity of radioactive material released from breached rods. This methodology characterizes the dynamic environment of the cask and its contents and deterministically models the peak stresses that are induced in spent-fuel cladding by the mechanical and thermal dynamic environments. The peak stresses are evaluated in relation to probabilistic failure criteria for generated or preexisting ductile tearing and material fractures at cracks partially through the wall in fuel rods. Activity concentrations in the cask cavity are predicted from estimates of the fraction of gases, volatiles, and fuel fines that are released when the rod cladding is breached. Containment requirements based on the source term are calculated in terms of maximum permissible volumetric leak rates from the cask. Calculations are included for representative cask designs.

  16. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    SciTech Connect (OSTI)

    IRWIN, J.J.

    2000-11-18T23:59:59.000Z

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed from the MCO back to the K Basins.

  17. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    SciTech Connect (OSTI)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01T23:59:59.000Z

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  18. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    SciTech Connect (OSTI)

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-10-11T23:59:59.000Z

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels.

  19. Interface agreement for the management of 308 Building Spent Nuclear Fuel. Revision 1

    SciTech Connect (OSTI)

    Danko, A.D.

    1995-12-22T23:59:59.000Z

    The Hanford Site Spent Nuclear Fuel (SNF) Project was formed to manage the SNF at Hanford. Specifically, the mission of the SNF Project on the Hanford Site is to ``provide safe, economic, environmentally sound management of Hanford SNF in a manner which stages it for final disposition.`` The current mission of the Fuel Fabrication Facilities Transition Project (FFFTP) is to transition the 308 Building for turn over to the Environmental Restoration Contractor for decontamination and decommissioning.

  20. Report to Congress on Plan for Interim Storage of Spent Nuclear...

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

    Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors More Documents & Publications Information Request, "THE REPORT TO THE PRESIDENT...

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

    SciTech Connect (OSTI)

    Shadrin, A.; Dvoeglazov, K.; Ivanov, V. [Bochvar Institute - VNIINM, Moscow (Russian Federation)

    2013-07-01T23:59:59.000Z

    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)

  2. U.S. Spent Nuclear Fuel Data as of December 31, 2002

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron Spin TransitionProgram | Department Home > Nuclear > Spent Nuclear

  3. U.S. Spent Nuclear Fuel Data as of December 31,2002 -Table 2

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron Spin TransitionProgram | Department Home > Nuclear > Spent

  4. U.S. Spent Nuclear Fuel Data as of December 31,2002 Table 3

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What'sis Taking Over OurThe Iron Spin TransitionProgram | Department Home > Nuclear > Spent7 Table

  5. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    SciTech Connect (OSTI)

    D.R. Jackson; G.R. Kiebel

    1999-08-24T23:59:59.000Z

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

  6. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    SciTech Connect (OSTI)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

    2012-07-01T23:59:59.000Z

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)

  7. Investigation of Electrochemical Recovery of Zirconium from Spent Nuclear Fuels

    SciTech Connect (OSTI)

    Michael Simpson; II-Soon Hwang

    2014-06-01T23:59:59.000Z

    This project uses both modeling and experimental studies to design optimal electrochemical technology methods for recovery of zirconium from used nuclear fuel rods for more effective waste management. The objectives are to provide a means of efficiently separating zirconium into metallic high-level waste forms and to support development of a process for decontamination of zircaloy hulls to enable their disposal as low- and intermediate-level waste. Modeling work includes extension of a 3D model previously developed by Seoul National University for uranium electrorefining by adding the ability to predict zirconium behavior. Experimental validation activities include tests for recovery of zirconium from molten salt solutions and aqueous tests using surrogate materials. *This is a summary of the FY 2013 progress for I-NERI project # 2010-001-K provided to the I-NERI office.

  8. Spent Fuel and Waste Management Technology Development Program. Annual progress report

    SciTech Connect (OSTI)

    Bryant, J.W.

    1994-01-01T23:59:59.000Z

    This report provides information on the progress of activities during fiscal year 1993 in the Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) at the Idaho Chemical Processing Plant (ICPP). As a new program, efforts are just getting underway toward addressing major issues related to the fuel and waste stored at the ICPP. The SF&WMTDP has the following principal objectives: Investigate direct dispositioning of spent fuel, striving for one acceptable waste form; determine the best treatment process(es) for liquid and calcine wastes to minimize the volume of high level radioactive waste (HLW) and low level waste (LLW); demonstrate the integrated operability and maintainability of selected treatment and immobilization processes; and assure that implementation of the selected waste treatment process is environmentally acceptable, ensures public and worker safety, and is economically feasible.

  9. 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.

    1982-01-19T23:59:59.000Z

    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.

  10. Effects of air oxidation on the dissolution rate of LWR spent fuel

    SciTech Connect (OSTI)

    Gray, W.J.; Thomas, L.E.; Einziger, R.E.

    1992-11-01T23:59:59.000Z

    Dissolution rates for air-oxidized spent fuel were measured in flowthrough tests. Results from two types of specimens, separated grains and multigrain particles, both in oxidized (U[sub 4]O[sub 9+x]) and unoxidized (UO[sub 2]) conditions indicated only minor effects of oxidation on the surface-area-normalized rates. Similar results were obtained for unirradiated specimens in three different oxidation states (UO[sub 2], U[sub 3]O[sub 7], and U[sub 3]O[sub 8]). These observations have important practical implications for disposal of spent fuel in a geologic repository as well as implications regarding the oxidative dissolution mechanism of UO[sub 2] fuel.

  11. Potential benefits and impacts on the CRWMS transportation system of filling spent fuel shipping casks with depleted uranium silicate glass

    SciTech Connect (OSTI)

    Pope, R.B.; Forsberg, C.W.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

    1996-06-01T23:59:59.000Z

    A new technology, the Depleted Uranium Silicate COntainer Fill System (DUSCOFS), is proposed to improve the performance and reduce the uncertainties of geological disposal of spent nuclear fuel (SNF), thus reducing both radionuclide release rates from the waste package and the potential for repository nuclear criticality events. DUSCOFS may also provide benefits for SNF storage and transport if it is loaded into the container early in the waste management cycle. Assessments have been made of the benefits to be derived by placing depleted uranium silicate (DUS) glass into SNF containers for enhancing repository performance assessment and controlling criticality over geologic times in the repository. Also, the performance, benefits, and impacts which can be derived if the SNF is loaded into a multi-purpose canister with DUS glass at a reactor site have been assessed. The DUSCOFS concept and the benefits to the waste management cycle of implementing DUSCOFS early in the cycle are discussed in this paper.

  12. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    SciTech Connect (OSTI)

    Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States)] [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States); Wagner, John C. [Oak Ridge National Laboratory (United States)] [Oak Ridge National Laboratory (United States)

    2013-07-01T23:59:59.000Z

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup fuel storage and transportation. This paper discusses the staff's preliminary considerations on the safety implication of fuel reconfiguration with respect to nuclear safety (subcriticality control), radiation shielding, containment, the performance of the thermal functions of the packages, and the retrievability of the contents from regulatory perspective. (authors)

  13. Spent fuel sabotage test program, characterization of aerosol dispersal : technical review and analysis supplement.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Lindgren, Eric Richard

    2009-07-01T23:59:59.000Z

    This project seeks to provide vital data required to assess the consequences of a terrorist attack on a spent fuel transportation cask. One such attack scenario involves the use of conical shaped charges (CSC), which are capable of damaging a spent fuel transportation cask. In the event of such an attack, the amount of radioactivity that may be released as respirable aerosols is not known with great certainty. Research to date has focused on measuring the aerosol release from single short surrogate fuel rodlets subjected to attack by a small CSC device in various aerosol chamber designs. The last series of three experiments tested surrogate fuel rodlets made with depleted uranium oxide ceramic pellets in a specially designed double chamber aerosol containment apparatus. This robust testing apparatus was designed to prevent any radioactive release and allow high level radioactive waste disposal of the entire apparatus following testing of actual spent fuel rodlets as proposed. DOE and Sandia reviews of the project to date identified a number of issues. The purpose of this supplemental report is to address and document the DOE review comments and to resolve the issues identified in the Sandia technical review.

  14. On selection and operation of an international interim storage facility for spent nuclear fuel

    E-Print Network [OSTI]

    Burns, Joe, 1966-

    2004-01-01T23:59:59.000Z

    Disposal of post-irradiation fuel from nuclear reactors has been an issue for the nuclear industry for many years. Most countries currently have no long-term disposal strategy in place. Therefore, the concept of an ...

  15. Advanced ceramic cladding for water reactor fuel

    SciTech Connect (OSTI)

    Feinroth, H.

    2000-07-01T23:59:59.000Z

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

  16. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect (OSTI)

    Chodak, P. III

    1996-05-01T23:59:59.000Z

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  17. Thermal performance of a buried nuclear waste storage container storing a hybrid mix of PWR and BWR spent fuel rods; Revision 1

    SciTech Connect (OSTI)

    Johnson, G.L.

    1991-11-01T23:59:59.000Z

    Lawrence Livermore National Laboratory will design, model, and test nuclear waste packages for use at the Nevada Nuclear Waste Storage Repository at Yucca Mountain, Nevada. One such package would store tightly packed spent fuel rods from both pressurized and boiling water reactors. The storage container provides the primary containment of the nuclear waste and the spent fuel rod cladding provides secondary containment. A series of transient conduction and radiation heat transfer analyses was run to determine for the first 1000 yr of storage if the temperature of the tuff at the borehole wall ever falls below 97{degrees}C and whether the cladding of the stored spent fuel ever exceeds 350{degrees}C. Limiting the borehole to temperatures of 97{degrees}C or greater helps minimize corrosion by assuring that no condensed water collects on the container. The 350{degrees}C cladding limit minimizes the possibility of creep- related failure in the spent fuel rod cladding. For a series of packages stored in a 8 {times} 30 m borehole grid where each package contains 10-yr-old spent fuel rods generating 4.74 kW or more, the borehole wall stays above 97{degrees}C for the full 10000-yr analysis period. For the 4.74-kW load, the peak cladding temperature rises to just below the 350{degrees}C limit about 4 years after emplacement. If the packages are stored using the spacing specified in the Site Characterization Plan (15 ft {times} 126 ft), a maximum of 4.1 kW per container may be stored. If the 0.05-m-thick void between the container and the borehole wall is filled with loosely packed bentonite, the peak cladding temperature rises more than 40{degrees}C above the allowed cladding limit. In all cases the dominant heat transfer mode between container components is thermal radiation.

  18. Analysis of radiation doses from operation of postulated commercial spent fuel transportation systems: Analysis of a system containing a monitored retrievable storage facility. Addendum 1

    SciTech Connect (OSTI)

    Smith, R.I.; Daling, P.M. [Pacific Northwest Lab., Richland, WA (United States); Faletti, D.W. [Westinghouse Hanford Co., Richland, WA (United States)

    1992-04-01T23:59:59.000Z

    This addendum report extends the original study of the estimated radiation doses to the public and to workers resulting from transporting spent nuclear fuel from commercial nuclear power reactor stations through the federal waste management system (FWMS), to a system that contains a monitored retrievable storage (MRS) facility. The system concepts and designs utilized herein are consistent with those used in the original study (circa 1985--1987). Because the FWMS design is still evolving, the results of these analyses may no longer apply to the design for casks and cask handling systems that are currently being considered. Four system scenarios are examined and compared with the reference No-MRS scenario (all spent fuel transported directly from the reactors to the western repository in standard-capacity truck and rail casks). In Scenarios 1 and 2, an MRS facility is located in eastern United States and ships either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters. In Scenarios 3 and 4, an MRS facility is located in the western United States and ship either intact fuel assemblies or consolidated fuel rods and compacted assembly hardware in canisters.

  19. 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.

    1982-01-19T23:59:59.000Z

    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).

  20. Performance Assessment Analyses Unique to Department of Energy Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Loo, Henry Hung Yiu; Duguid, J. O.

    2000-06-01T23:59:59.000Z

    This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented.

  1. A GAMMA RAY SCANNING APPROACH TO QUANTIFY SPENT FUEL CASK RADIONUCLIDE CONTENTS

    SciTech Connect (OSTI)

    Branney, S.

    2011-07-01T23:59:59.000Z

    The International Atomic Energy Agency (IAEA) has outlined a need to develop methods of allowing re-verification of LWR spent fuel stored in dry storage casks without the need of a reference baseline measurement. Some scanning methods have been developed, but improvements can be made to readily provide required data for spent fuel cask verification. The scanning process should be conditioned to both confirm the contents and detect any changes due to container/contents degradation or unauthorized removal or tampering. Savannah River National Laboratory and The University of Tennessee are exploring a new method of engineering a high efficiency, cost effective detection system, capable of meeting the above defined requirements in a variety of environmental situations. An array of NaI(Tl) detectors, arranged to form a 'line scan' along with a matching array of 'honeycomb' collimators provide a precisely defined field of view with minimal degradation of intrinsic detection efficiency and with significant scatter rejection. Scanning methods are adapted to net optimum detection efficiency of the combined system. In this work, and with differing detectors, a series of experimental demonstrations are performed that map system spatial performance and counting capability before actual spent fuel cask scans are performed. The data are evaluated to demonstrate the prompt ability to identify missing fuel rods or other content abnormalities. To also record and assess cask tampering, the cask is externally examined utilizing FTIR hyper spectral and other imaging/sensing approaches. This provides dated records and indications of external abnormalities (surface deposits, smears, contaminants, corrosion) attributable to normal degradation or to tampering. This paper will describe the actual gathering of data in both an experimental climate and from an actual spent fuel dry storage cask, and how an evaluation may be performed by an IAEA facility inspector attempting to draw an independent safeguards conclusion concerning the status of the special nuclear material.

  2. DESIGN OF A SYSTEM TO RETRIEVE SLUDGE FROM THE K EAST SPENT FUEL BASIN AT HANFORD

    SciTech Connect (OSTI)

    Twitchell, A.L.; MacLean, G.T.; Ho, Q.T.; Fort, D.L.

    2003-02-27T23:59:59.000Z

    This paper describes the Sludge Retrieval System (SRS), which was designed to safely remove radioactive sludge from the K East spent fuel basin at the 100 K Area of the Hanford Site. Basin water and sludge have the potential to leak to the environment due to the age and condition of the basins. Since the 100 K Area spent fuel basins are located next to the Columbia River, the Spent Nuclear Fuel Project mission includes the safe removal, containment, and transportation of sludge from the basins to a secure storage location. The scope of the SRS includes: A system capable of retrieving sludge from the K East basin floor, pits, and fuel canisters; Separation of debris from sludge, where debris is defined as any material greater than 0.64 cm (0.25 in.) in diameter; Collection of sludge particles in a container that can be transported away from the basin; Modifications to the K East basin to allow installation of the SRS. The SRS was designed by Fluor Federal Services. Changes to the designed system were made by Fluor Hanford as a result of full-scale testing performed after design. This paper discusses this testing, as well as operation and control of the system. Construction and startup testing was initially scheduled to be complete by the end of December 2002. Startup of the system is now expected in April 2003.

  3. Radionuclide release from spent fuel under geologic disposal conditions: An overview of experimental and theoretical work through 1985

    SciTech Connect (OSTI)

    Reimus, P.W.; Simonson, S.A.

    1988-04-01T23:59:59.000Z

    This report presents an overview of experimental and theoretical work on radionuclide release from spent fuel and uranium dioxide (UO/sub 2/) under geologic disposal conditions. The purpose of the report is to provide a source book of information that can be used to develop models that describe radionuclide release from spent fuel waste packages. Modeling activities of this nature will be conducted within the Waste Package Program (WPP) of the Department of Energy's Salt Repository Project (SRP). The topics discussed include experimental methods for investigating radionuclide release, how results have been reported from radionuclide release experiments, theoretical studies of UO/sub 2/ and actinide solubility, results of experimental studies of radionuclide release from spent fuel and UO/sub 2/ (i.e., the effects of different variables on radionuclide release), characteristics of spent fuel pertinent to radionuclide release, and status of modeling of radionuclide release from spent fuel. Appendix A presents tables of data from spent fuel radionuclide release experiments. These data have been digitized from graphs that appear in the literature. An annotated bibliography of literature on spent fuel characterization is provided in Appendix B.

  4. Feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer based on single-element transfer cask experience

    SciTech Connect (OSTI)

    Schmoker, D.S.; Bowser, R.C.

    1993-12-31T23:59:59.000Z

    Spent fuel transportation casks and canister-based storage systems are generally loaded underwater in a nuclear plant`s spent fuel pool/cask loading pit. Several reasons exist for exploring the feasibility of dry cask-to-cask and pool-to-cask spent fuel transfer. These include: the accommodation of plants which do not have sufficient crane capacity to handle large 90 tonne (100 ton) storage canisters or shipping casks, and construction of an MRS without the need for extensive hot cell facilities. In the case of DOE`s ``Multi-Purpose Canister`` (MPC) scenario, use of such a transfer system would allow all plants with adequate transport routes to use large canisters at-reactor, and those without adequate transport routes to use the MRS for loading of large canisters without the need for hot cell facilities. The dry transfer option would not only allow the use of large canisters for all fuel, but would assist DOE in meeting MRS deadlines since licensing and construction of hot-cell facilities significantly affect schedule. This paper reviews the regulatory issues and technical design considerations for a single-element dry transfer system. Also summarized are lessons learned from the TMI-2 fuel transfer system which are directly applicable to the design, testing, startup, and use of a future dry cask-to-cask or pool-to-cask transfer system.

  5. Non-Destructive Spent Fuel Characterization with Semi-Conducting Gallium Arsinde Neutron Imaging Arrays

    SciTech Connect (OSTI)

    Douglas S. McGregor; Holly K. Gersch; Jeffrey D. Sanders; John C. Lee; Mark D. Hammig; Michael R. Hartman; Yong Hong Yang; Raymond T. Klann; Brian Van Der Elzen; John T. Lindsay; Philip A. Simpson

    2002-01-30T23:59:59.000Z

    High resistivity bulk grown GaAs has been used to produce thermal neutron imaging devices for use in neutron radiography and characterizing burnup in spent fuel. The basic scheme utilizes a portable Sb/Be source for monoenergetic (24 keV) neutron radiation source coupled to an Fe filter with a radiation hard B-coated pixellated GaAs detector array as the primary neutron detector. The coated neutron detectors have been tested for efficiency and radiation hardness in order to determine their fitness for the harsh environments imposed by spent fuel. Theoretical and experimental results are presented, showing detector radiation hardness, expected detection efficiency and the spatial resolution from such a scheme. A variety of advanced neutron detector designs have been explored, with experimental results achieving 13% thermal neutron detection efficiency while projecting the possibility of over 30% thermal neutron detection efficiency.

  6. Long-term kinetic effects and colloid formations in dissolution of LWR spent fuels

    SciTech Connect (OSTI)

    Ahn, T.M.

    1996-11-01T23:59:59.000Z

    This report evaluates continuous dissolution and colloid formation during spent-fuel performance under repository conditions in high-level waste disposal. Various observations suggest that reprecipitated layers formed on spent-fuel surfaces may not be protective. This situation may lead to continuous dissolution of highly soluble radionuclides such as C-14, Cl-36, Tc-99, I-129, and Cs-135. However, the diffusion limits of various species involved may retard dissolution significantly. For low-solubility actinides such as Pu-(239+240) or Am-(241+243), various processes regarding colloid formation have been analyzed. The processes analyzed are condensation, dispersion, and sorption. Colloid formation may lead to significant releases of low-solubility actinides. However, because there are only limited data available on matrix dissolution, colloid formation, and solubility limits, many uncertainties still exist. These uncertainties must be addressed before the significance of radionuclide releases can be determined. 118 refs.

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

    DOE Patents [OSTI]

    Heckman, R.A.

    1980-12-19T23:59:59.000Z

    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.

  8. Evaluation of alternative treatments for spent fuel rod consolidation wastes and other miscellaneous commercial transuranic wastes

    SciTech Connect (OSTI)

    Ross, W.A.; Schneider, K.J.; Oma, K.H.; Smith, R.I.; Bunnell, L.R.

    1986-05-01T23:59:59.000Z

    Eight alternative treatments (and four subalternatives) are considered for both existing commercial transuranic wastes and future wastes from spent fuel consolidation. Waste treatment is assumed to occur at a hypothetical central treatment facility (a Monitored Retrieval Storage facility was used as a reference). Disposal in a geologic repository is also assumed. The cost, process characteristics, and waste form characteristics are evaluated for each waste treatment alternative. The evaluation indicates that selection of a high-volume-reduction alternative can save almost $1 billion in life-cycle costs for the management of transuranic and high-activity wastes from 70,000 MTU of spent fuel compared to the reference MRS process. The supercompaction, arc pyrolysis and melting, and maximum volume reduction alternatives are recommended for further consideration; the latter two are recommended for further testing and demonstration.

  9. The release of actinides, cesium, strontium, technetium, and iodine from spent fuel under unsaturated conditions

    SciTech Connect (OSTI)

    Finn, P.A.; Hoh, J.C.; Wolf, S.F. [and others

    1995-12-31T23:59:59.000Z

    Drip tests to measure radionuclide release from spent nuclear fuel are being performed at 90{degrees}C at a drip rate of 0.75 mL/3.5 days; the test conditions are designed to simulate the behavior of spent fuel under the unsaturated and oxidizing conditions expected in the potential repository at Yucca Mountain. This paper presents measurements of the actinide, {sup 137}Cs, {sup 90}Sr, {sup 99}Tc, and {sup 129}I contents in the leachates after 581 days of testing at 90{degrees}C. These values provide an estimate of the source term for the long-lived radionuclide release under these test conditions. Comparisons are made between our results and those of other researchers.

  10. End-to-end calculation of the radiation characteristics of VVER-1000 spent fuel assemblies

    SciTech Connect (OSTI)

    Linge, I. I.; Mitenkova, E. F., E-mail: mit@ibrae.ac.ru; Novikov, N. V. [Russian Academy of Sciences, Nuclear Safety Institute (Russian Federation)

    2012-12-15T23:59:59.000Z

    The results of end-to-end calculation of the radiation characteristics of VVER-1000 spent nuclear fuel are presented. Details of formation of neutron and gamma-radiation sources are analyzed. Distributed sources of different types of radiation are considered. A comparative analysis of calculated radiation characteristics is performed with the use of nuclear data from different ENDF/B and EAF files and ANSI/ANS and ICRP standards.

  11. Annual Radiological Environmental Monitoring Program Report for the Fort St. Vrain Independent Spent Fuel Storage Installation

    SciTech Connect (OSTI)

    Hall, Gregory Graham; Newkirk, Jay Ronald; Borst, Frederick Jon

    2002-02-01T23:59:59.000Z

    This report presents the results of the 2001 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Fort St. Vrain Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the predominant radiation exposure pathway, direct and scattered radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

  12. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Pruett, D.J.; McTaggart, D.R.

    1983-08-31T23:59:59.000Z

    Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

  13. Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1997-03-01T23:59:59.000Z

    A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs.

  14. Comparative analyses of spent nuclear fuel transport modal options: Transport options under existing site constraints

    SciTech Connect (OSTI)

    Brentlinger, L.A.; Hofmann, P.L.; Peterson, R.W.

    1989-08-01T23:59:59.000Z

    The movement of nuclear waste can be accomplished by various transport modal options involving different types of vehicles, transport casks, transport routes, and intermediate intermodal transfer facilities. A series of systems studies are required to evaluate modal/intermodal spent fuel transportation options in a consistent fashion. This report provides total life-cycle cost and life-cycle dose estimates for a series of transport modal options under existing site constraints. 14 refs., 7 figs., 28 tabs.

  15. A comparison of spent fuel assembly control instruments: The Cadarache PYTHON and the Los Alamos Fork

    SciTech Connect (OSTI)

    Bignan, G.; Capsie, J.; Romeyer-Dherbey, J. (CEA Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires); Rinard, P. (Los Alamos National Lab., NM (United States))

    1991-01-01T23:59:59.000Z

    Devices to monitor spent fuel assemblies while stored under water with nondestructive assay methods, have been developed in France and in the United States. Both devices are designed to verify operator's declared values of exposures and cooling-time but the applications and thus the designs of the systems differ. A study, whose results are presented in this paper, has been conducted to compare the features and the performances of the two instruments. 4 refs., 9 figs.

  16. Comparison of potential radiological consequences from a spent-fuel repository and natural uranium deposits

    SciTech Connect (OSTI)

    Wick, O.J.; Cloninger, M.O.

    1980-09-01T23:59:59.000Z

    A general criterion has been suggested for deep geological repositories containing spent fuel - the repositories should impose no greater radiological risk than due to naturally occurring uranium deposits. The following analysis investigates the rationale of that suggestion and determines whether current expectations of spent-fuel repository performance are consistent with such a criterion. In this study, reference spent-fuel repositories were compared to natural uranium-ore deposits. Comparisons were based on intrinsic characteristics, such as radionuclide inventory, depth, proximity to aquifers, and regional distribution, and actual and potential radiological consequences that are now occurring from some ore deposits and that may eventually occur from repositories and other ore deposits. The comparison results show that the repositories are quite comparable to the natural ore deposits and, in some cases, present less radiological hazard than their natural counterparts. On the basis of the first comparison, placing spent fuel in a deep geologic repository apparently reduces the hazard from natural radioactive materials occurring in the earth's crust by locating the waste in impermeable strata without access to oxidizing conditions. On the basis of the second comparison, a repository constructed within reasonable constraints presents no greater hazard than a large ore deposit. It is recommended that if the naturally radioactive environment is to be used as a basis for a criterion regarding repositories, then this criterion should be carefully constructed. The criterion should be based on the radiological quality of the waters in the immediate region of a specific repository, and it should be in terms of an acceptable potential increase in the radiological content of those waters due to the existence of the repository.

  17. The benefits of an advanced fast reactor fuel cycle for plutonium management

    SciTech Connect (OSTI)

    Hannum, W.H.; McFarlane, H.F.; Wade, D.C.; Hill, R.N.

    1996-12-31T23:59:59.000Z

    The United States has no program to investigate advanced nuclear fuel cycles for the large-scale consumption of plutonium from military and civilian sources. The official U.S. position has been to focus on means to bury spent nuclear fuel from civilian reactors and to achieve the spent fuel standard for excess separated plutonium, which is considered by policy makers to be an urgent international priority. Recently, the National Research Council published a long awaited report on its study of potential separation and transmutation technologies (STATS), which concluded that in the nuclear energy phase-out scenario that they evaluated, transmutation of plutonium and long-lived radioisotopes would not be worth the cost. However, at the American Nuclear Society Annual Meeting in June, 1996, the STATS panelists endorsed further study of partitioning to achieve superior waste forms for burial, and suggested that any further consideration of transmutation should be in the context of energy production, not of waste management. 2048 The U.S. Department of Energy (DOE) has an active program for the short-term disposition of excess fissile material and a `focus area` for safe, secure stabilization, storage and disposition of plutonium, but has no current programs for fast reactor development. Nevertheless, sufficient data exist to identify the potential advantages of an advanced fast reactor metallic fuel cycle for the long-term management of plutonium. Advantages are discussed.

  18. Dry halide method for separating the components of spent nuclear fuels

    DOE Patents [OSTI]

    Christian, Jerry Dale (Idaho Falls, ID); Thomas, Thomas Russell (Rigby, ID); Kessinger, Glen F. (Idaho Falls, ID)

    1998-01-01T23:59:59.000Z

    The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

  19. Dry halide method for separating the components of spent nuclear fuels

    DOE Patents [OSTI]

    Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

    1998-06-30T23:59:59.000Z

    The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

  20. Classification of transportation packaging and dry spent fuel storage system components according to importance to safety

    SciTech Connect (OSTI)

    McConnell, J.W., Jr; Ayers, A.L. Jr; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1996-02-01T23:59:59.000Z

    This report provides a graded approach for classification of components used in transportation packaging and dry spent fuel storage systems. This approach provides a method for identifying, the classification of components according to importance to safety within transportation packagings and dry spent fuel storage systems. Record retention requirements are discussed to identify the documentation necessary to validate that the individual components were fabricated in accordance with their assigned classification. A review of the existing regulations pertaining to transportation packagings and dry storage systems was performed to identify current requirements The general types of transportation packagings and dry storage systems were identified. Discussions were held with suppliers and fabricators of packagings and storage systems to determine current practices. The methodology used in this report is based on Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material. This report also includes a list of generic components for each of the general types of transportation packagings and spent fuel storage systems. The safety importance of each component is discussed, and a classification category is assigned.

  1. The corrosion of aluminum-clad spent nuclear fuel in wet basin storage

    SciTech Connect (OSTI)

    Howell, J.P.; Burke, S.D.

    1996-02-20T23:59:59.000Z

    Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980`s and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy.

  2. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    SciTech Connect (OSTI)

    Shropshire, D.E.; Herring, J.S.

    2004-10-03T23:59:59.000Z

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; and low-level waste (LLW) and high-level waste (HLW) disposal.

  3. Dynamic modeling of physical phenomena for probabilistic assessment of spent fuel accidents

    SciTech Connect (OSTI)

    Benjamin, A.S.

    1997-11-01T23:59:59.000Z

    If there should be an accident involving drainage of all the water from a spent fuel pool, the fuel elements will heat up until the heat produced by radioactive decay is balanced by that removed by natural convection to air, thermal radiation, and other means. If the temperatures become high enough for the cladding or other materials to ignite due to rapid oxidation, then some of the fuel might melt, leading to an undesirable release of radioactive materials. The amount of melting is dependent upon the fuel loading configuration and its age, the oxidation and melting characteristics of the materials, and the potential effectiveness of recovery actions. The authors have developed methods for modeling the pertinent physical phenomena and integrating the results with a probabilistic treatment of the uncertainty distributions. The net result is a set of complementary cumulative distribution functions for the amount of fuel melted.

  4. Spent Nuclear Fuel Project technical baseline document. Fiscal year 1995: Volume 1, Baseline description

    SciTech Connect (OSTI)

    Womack, J.C. [Westinghouse Hanford Co., Richland, WA (United States); Cramond, R. [TRW (United States); Paedon, R.J. [SAIC (United States)] [and others

    1995-03-13T23:59:59.000Z

    This document is a revision to WHC-SD-SNF-SD-002, and is issued to support the individual projects that make up the Spent Nuclear Fuel Project in the lower-tier functions, requirements, interfaces, and technical baseline items. It presents results of engineering analyses since Sept. 1994. The mission of the SNFP on the Hanford site is to provide safety, economic, environmentally sound management of Hanford SNF in a manner that stages it to final disposition. This particularly involves K Basin fuel, although other SNF is involved also.

  5. The release of cesium and the actinides from spent fuel under unsaturated conditions

    SciTech Connect (OSTI)

    Finn, P.A.; Hoh, J.C.; Wolf, S.F.; Slater, S.A.; Bates, J.K.

    1995-12-31T23:59:59.000Z

    Tests designed to be similar to the unsaturated and oxidizing conditions expected in the candidate repository at Yucca Mountain are in progress with spent fuel at 90{degree}C. The similarities and the differences in release behavior for {sup 137}Cs during the first 2.6 years and the actinides during the first 1.6 years of testing are presented for tests done with (1) water dripped on the fuel at a rate of 0.075 and 0.75 mL every 3.5 days and (2) in a saturated water vapor environment.

  6. Discovery sheds light on nuclear reactor fuel behavior during...

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

    and Reports Summer Science Writing Internship Discovery sheds light on nuclear reactor fuel behavior during a severe event By Angela Hardin * November 20, 2014 Tweet...

  7. 105-K Basin Material Design Basis Feed Description for Spent Nuclear Fuel (SNF) Project Facilities VOL 1 Fuel

    SciTech Connect (OSTI)

    PACKER, M.J.

    1999-11-04T23:59:59.000Z

    Metallic uranium Spent Nuclear Fuel (SNF) is currently stored within two water filled pools, 105-KE Basin (KE Basin) and 105-KW Basin (KW Basin), at the United States Department of Energy (U.S. DOE) Hanford Site, in southeastern Washington State. The Spent Nuclear Fuel Project (SNF Project) is responsible to DOE for operation of these fuel storage pools and for the 2100 metric tons of SNF materials that they contain. The SNF Project mission includes safe removal and transportation of all SNF from these storage basins to a new storage facility in the 200 East Area. To accomplish this mission, the SNF Project modifies the existing KE Basin and KW Basin facilities and constructs two new facilities: the 100 K Area Cold Vacuum Drying Facility (CVDF), which drains and dries the SNF; and the 200 East Area Canister Storage Building (CSB), which stores the SNF. The purpose of this document is to describe the design basis feed compositions for materials stored or processed by SNF Project facilities and activities. This document is not intended to replace the Hanford Spent Fuel Inventory Baseline (WHC 1994b), but only to supplement it by providing more detail on the chemical and radiological inventories in the fuel (this volume) and sludge. A variety of feed definitions is required to support evaluation of specific facility and process considerations during the development of these new facilities. Six separate feed types have been identified for development of new storage or processing facilities. The approach for using each feed during design evaluations is to calculate the proposed facility flowsheet assuming each feed. The process flowsheet would then provide a basis for material compositions and quantities which are used in follow-on calculations.

  8. Dose Rate Analysis Capability for Actual Spent Fuel Transportation Cask Contents

    SciTech Connect (OSTI)

    Radulescu, Georgeta [ORNL] [ORNL; Lefebvre, Robert A [ORNL] [ORNL; Peplow, Douglas E. [ORNL] [ORNL; Williams, Mark L [ORNL] [ORNL; Scaglione, John M [ORNL] [ORNL

    2014-01-01T23:59:59.000Z

    The approved contents for a U.S. Nuclear Regulatory Commission (NRC) licensed spent nuclear fuel casks are typically based on bounding used nuclear fuel (UNF) characteristics. However, the contents of the UNF canisters currently in storage at independent spent fuel storage installations are considerably heterogeneous in terms of fuel assembly burnup, initial enrichment, decay time, cladding integrity, etc. Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF ST&DARDS) is an integrated data and analysis system that facilitates automated cask-specific safety analyses based on actual characteristics of the as-loaded UNF. The UNF-ST&DARDS analysis capabilities have been recently expanded to include dose rate analysis of as-loaded transportation packages. Realistic dose rate values based on actual canister contents may be used in place of bounding dose rate values to support development of repackaging operations procedures, evaluation of radiation-related transportation risks, and communication with stakeholders. This paper describes the UNF-ST&DARDS dose rate analysis methodology based on actual UNF canister contents and presents sample dose rate calculation results.

  9. Spent nuclear fuel as a waste form for geologic disposal: Assessment and recommendations on data and modeling needs

    SciTech Connect (OSTI)

    Van Luik, A.E.; Apted, M.J.; Bailey, W.J.; Haberman, J.H.; Shade, J.S.; Guenther, R.E.; Serne, R.J.; Gilbert, E.R.; Peters, R.; Williford, R.E.

    1987-09-01T23:59:59.000Z

    This study assesses the status of knowledge pertinent to evaluating the behavior of spent nuclear fuel as a waste form in geologic disposal systems and provides background information that can be used by the DOE to address the information needs that pertain to compliance with applicable standards and regulations. To achieve this objective, applicable federal regulations were reviewed, expected disposal environments were described, the status of spent-fuel modeling was summarized, and information regarding the characteristics and behavior of spent fuel was compiled. This compiled information was then evaluated from a performance modeling perspective to identify further information needs. A number of recommendations were made concerning information still needed to enhance understanding of spent-fuel behavior as a waste form in geologic repositories. 335 refs., 22 figs., 44 tabs.

  10. Determining Plutonium Mass in Spent Fuel with Nondestructive Assay Techniques NGSI Research Overview and Update on NDA Techniques

    E-Print Network [OSTI]

    A., V. Mozin, S.J. Tobin, L.W. Cambell, J.R. Cheatham, C.R. Freeman, C.J. Gesh,

    2012-01-01T23:59:59.000Z

    the target delayed gamma peaks. 3. X-Ray Fluorescence (XRF)The XRF assay technique is being developed by the Los Alamosquantities in the spent fuel. XRF is unique among the other

  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)

    2013-07-01T23:59:59.000Z

    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. A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544

    E-Print Network [OSTI]

    Croft, S.

    2012-01-01T23:59:59.000Z

    Determining Fissile Content in PWR Spent Fuel Assembliesalong the length of several PWR fuel rods (including somebeen studied for a wide range of PWR assembly cases and two

  13. Integral reactor system and method for fuel cells

    DOE Patents [OSTI]

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19T23:59:59.000Z

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  14. ON-LINE MONITORING FOR CONTROL AND SAFEGUARDING OF RADIOCHEMICAL STREAMS AT SPENT FUEL REPROCESSING PLANT

    SciTech Connect (OSTI)

    Bryan, Samuel A.; Levitskaia, Tatiana G.; Lines, Amanda M.; Billing, Justin M.; Casella, Amanda J.; Johnsen, Amanda M.; Peterson, James M.; Thomas, Elizabeth M.

    2009-11-10T23:59:59.000Z

    Advanced techniques that enhance safeguarding of spent fuel reprocessing plants are urgently needed. Our approach is based on the prerequisite that real-time monitoring of solvent extraction flowsheets at a spent fuel reprocessing plant provides the unique capability to quickly detect unwanted manipulations with fissile isotopes present in the radiochemical streams during reprocessing activities. The methods used to monitor these processes must be robust and capable of withstanding harsh radiation and chemical environments. A new on-line monitoring system satisfying these requirements and featuring Raman spectroscopy combined with a Coriolis and conductivity probes recently has been developed by our research team for tank waste retrieval. It provides immediate chemical data and flow parameters of high-level radioactive waste streams with high brine content generated during retrieval activities from nuclear waste storage tanks at the Hanford Site. The nature of the radiochemical streams at the spent fuel reprocessing plant calls for additional spectroscopic information that can be gained by using Vis-NIR capabilities augmenting Raman spectroscopy. A fiber optic Raman probe allows monitoring of high concentration species encountered in both aqueous and organic phases within the UREX suite of flowsheets, including metal oxide ions, such as uranyl, components of the organic solvent, inorganic oxo-anions, and water. Actinides and lanthanides are monitored remotely by Vis-NIR spectroscopy in aqueous and organic phases. In this report, we present our results on spectroscopic measurements of simulant flowsheet solutions and commercial fuels designed to demonstrate the applicability of Raman and Vis-NIR spectroscopic analysis for actual dissolver feed solutions.

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

    DOE Patents [OSTI]

    Fanning, Alan W. (San Jose, CA); Ramsour, Nicholas L. (San Jose, CA)

    1991-01-01T23:59:59.000Z

    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.

  16. The economics of fuel depletion in fast breeder reactor blankets

    E-Print Network [OSTI]

    Brewer, Shelby Templeton

    1972-01-01T23:59:59.000Z

    A fast breeder reactor fuel depletion-economics model was developed and applied to a number of 1000 MWe UMBR case studies, involving radial blanket-radial reflector design, radial blanket fuel management, and sensitivity ...

  17. Optimization of hydride fueled pressurized water reactor cores

    E-Print Network [OSTI]

    Shuffler, Carter Alexander

    2004-01-01T23:59:59.000Z

    This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves ...

  18. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    SciTech Connect (OSTI)

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom)] [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)] [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01T23:59:59.000Z

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  19. Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Guida, Tracey [University of Pittsburgh

    2010-02-01T23:59:59.000Z

    Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

  20. Spent nuclear fuel characterization for a bounding reference assembly for the receiving basin for off-site fuel

    SciTech Connect (OSTI)

    Kahook, S.D.; Garrett, R.L.; Canas, L.R.; Beckum, M.J. [Westinghouse Savannah River, Aiken, SC (United States)

    1995-07-01T23:59:59.000Z

    The Basis for Interim Operation (BIO) for the Receiving Basin for Off-Site Fuel (RBOF) facility at the Department of Energy (DOE) Savannah River Site (SRS) nuclear materials production complex, developed in accordance with draft DOE-STD-0019-93, required a hazard categorization for the safety analysis section as outlined in DOE-STD-1027-92. The RBOF facility was thus established as a Category-2 facility (having potential for significant on-site consequences from a radiological release) as defined in DOE 5480.23. Given the wide diversity of spent nuclear fuel stored in the RBOF facility, which made a detailed assessment of the total nuclear inventory virtually impossible, the categorization required a conservative calculation based on the concept of a hypothetical, bounding reference fuel assembly integrated over the total capacity of the facility. This scheme not only was simple but also precluded a potential delay in the completion of the BIO.

  1. Technical data summary supporting the spent nuclear fuel environment impact statement, March 1994

    SciTech Connect (OSTI)

    Geddes, R.L.; Claxton, R.E.; Lengel, J.D. [and others

    1994-03-01T23:59:59.000Z

    This report has been compiled by the WSRC Nuclear Materials Processing Division`s Planning Section at the request of the Office of Spent Fuel Management and Special Projects (EM-37) to support issuance of the Spent Nuclear Fuel Environmental Impact Statement. Savannah River Site input data evaluates five programmatic options (including {open_quotes}No Action{close_quotes}) ranging up to transfer of all DOE responsibility spent fuel to the SRS. For each option, a range of management/disposition scenarios has been examined. Each case summary provides information relative to the technical proposal, technical issues, environmental impacts, and projected costs for a forty year period (FY-35) when it is assumed that the material will be dispositioned from the SRS. The original issue of the report which was prepared under severe time constraints contained many simplifications and assumptions. Although the revisions have corrected some of the shortcomings of the original report, it is still highly recommended that significant additional study be performed before basing key decisions upon the data contained in this report. The data represents the best effort by a significant group of technical personnel familiar with nuclear materials processing, handling, and storage; but it is likely that careful scrutiny will reveal numerous discrepancies, inconsistencies and omissions. Nor does this report attempt to analyze every potential disposal pathway, but probably establishes the bounds for the most of the viable pathways. The bulk of the effort went into defining the engineering approaches necessary to execute the various mission scenarios which were changed since the last revision. The decision to limit reprocessing to only SRS aluminum clad required a major alteration of the TDS. Collection and/or calculation of much of the various waste, emission, and utility consumption data, so important to an EIS, has been updated since the last revision, but not thoroughly completed.

  2. Research on Spent Fuel Storage and Transportation in CRIEPI (Part 2 Concrete Cask Storage)

    SciTech Connect (OSTI)

    Koji Shirai; Jyunichi Tani; Taku Arai; Masumi Watatu; Hirofumi Takeda; Toshiari Saegusa; Philip L. Winston

    2008-10-01T23:59:59.000Z

    Concrete cask storage has been implemented in the world. At a later stage of storage period, the containment of the canister may deteriorate due to stress corrosion cracking phenomena in a salty air environment. High resistant stainless steels against SCC have been tested as compared with normal stainless steel. Taking account of the limited time-length of environment with certain level of humidity and temperature range, the high resistant stainless steels will survive from SCC damage. In addition, the adhesion of salt from salty environment on the canister surface will be further limited with respect to the canister temperature and angle of the canister surface against the salty air flow in the concrete cask. Optional countermeasure against SCC with respect to salty air environment has been studied. Devices consisting of various water trays to trap salty particles from the salty air were designed to be attached at the air inlet for natural cooling of the cask storage building. Efficiency for trapping salty particles was evaluated. Inspection of canister surface was carried out using an optical camera inserted from the air outlet through the annulus of a concrete cask that has stored real spent fuel for more than 15 years. The camera image revealed no gross degradation on the surface of the canister. Seismic response of a full-scale concrete cask with simulated spent fuel assemblies has been demonstrated. The cask did not tip over, but laterally moved by the earthquake motion. Stress generated on the surface of the spent fuel assemblies during the earthquake motion were within the elastic region.

  3. Truck and rail charges for shipping spent fuel and nuclear waste

    SciTech Connect (OSTI)

    McNair, G.W.; Cole, B.M.; Cross, R.E.; Votaw, E.F.

    1986-06-01T23:59:59.000Z

    The Pacific Northwest Laboratory developed techniques for calculating estimates of nuclear-waste shipping costs and compiled a listing of representative data that facilitate incorporation of reference shipping costs into varius logistics analyses. The formulas that were developed can be used to estimate costs that will be incurred for shipping spent fuel or nuclear waste by either legal-weight truck or general-freight rail. The basic data for this study were obtained from tariffs of a truck carrier licensed to serve the 48 contiguous states and from various rail freight tariff guides. Also, current transportation regulations as issued by the US Department of Transportation and the Nuclear Regulatory Commission were investigated. The costs that will be incurred for shipping spent fuel and/or nuclear waste, as addressed by the tariff guides, are based on a complex set of conditions involving the shipment origin, route, destination, weight, size, and volume and the frequency of shipments, existing competition, and the length of contracts. While the complexity of these conditions is an important factor in arriving at a ''correct'' cost, deregulation of the transportation industry means that costs are much more subject to negotiation and, thus, the actual fee that will be charged will not be determined until a shipping contract is actually signed. This study is designed to provide the baseline data necessary for making comparisons of the estimated costs of shipping spent fuel and/or nuclear wastes by truck and rail transportation modes. The scope of the work presented in this document is limited to the costs incurred for shipping, and does not include packaging, cask purchase/lease costs, or local fees placed on shipments of radioactive materials.

  4. Realization of the German Concept for Interim Storage of Spent Nuclear Fuel - Current Situation and Prospects

    SciTech Connect (OSTI)

    Thomauske, B. R.

    2003-02-25T23:59:59.000Z

    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. New Perspectives for the Spent Nuclear Fuel Radionuclides Release Model in a Deep Geological Repository

    SciTech Connect (OSTI)

    Poinssot, Christophe; Ferry, Cecile; Poulesquen, Arnaud [Department of Physics and Chemistry, Service for the Studies of the Radionuclides Behaviour, CEA, Nuclear Energy Division, CEA Saclay, BP.11, Gif Sur Yvette, F-91650 (France)

    2007-07-01T23:59:59.000Z

    Spent Nuclear Fuel (SNF) source terms are used to define the release rate of radionuclides (RN) in a direct disposal and to assess the performance of this waste form. They classically distinguish between two contributions: (i) the Instant Release Fraction (IRF) of RN which are directly leached when water contacts the fuel, (ii) the slow and long term release of RN which are embedded within the fuel matrix. Recent experimental results bring significant input in our understanding and assessment of both contributions. However, they have not yet been integrated in the definition of the SNF source term. This paper will present the impact on the RN source term of the latest results on the SNF long term evolution and the key remaining scientific issues. (authors)

  6. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Morrow, Charles W.

    2013-01-01T23:59:59.000Z

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level - 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  7. Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask

    SciTech Connect (OSTI)

    Sheryl L. Morton; Philip L. Winston; Toshiari Saegusa; Koji Shirai; Akihiro Sasahara; Takatoshi Hattori

    2006-04-01T23:59:59.000Z

    In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC-17) spent nuclear fuel storage cask as a candidate to study cask performance, because it had been used to store fuel as part of a dry cask storage demonstration project for more than 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. Preliminary cask evaluations performed in 2003 indicated that the cask has no visual degradation. However, a 4-5 mrem/hr step-change in the radiation levels about halfway up the cask and a localized hot spot beneath an upper air vent indicate that there may be variability in the density of the concrete or localized cracking. In 2005, INL and CRIEPI scientists performed additional surveys on the VSC-17 cask. This document summarizes the methods used on the VSC-17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.

  8. Integrated data base report--1995: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    NONE

    1996-12-01T23:59:59.000Z

    The information in this report summarizes the U.S. Department of Energy (DOE) data base for inventories, projections, and characteristics of domestic spent nuclear fuel and radioactive waste. This report is updated annually to keep abreast of continual waste inventory and projection changes in both the government and commercial sectors. Baseline information is provided for DOE program planning purposes and to support DOE program decisions. Although the primary purpose of this document is to provide background information for program planning within the DOE community, it has also been found useful by state and local governments, the academic community, and some private citizens.

  9. What are Spent Nuclear Fuel and High-Level Radioactive Waste ?

    SciTech Connect (OSTI)

    DOE

    2002-12-01T23:59:59.000Z

    Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository.

  10. Radiation Doses to the Public From the Transport of Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Best, R. E.; Maheras, S. J.; Ross, S. S.; Weiner, R.

    2003-02-25T23:59:59.000Z

    This paper reviews issues that have been raised concerning radiological risks and safety of the public exposed to shipments of spent nuclear fuel and high-level radioactive waste to a Yucca Mountain repository. It presents and analyzes the contrasting viewpoints of opponents and proponents, presents facts about radiological exposures and risks, and provides perspective from which to observe the degree of risk that would devolve from the shipments. The paper concludes that the risks to the public's health and safety from being exposed to radiation from the shipments will not be discernable.

  11. Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai [Ishikawajima-Harima Heavy Industries Company Ltd., 1 Shin-Nakaharacho, Isogoku, Yokohama 235-8501 (Japan)

    2002-07-01T23:59:59.000Z

    Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis was to estimate the concrete temperature increase in the case a canister contacts with guide rails in normal storage. It has a possibility that a canister contacts with guide rails during storage period after concrete cask is upended from transfer operation. The temperature increase due to this contact was calculated 5 deg. C at small local area. This result implies that the affect of the contact is very small. This paper addresses that the storage cask concrete is kept its integrity in transfer operation period and normal storage period. (authors)

  12. Hanford Spent Nuclear Fuel Project evaluation of multi-canister overpack venting and monitoring options during staging of K basins fuel

    SciTech Connect (OSTI)

    Wiborg, J.C.

    1995-12-01T23:59:59.000Z

    This engineering study recommends whether multi-canister overpacks containing spent nuclear fuel from the Hanford K Basins should be staged in vented or a sealed, but ventable, condition during staging at the Canister Storage Building prior to hot vacuum conditioning and interim storage. The integrally related issues of MCO monitoring, end point criteria, and assessing the practicality of avoiding venting and Hot Vacuum Conditioning for a portion of the spent fuel are also considered.

  13. Developments of Spent Nuclear Fuel Pyroprocessing Technology at Idaho National Laboratory

    SciTech Connect (OSTI)

    Michael F. Simpson

    2012-03-01T23:59:59.000Z

    This paper summarizes research in used fuel pyroprocessing that has been published by Idaho National Laboratory over the last decade. It includes work done both on treatment of Experimental Breeder Reactor-II and development of advanced technology for potential scale-up and commercialization. Collaborations with universities and other laboratories is included in the cited work.

  14. THORIUM FUEL CYCLES: A GRAPHITE-MODERATED MOLTEN SALT REACTOR

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    THORIUM FUEL CYCLES: A GRAPHITE-MODERATED MOLTEN SALT REACTOR VERSUS A FAST SPECTRUM SOLID FUEL is to compare two main options dedicated to long-term energy production with Thorium: solid fuel with fast its be- haviour until it reaches the 232Th/233U equilibrium from two di erent starting fuels: 232Th

  15. Technical basis for storage of Zircaloy-clad spent fuel in inert gases

    SciTech Connect (OSTI)

    Johnson, A.B. Jr.; Gilbert, E.R.

    1983-09-01T23:59:59.000Z

    This report summarizes the technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. In addition, dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor (PWR) fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved {similar_to}5,000 fuel rods, and {similar_to}600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570{sup 0}C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at {similar_to}70{sup 0}C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the United States. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380{sup 0}C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400{sup 0}C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved.

  16. Nuclear reactor vessel fuel thermal insulating barrier

    DOE Patents [OSTI]

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19T23:59:59.000Z

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  17. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    SciTech Connect (OSTI)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01T23:59:59.000Z

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450MWth DB-HTRs. The TRISO fuel microanalysis covers the gas pressure buildup in a coated fuel particle including helium production, the thermo-mechanical behavior of a CFP, the failure probabilities of CFPs, the temperature distribution in a CPF, and the fission product (FP) transport in a CFP and a graphite. In Chapter VIII, it contains the core design and analysis of sodium cooled fast reactor (SFR) with deep burn HTR reactor. It considers a synergistic combination of the DB-MHR and an SFR burner for a safe and efficient transmutation of the TRUs from LWRs. Chapter IX describes the design and analysis results of the self-cleaning (or self-recycling) HTR core. The analysis is considered zero and 5-year cooling time of the spent LWR fuels.

  18. National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

    SciTech Connect (OSTI)

    NONE

    1995-07-01T23:59:59.000Z

    This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR {section}261.4(a)(4), ``Identification and Listing of Hazardous Waste,`` as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues.

  19. Incorporation of radionuclides in the alteration phases of spent nuclear fuel.

    SciTech Connect (OSTI)

    Buck, E. C.; Kim, C.-W.; Wronkiewicz, D. J.

    1999-08-25T23:59:59.000Z

    Alteration may be expected for spent nuclear fuel exposed to groundwater under oxidizing conditions such as that which exist at the proposed nuclear waste repository at Yucca Mountain, Nevada. The actinide elements released during the corrosion of spent fuel may be incorporated into the structures of secondary U{sup 6+} phases. The incorporation of transuranics into the crystal structures of the alteration products may significantly decrease their mobility. A series of precipitation tests were conducted at 90 C to determine the potential incorporation of Ce{sup 4+} and Nd{sup 3+} (surrogates for Pu{sup 4+} and Am{sup 3+}, respectively) into uranyl phase. Dehydrated schoepite (UO{sub 3}{center_dot}0.8-1.0HP{sub 2}O) was produced by hydrolysis of a uranium oxyacetate solution containing either cerium or neodymium. ICP-MS analysis of the leachant, leachate, and solid phase reaction products which were dissolved in a HNO{sub 3} solution indicates that 26 ppm of Ce was incorporated into dehydrated schoepite. ICP-MS results from the Nd-doped tests indicate significant neodymium incorporation as well, however, the heterogeneous distribution of Nd in the solid phase noted during the AEM/EELS examination implies that neodymium may not incorporate into the structure of dehydrated schoepite.

  20. Geant4 Model Validation of Compton Suppressed System for Process monitoring of Spent Fuel

    SciTech Connect (OSTI)

    Bender, Sarah; Unlu, Kenan; Orton, Christopher R.; Schwantes, Jon M.

    2013-05-01T23:59:59.000Z

    Nuclear material accountancy is of continuous concern for the regulatory, safeguards, and verification communities. In particular, spent nuclear fuel reprocessing facilities pose one of the most difficult accountancy challenges: monitoring highly radioactive, fluid sample streams in near real-time. The Multi-Isotope Process monitor will allow for near-real-time indication of process alterations using passive gamma-ray detection coupled with multivariate analysis techniques to guard against potential material diversion or to enhance domestic process monitoring. The Compton continuum from the dominant 661.7 keV 137Cs fission product peak obscures lower energy lines which could be used for spectral and multivariate analysis. Compton suppression may be able to mitigate the challenges posed by the high continuum caused by scattering. A Monte Carlo simulation using the Geant4 toolkit is being developed to predict the expected suppressed spectrum from spent fuel samples to estimate the reduction in the Compton continuum. Despite the lack of timing information between decay events in the particle management of Geant4, encouraging results were recorded utilizing only the information within individual decays without accounting for accidental coincidences. The model has been validated with single and cascade decay emitters in two steps: as an unsuppressed system and with suppression activated. Results of the Geant4 model validation will be presented.