Science.gov

Sample records for spent fuel regeneration

  1. Regeneration of ammonia borane spent fuel

    SciTech Connect

    Sutton, Andrew David; Davis, Benjamin L; Gordon, John C

    2009-01-01

    A necessary target in realizing a hydrogen (H{sub 2}) economy, especially for the transportation sector, is its storage for controlled delivery, presumably to an energy producing fuel cell. In this vein, the U.S. Department of Energy's Centers of Excellence (CoE) in Hydrogen Storage have pursued different methodologies, including metal hydrides, chemical hydrides, and sorbents, for the expressed purpose of supplanting gasoline's current > 300 mile driving range. Chemical H{sub 2} storage has been dominated by one appealing material, ammonia borane (H{sub 3}N-BH{sub 3}, AB), due to its high gravimetric capacity of H{sub 2} (19.6 wt %) and low molecular weight (30.7 g mol{sup -1}). In addition, AB has both hydridic and protic moieties, yielding a material from which H{sub 2} can be readily released in contrast to the loss of H{sub 2} from C{sub 2}H{sub 6} which is substantially endothermic. As such, a number of publications have described H{sub 2} release from amine boranes, yielding various rates depending on the method applied. The viability of any chemical H{sub 2} storage system is critically dependent on efficient recyclability, but reports on the latter subject are sparse, invoke the use of high energy reducing agents, and suffer from low yields. Our group is currently engaged in trying to find and fully demonstrate an energy efficient regeneration process for the spent fuel from H{sub 2} depleted AB with a minimum number of steps. Although spent fuel composition depends on the dehydrogenation method, we have focused our efforts on the spent fuel resulting from metal-based catalysis, which has thus far shown the most promise. Metal-based catalysts have produced the fastest rates for a single equivalent of H{sub 2} released from AB and up to 2.5 equiv. of H{sub 2} can be produced within 2 hours. While ongoing work is being carried out to tailor the composition of spent AB fuel, a method has been developed for regenerating the predominant product, polyborazylene

  2. Regeneration of ammonia borane from spent fuel materials.

    PubMed

    Summerscales, Owen T; Gordon, John C

    2013-07-28

    A shift to the hydrogen economy requires the development of an effective hydrogen fuel carrier with high volumetric and gravimetric storage capacity. Ammonia borane (AB) has emerged as a leading candidate due to its light weight and multiple protic (N-H) and hydridic (B-H) hydrogens. As a consequence, much work has been directed towards fine tuning the release of H2 from AB, in addition to its regeneration from the dehydrogenated "spent fuel" materials. This review summarizes the development of these regeneration methodologies. PMID:23571860

  3. Efficient regeneration of partially spent ammonia borane fuel

    SciTech Connect

    Davis, Benjamin Lee; Gordon, John C; Stephens, Frances; Dixon, David A; Matus, Myrna H

    2008-01-01

    A necessary target in realizing a hydrogen (H{sub 2}) economy, especially for the transportation sector, is its storage for controlled delivery, presumably to an energy producing fuel cell. In this vein, the U.S. Department of Energy's (DOE) Centers of Excellence (CoE) in Hydrogen Storage have pursued different methodologies, including metal hydrides, chemical hydrides, and sorbents, for the expressed purpose of supplanting gasoline's current > 300 mile driving range. Chemical hydrogen storage has been dominated by one appealing material, ammonia borane (H{sub 3}B-NH{sub 3}, AB), due to its high gravimetric capacity of hydrogen (19.6 wt %) and low molecular weight (30.7 g mol{sup -1}). In addition, AB has both hydridic and protic moieties, yielding a material from which H2 can be readily released. As such, a number of publications have described H{sub 2} release from amine boranes, yielding various rates depending on the method applied. Even though the viability of any chemical hydrogen storage system is critically dependent on efficient recyclability, reports on the latter subject are sparse, invoke the use of high energy reducing agents, and suffer from low yields. For example, the DOE recently decided to no longer pursue the use of NaBH{sub 4} as a H{sub 2} storage material, in part because of inefficient regeneration. We thus endeavored to find an energy efficient regeneration process for the spent fuel from H{sub 2} depleted AB with a minimum number of steps.

  4. Hydrogen storage by boron-nitrogen heterocycles: a simple route for spent fuel regeneration.

    PubMed

    Campbell, Patrick G; Zakharov, Lev N; Grant, Daniel J; Dixon, David A; Liu, Shih-Yuan

    2010-03-17

    We describe a new hydrogen storage platform based on well-defined BN heterocyle materials. Specifically, we demonstrate that regeneration of the spent fuel back to the charged fuel can be accomplished using molecular H(2) and H(-)/H(+) sources. Crystallographic characterization of intermediates along the regeneration pathway confirms our structural assignments and reveals unique bonding changes associated with increasing hydrogen content on boron and nitrogen. Synthetic access to the fully charged BN cyclohexane fuels will now enable investigations of these materials in hydrogen desorption studies. PMID:20214402

  5. Regeneration of ammonia borane spent fuel by direct reaction with hydrazine and liquid ammonia.

    PubMed

    Sutton, Andrew D; Burrell, Anthony K; Dixon, David A; Garner, Edward B; Gordon, John C; Nakagawa, Tessui; Ott, Kevin C; Robinson, J Pierce; Vasiliu, Monica

    2011-03-18

    Ammonia borane (H(3)N-BH(3), AB) is a lightweight material containing a high density of hydrogen (H(2)) that can be readily liberated for use in fuel cell-powered applications. However, in the absence of a straightforward, efficient method for regenerating AB from dehydrogenated polymeric spent fuel, its full potential as a viable H(2) storage material will not be realized. We demonstrate that the spent fuel type derived from the removal of greater than two equivalents of H(2) per molecule of AB (i.e., polyborazylene, PB) can be converted back to AB nearly quantitatively by 24-hour treatment with hydrazine (N(2)H(4)) in liquid ammonia (NH(3)) at 40°C in a sealed pressure vessel. PMID:21415349

  6. Down Select Report of Chemical Hydrogen Storage Materials, Catalysts, and Spent Fuel Regeneration Processes

    SciTech Connect

    Ott, Kevin; Linehan, Sue; Lipiecki, Frank; Aardahl, Christopher L.

    2008-08-24

    The DOE Hydrogen Storage Program is focused on identifying and developing viable hydrogen storage systems for onboard vehicular applications. The program funds exploratory research directed at identifying new materials and concepts for storage of hydrogen having high gravimetric and volumetric capacities that have the potential to meet long term technical targets for onboard storage. Approaches currently being examined are reversible metal hydride storage materials, reversible hydrogen sorption systems, and chemical hydrogen storage systems. The latter approach concerns materials that release hydrogen in endothermic or exothermic chemical bond-breaking processes. To regenerate the spent fuels arising from hydrogen release from such materials, chemical processes must be employed. These chemical regeneration processes are envisioned to occur offboard the vehicle.

  7. Thermally activated persulfate oxidation regeneration of NOM- and MTBE- spent granular activated carbon

    EPA Science Inventory

    Chemical oxidation is a developing technology used to regenerate contaminant-spent GAC. Chemical regeneration of GAC represents a viable option to thermal regeneration methods that are energy intensive resulting in significant consumption of fossil fuels and production of greenho...

  8. The Feasibility Study of Persulfate Oxidation to Regenerating of Spent Granular Activated Carbon

    EPA Science Inventory

    Chemical oxidation is a developing technology used to regenerate contaminant-spent GAC. Chemical regeneration of GAC represents a viable option to thermal regeneration methods that are energy intensive resulting in significant consumption of fossil fuels and production of greenho...

  9. Spent fuel shortage: Facts booklet

    NASA Astrophysics Data System (ADS)

    1980-04-01

    In October 1977, the Department of Energy (DOE) announced a spent nuclear fuel policy where the Government would, under certain conditions, take title of and store spent nuclear fuel from commercial power reactors. The policy is intended to provide spent fuel storage until final disposition is available. The DOE has programs for providing safe, long-term disposal of nuclear waste. The spent fuel storage program is one element of waste management and compliments the disposal program. The costs for spent fuel services are to be fully recovered by the Government from the utilities. This will allow the utilities to confidently consider the costs for disposition of spent fuel in their rate structure. The United States would also store limited amounts of foreign spent fuel to meet nonproliferation objectives. This booklet summarizes information on many aspects of spent fuel storage.

  10. Spent fuel storage. Facts booklet

    SciTech Connect

    1980-04-01

    In October 1977, the Department of Energy (DOE) announced a spent nuclear fuel policy where the Government would, under certain conditions, take title to and store spent nuclear fuel from commercial power reactors. The policy is intended to provide spent fuel storage until final disposition is available. DOE has programs for providing safe, long-term disposal of nuclear waste. The spent fuel storage program is one element of waste management and compliments the disposal program. The costs for spent fuel services are to be fully recovered by the Government from the utilities. This will allow the utilities to confidently consider the costs for disposition of spent fuel in their rate structure. The United States would also store limited amounts of foreign spent fuel to meet nonproliferation objectives. This booklet summarizes information on many aspects of spent fuel storage.

  11. FENTON-DRIVEN CHEMICAL REGENERATION OF MTBE-SPENT GAC

    EPA Science Inventory

    Methyl tert-butyl ether (MTBE)-spent granular activated carbon (GAC) was chemically regenerated utilizing the Fenton mechanism. Two successive GAC regeneration cycles were performed involving iterative adsorption and oxidation processes: MTBE was adsorbed to the GAC, oxidized, r...

  12. Assessment of spent fuel cooling

    SciTech Connect

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F.

    1997-02-01

    The paper presents the methodology, the findings, and the conclusions of a study that was done by the Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) on loss of spent fuel pool cooling. The study involved an examination of spent fuel pool designs, operating experience, operating practices, and procedures. AEOD`s work was augmented in the area of statistics and probabilistic risk assessment by experts from the Idaho Nuclear Engineering Laboratory. Operating experience was integrated into a probabilistic risk assessment to gain insight on the risks from spent fuel pools.

  13. Spent-fuel-storage alternatives

    SciTech Connect

    Not Available

    1980-01-01

    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)

  14. Intermodal transportation of spent fuel

    SciTech Connect

    Elder, H.K.

    1983-09-01

    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.

  15. Active Interrogation for Spent Fuel

    SciTech Connect

    Swinhoe, Martyn Thomas; Dougan, Arden

    2015-11-05

    The DDA instrument for nuclear safeguards is a fast, non-destructive assay, active neutron interrogation technique using an external 14 MeV DT neutron generator for characterization and verification of spent nuclear fuel assemblies.

  16. Transportation of spent MTR fuels

    SciTech Connect

    Raisonnier, D.

    1997-08-01

    This paper gives an overview of the various aspects of MTR spent fuel transportation and provides in particular information about the on-going shipment of 4 spent fuel casks to the United States. Transnucleaire is a transport and Engineering Company created in 1963 at the request of the French Atomic Energy Commission. The company followed the growth of the world nuclear industry and has now six subsidiaries and affiliated companies established in countries with major nuclear programs.

  17. HFIR spent fuel management alternatives

    SciTech Connect

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

    1992-10-15

    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. HFIR spent fuel management alternatives

    SciTech Connect

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

    1992-10-15

    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.

  19. Spent graphite fuel element processing

    SciTech Connect

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

    1981-07-01

    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.

  20. Spent Nuclear Fuel project, project management plan

    SciTech Connect

    Fuquay, B.J.

    1995-10-25

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

  1. Specific features of external heat and mass transfer in the vibration apparatuses used for regenerating spent fuel from nuclear power plants

    NASA Astrophysics Data System (ADS)

    Sapozhnikov, B. G.; Gorbunova, A. M.; Zelenkova, Yu. O.; Sapozhnikov, G. B.; Shiryaeva, N. P.

    2014-06-01

    We present experimental data on the coefficients of heat and mass transfer for freely floating bodies simulating fragments of cladding and large conglomerates of fuel, as well as on the local coefficients of heat and mass transfer over the bed height, which point to high intensity of heat and mass transfer processes that take place in the elements of vibration apparatuses intended for subjecting spent fuel from nuclear power plants to oxidative recrystallization.

  2. Intermodal transfer of spent fuel

    SciTech Connect

    Neuhauser, K. S.; Weiner, R. F.

    1991-01-01

    As a result of the international standardization of containerized cargo handling in ports around the world, maritime shipment handling is particularly uniform. Thus, handier exposure parameters will be relatively constant for ship-truck and ship-rail transfers at ports throughout the world. Inspectors' doses are expected to vary because of jurisdictional considerations. The results of this study should be applicable to truck-to-rail transfers. A study of the movement of spent fuel casks through ports, including the loading and unloading of containers from cargo vessels, afforded an opportunity to estimate the radiation doses to those individuals handling the spent fuels with doses to the public along subsequent transportation routes of the fuel. A number of states require redundant inspections and for escorts over long distances on highways; thus handlers, inspectors, escort personnel, and others who are not normally classified as radiation workers may sustain doses high enough to warrant concern about occupational safety. This paper addresses the question of radiation safety for these workers. Data were obtained during, observation of the offloading of reactor spent fuel (research reactor spent fuel, in this instance) which included estimates of exposure times and distances for handlers, inspectors and other workers during offloading and overnight storage. Exposure times and distance were also for other workers, including crane operators, scale operators, security personnel and truck drivers. RADTRAN calculational models and parameter values then facilitated estimation of the dose to workers during incident-free ship-to-truck transfer of spent fuel.

  3. Criticality of spent reactor fuel

    SciTech Connect

    Harris, D.R.

    1987-01-01

    The storage capacity of spent reactor fuel pools can be greatly increased by consolidation. In this process, the fuel rods are removed from reactor fuel assemblies and are stored in close-packed arrays in a canister or skeleton. An earlier study examined criticality consideration for consolidation of Westinghouse fuel, assumed to be fresh, in canisters at the Millstone-2 spent-fuel pool and in the General Electric IF-300 shipping cask. The conclusions were that the fuel rods in the canister are so deficient in water that they are adequately subcritical, both in normal and in off-normal conditions. One potential accident, the water spill event, remained unresolved in the earlier study. A methodology is developed here for spent-fuel criticality and is applied to the water spill event. The methodology utilizes LEOPARD to compute few-group cross sections for the diffusion code PDQ7, which then is used to compute reactivity. These codes give results for fresh fuel that are in good agreement with KENO IV-NITAWL Monte Carlo results, which themselves are in good agreement with continuous energy Monte Carlo calculations. These methodologies are in reasonable agreement with critical measurements for undepleted fuel.

  4. GNS spent fuel cask experience

    SciTech Connect

    Weh, R. )

    1993-05-01

    The Gesellschaft fuer Nuklear-Service mbH (GNS), which is owned by German utilities, is responsible for the management of spent fuel and nuclear waste on behalf of the German utilities operating nuclear power plants. This paper describes the spent reactor fuel and waste shipping and/or storage casks that GNS manufacturers for nuclear facilities in Germany, and worldwide. So far more than 30 different casks have been produced in quantities ranging from one to several hundred of each type. GNS participates in the German Support Program to assist the International Atomic Energy Agency (IAEA) in developing verification procedures for dry storage casks containing spent fuel. This activity is also summarized.

  5. Spent fuel data for waste storage programs

    SciTech Connect

    Greene, E M

    1980-09-01

    Data on LWR spent fuel were compiled for dissemination to participants in DOE-sponsored waste storage programs. Included are mechanical descriptions of the existing major types of LWR fuel assemblies, spent LWR fuel fission product inventories and decay heat data, and inventories of LWR spent fuel currently in storage, with projections of future quantities.

  6. Spent fuel integrity during transportation

    SciTech Connect

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

    1980-01-01

    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.

  7. Spent-fuel storage requirements

    NASA Astrophysics Data System (ADS)

    1982-06-01

    Spent fuel storage requirements, as projected through the year 2000 for U.S. LWRs, were calculated using information supplied by the utilities reflecting plant status as of December 31, 1981. Projections through the year 2000 combined fuel discharge projections of the utilities with the assumed discharges of typical reactors required to meet the nuclear capacity of 165 GWe projected by the Energy Information Administration for the year 2000. Three cases were developed and are summarized. A reference case, or maximum at-reactor capacity case, assumes that all reactor storage pools are increased to their maximum capacities as estimated by the utilities for spent fuel storage utilizing currently licensed technologies. The reference case assumes no transshipments between pools except as current licensed by the Nuclear Regulatory Commission. This case identifies an initial requirement for 13 MTU of additional storage in 1984, and a cumulative requirement for 14,490 MTU additional storage in the year 2000.

  8. Spent fuel receipt scenarios study

    SciTech Connect

    Ballou, L.B.; Montan, D.N.; Revelli, M.A.

    1990-09-01

    This study reports on the results of an assignment from the DOE Office of Civilian Radioactive Waste Management to evaluate of the effects of different scenarios for receipt of spent fuel on the potential performance of the waste packages in the proposed Yucca Mountain high-level waste repository. The initial evaluations were performed and an interim letter report was prepared during the fall of 1988. Subsequently, the scope of work was expanded and additional analyses were conducted in 1989. This report combines the results of the two phases of the activity. This study is a part of a broader effort to investigate the options available to the DOE and the nuclear utilities for selection of spent fuel for acceptance into the Federal Waste Management System for disposal. Each major element of the system has evaluated the effects of various options on its own operations, with the objective of providing the basis for performing system-wide trade-offs and determining an optimum acceptance scenario. Therefore, this study considers different scenarios for receipt of spent fuel by the repository only from the narrow perspective of their effect on the very-near-field temperatures in the repository following permanent closure. This report is organized into three main sections. The balance of this section is devoted to a statement of the study objective, a summary of the assumptions. The second section of the report contains a discussion of the major elements of the study. The third section summarizes the results of the study and draws some conclusions from them. The appendices include copies of the waste acceptance schedule and the existing and projected spent fuel inventory that were used in the study. 10 refs., 27 figs.

  9. Spent nuclear fuel sampling strategy

    SciTech Connect

    Bergmann, D.W.

    1995-02-08

    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.

  10. Transportation accident scenarios for commercial spent fuel

    SciTech Connect

    Wilmot, E L

    1981-02-01

    A spectrum of high severity, low probability, transportation accident scenarios involving commercial spent fuel is presented together with mechanisms, pathways and quantities of material that might be released from spent fuel to the environment. These scenarios are based on conclusions from a workshop, conducted in May 1980 to discuss transportation accident scenarios, in which a group of experts reviewed and critiqued available literature relating to spent fuel behavior and cask response in accidents.

  11. Fenton- and Persulfate-driven Regeneration of Contaminant-spent Granular Activated Carbon

    EPA Science Inventory

    Fenton- or persulfate-driven chemical oxidation regeneration of spent granular activated carbon (GAC) involves the combined, synergistic use of two treatment technologies: adsorption of organic chemicals onto GAC and chemical oxidation regeneration of the spent-GAC. Environmental...

  12. Regeneration of spent catalysts in oxy-combustion atmosphere

    SciTech Connect

    Ammendola, P.; Chirone, R.; Ruoppolo, G.; Russo, G.

    2010-04-15

    The feasibility of adopting an oxy-combustion stage to regenerate a spent catalyst proposed for methane thermo-catalytic decomposition has been investigated in a laboratory scale bubbling fluidized bed reactor operated at 800 C and different inlet oxygen concentrations. The efficiency of carbon oxy-combustion regeneration strategy has been evaluated on the basis of the efficiency of carbon removed from the catalyst and the performance of regenerated catalyst. The effect of multiple cycles of decomposition and regeneration steps has been also quantified. Experimental activity confirmed the possibility of producing a CO{sub 2} stream that can be finalized to a sequestration unit but also indicated the requirement of a good temperature control of catalytic particles. (author)

  13. Spent nuclear fuel reprocessing modeling

    SciTech Connect

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

    2013-07-01

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

  14. Spent Nuclear Fuel (SNF) Project Execution Plan

    SciTech Connect

    LEROY, P.G.

    2000-11-03

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

  15. Spent nuclear fuel disposal liability insurance

    SciTech Connect

    Martin, D.W.

    1984-01-01

    This thesis examines the social efficiency of nuclear power when the risks of accidental releases of spent fuel radionuclides from a spent fuel disposal facility are considered. The analysis consists of two major parts. First, a theoretical economic model of the use of nuclear power including the risks associated with releases of radionuclides from a disposal facility is developed. Second, the costs of nuclear power, including the risks associated with a radionuclide release, are empirically compared to the costs of fossil fuel-fired generation of electricity. Under the provisions of the Nuclear Waste Policy Act of 1982, the federally owned and operated spent nuclear fuel disposal facility is not required to maintain a reserve fund to cover damages from an accidental radionuclide release. Thus, the risks of a harmful radionuclide release are not included in the spent nuclear fuel disposal fee charged to the electric utilities. Since the electric utilities do not pay the full, social costs of spent fuel disposal, they use nuclear fuel in excess of the social optimum. An insurance mechanism is proposed to internalize the risks associated with spent fueled disposal. Under this proposal, the Federal government is required to insure the disposal facility against any liabilities arising from accidental releases of spent fuel radionuclides.

  16. Method for the regeneration of spent molten zinc chloride

    DOEpatents

    Zielke, Clyde W.; Rosenhoover, William A.

    1981-01-01

    In a process for regenerating spent molten zinc chloride which has been used in the hydrocracking of coal or ash-containing polynuclear aromatic hydrocarbonaceous materials derived therefrom and which contains zinc chloride, zinc oxide, zinc oxide complexes and ash-containing carbonaceous residue, by incinerating the spent molten zinc chloride to vaporize the zinc chloride for subsequent condensation to produce a purified molten zinc chloride: an improvement comprising the use of clay in the incineration zone to suppress the vaporization of metals other than zinc. Optionally water is used in conjunction with the clay to further suppress the vaporization of metals other than zinc.

  17. Rack for storing spent nuclear fuel elements

    DOEpatents

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

    1978-06-20

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

  18. Nuclear criticality safety studies applicable to spent fuel shipping cask designs and spent fuel storage

    SciTech Connect

    Tang, J.S.

    1980-11-01

    Criticality analyses of water-moderated and reflected arrays of LWR fresh and spent fuel assemblies were carried out in this study. The calculated results indicate that using the assumption of fresh fuel loading in spent fuel shipping cask design leads to assembly spacings which are about twice the spacings of spent fuel loadings. Some shipping cask walls of composite lead and water are more effective neutron reflectors than water of 30.48 cm (12 in).

  19. Spent fuel transportation in the United States: commercial spent fuel shipments through December 1984

    SciTech Connect

    Not Available

    1986-04-01

    This report has been prepared to provide updated transportation information on light water reactor (LWR) spent fuel in the United States. Historical data are presented on the quantities of spent fuel shipped from individual reactors on an annual basis and their shipping destinations. Specifically, a tabulation is provided for each present-fuel shipment that lists utility and plant of origin, destination and number of spent-fuel assemblies shipped. For all annual shipping campaigns between 1980 and 1984, the actual numbers of spent-fuel shipments are defined. The shipments are tabulated by year, and the mode of shipment and the casks utilized in shipment are included. The data consist of the current spent-fuel inventories at each of the operating reactors as of December 31, 1984. This report presents historical data on all commercial spent-fuel transportation shipments have occurred in the United States through December 31, 1984.

  20. Spent Nuclear Fuel Project dose management plan

    SciTech Connect

    Bergsman, K.H.

    1996-03-01

    This dose management plan facilitates meeting the dose management and ALARA requirements applicable to the design activities of the Spent Nuclear Fuel Project, and establishes consistency of information used by multiple subprojects in ALARA evaluations. The method for meeting the ALARA requirements applicable to facility designs involves two components. The first is each Spent Nuclear Fuel Project subproject incorporating ALARA principles, ALARA design optimizations, and ALARA design reviews throughout the design of facilities and equipment. The second component is the Spent Nuclear Fuel Project management providing overall dose management guidance to the subprojects and oversight of the subproject dose management efforts.

  1. Apparatus for shearing spent nuclear fuel assemblies

    DOEpatents

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

    1980-01-01

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

  2. Spent Fuel Background Report Volume I

    SciTech Connect

    Abbott, D.

    1994-03-01

    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

  3. Catalytic iron oxide for lime regeneration in carbonaceous fuel combustion

    DOEpatents

    Shen, Ming-Shing; Yang, Ralph T.

    1980-01-01

    Lime utilization for sulfurous oxides absorption in fluidized combustion of carbonaceous fuels is improved by impregnation of porous lime particulates with iron oxide. The impregnation is achieved by spraying an aqueous solution of mixed iron sulfate and sulfite on the limestone before transfer to the fluidized bed combustor, whereby the iron compounds react with the limestone substrate to form iron oxide at the limestone surface. It is found that iron oxide present in the spent limestone acts as a catalyst to regenerate the spent limestone in a reducing environment. With only small quantities of iron oxide the calcium can be recycled at a significantly increased rate.

  4. Method for shearing spent nuclear fuel assemblies

    DOEpatents

    Weil, Bradley S.; Watson, Clyde D.

    1977-01-01

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

  5. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Rector, D. R.; Cuta, J. M.; Enderlin, C. W.

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.

  6. 78 FR 3853 - Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an Independent Spent Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-17

    ... COMMISSION 10 CFR Parts 71 and 72 Retrievability, Cladding Integrity and Safe Handling of Spent Fuel at an Independent Spent Fuel Storage Installation and During Transportation AGENCY: Nuclear Regulatory Commission... transport of spent nuclear fuel are separate from requirements for storage of spent nuclear fuel....

  7. Reexamination of spent fuel shipment risk estimates

    SciTech Connect

    COOK,J.R.; SPRUNG,JEREMY L.

    2000-04-25

    The risks associated with the transport of spent nuclear fuel by truck and rail have been reexamined and compared to results published in NUREG-O170 and the Modal Study. The full reexamination considered transport of PWR and BWR spent fuel by truck and rail in four generic Type B spent fuel casks. Because they are typical, this paper presents results only for transport of PWR spent fuel in steel-lead steel casks. Cask and spent fuel response to collision impacts and fires were evaluated by performing three-dimensional finite element and one-dimensional heat transport calculations. Accident release fractions were developed by critical review of literature data. Accident severity fractions were developed from Modal Study truck and rail accident event trees, modified to reflect the frequency of occurrence of hard and soft rock wayside route surfaces as determined by analysis of geographic data. Incident-free population doses and the population dose risks associated with the accidents that might occur during transport were calculated using the RADTRAN 5 transportation risk code. The calculated incident-free doses were compared to those published in NUREG-O170. The calculated accident dose risks were compared to dose risks calculated using NUREG-0170 and Modal Study accident source terms. The comparisons demonstrated that both of these studies made a number of very conservative assumptions about spent fuel and cask response to accident conditions, which caused their estimates of accident source terms, accident frequencies, and accident consequences to also be very conservative. The results of this study and the previous studies demonstrate that the risks associated with the shipment of spent fuel by truck or rail are very small.

  8. Spent Nuclear Fuel Project Technical Databook

    SciTech Connect

    Reilly, M.A.

    1998-10-23

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

  9. Spent Nuclear Fuel Transport Reliability Study

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao

    2016-01-01

    This conference paper was orignated and shorten from the following publisehd PTS documents: 1. Jy-An Wang, Hao Jiang, and Hong Wang, Dynamic Deformation Simulation of Spent Nuclear Fuel Assembly and CIRFT Deformation Sensor Stability Investigation, ORNL/SPR-2015/662, November 2015. 2. Jy-An Wang, Hong Wang, Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, NUREG/CR-7198, ORNL/TM-2014/214, May 2015. 3. Jy-An Wang, Hong Wang, Hao Jiang, Yong Yan, Bruce Bevard, Spent Nuclear Fuel Vibration Integrity Study 16332, WM2016 Conference, March 6 10, 2016, Phoenix, Arizona.

  10. Code System for Spent Fuel Heating Analysis.

    Energy Science and Technology Software Center (ESTSC)

    1999-05-24

    Version 00 SFHA calculates steady-state fuel rod temperatures for hexagon and square-fuel bundles. The code is used to perform sensitivity studies and confirmatory analyses of results submitted by applicants for spent fuel storage licenses. All three modes of heat transfer are considered; radiation, convection, and conduction. Each is modeled separately. SFHA benchmark calculations were made with test data to validate the use of a simple one-dimensional heat transfer model for estimating fuel rod temperatures. Benchmarkmore » results show that SFHA is capable of calculating spent fuel rod temperatures for square and hexagonal fuel bundles under various environments for the consolidated or unconsolidated condition. The program is menu-driven and executes automatically after all required information is entered.« less

  11. Corrosion of spent Advanced Test Reactor fuel

    SciTech Connect

    Lundberg, L.B.; Croson, M.L.

    1994-11-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented.

  12. Geomechanics of the Spent Fuel Test: Climax

    SciTech Connect

    Wilder, D.G.; Yow, J.L. Jr.

    1987-07-01

    Three years of geomechanical measurements were made at the Spent Fuel Test-Climax (SFT-C) 1400 feet underground in fractured granitic rock. Heating of the rock mass resulted from emplacement of spent fuel as well as the heating by electrical heaters. Cooldown of the rock occurred after the spent fuel was removed and the heaters were turned off. The measurements program examines both gross and localized responses of the rock mass to thermal loading, to evaluate the thermomechanical response of sheared and fractured rock with that of relatively unfractured rock, to compare the magnitudes of displacements during mining with those induced by extensive heating of the rock mass, and to check assumptions regarding symmetry and damaged zones made in numerical modeling of the SFT-C. 28 refs., 113 figs., 10 tabs.

  13. Laser Surveillance System for Spent Fuel

    SciTech Connect

    Fiarman, S.; Zucker, M.S.; Bieber, A.M. Jr.

    1980-01-01

    A laser surveillance system installed at spent fuel storage pools (SFSP's) will provide the safeguard inspector with specific knowledge of spent fuel movement that cannot be obtained with current surveillance systems. The laser system will allow for the division of the pool's spent fuel inventory into two populations - those assemblies which have been moved and those which haven't - which is essential for maximizing the efficiency and effectiveness of the inspection effort. We have designed, constructed, and tested a full size laser system operating in air and have used an array of 6 zircaloy BWR tubes to simulate an assembly. The reflective signal from the zircaloy rods is a strong function of position of the assembly, but in all cases is easily discernable from the reference scan of the background with no assembly. A design for a SFSP laser surveillance system incorporating laser ranging is discussed. 10 figures.

  14. Spent Nuclear Fuel Alternative Technology Decision Analysis

    SciTech Connect

    Shedrow, C.B.

    1999-11-29

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

  15. SEU43 fuel bundle shielding analysis during spent fuel transport

    SciTech Connect

    Margeanu, C. A.; Ilie, P.; Olteanu, G.

    2006-07-01

    The basic task accomplished by the shielding calculations in a nuclear safety analysis consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper investigates the effects induced by fuel bundle geometry modifications on the CANDU SEU spent fuel shielding analysis during transport. For this study, different CANDU-SEU43 fuel bundle projects, developed in INR Pitesti, have been considered. The spent fuel characteristics will be obtained by means of ORIGEN-S code. In order to estimate the corresponding radiation doses for different measuring points the Monte Carlo MORSE-SGC code will be used. Both codes are included in ORNL's SCALE 5 programs package. A comparison between the considered SEU43 fuel bundle projects will be also provided, with CANDU standard fuel bundle taken as reference. (authors)

  16. Spent fuel container alignment device and method

    DOEpatents

    Jones, Stewart D.; Chapek, George V.

    1996-01-01

    An alignment device is used with a spent fuel shipping container including a plurality of fuel pockets for spent fuel arranged in an annular array and having a rotatable cover including an access opening therein. The alignment device includes a lightweight plate which is installed over the access opening of the cover. A laser device is mounted on the plate so as to emit a laser beam through a laser admittance window in the cover into the container in the direction of a pre-established target associated with a particular fuel pocket. An indexing arrangement on the container provides an indication of the angular position of the rotatable cover when the laser beam produced by the laser is brought into alignment with the target of the associated fuel pocket.

  17. Hanford spent fuel inventory baseline

    SciTech Connect

    Bergsman, K.H.

    1994-07-15

    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.

  18. TRIGA spent-fuel storage criticality analysis

    SciTech Connect

    Ravnik, M.; Glumac, B.

    1996-06-01

    A criticality safety analysis of a pool-type storage for spent TRIGA Mark II reactor fuel is presented. Two independent computer codes are applied: the MCNP Monte Carlo code and the WIMS lattice cell code. Two types of fuel elements are considered: standard fuel elements with 12 wt% uranium concentration and FLIP fuel elements. A parametric study of spent-fuel storage lattice pitch, fuel element burnup, and water density is presented. Normal conditions and postulated accident conditions are analyzed. A strong dependence of the multiplication factor on the distance between the fuel elements and on the effective water density is observed. A multiplication factor <1 may be expected for an infinite array of fuel rods at center-to-center distances >6.5 cm, regardless of the fuel element type and burnup. At shorter distances, the subcriticality can be ensured only by adding absorbers to the array of fuel rods even if the fuel rods were burned to {approximately}20% burnup. The results of both codes agree well for normal conditions. The results show that WIMS may be used as a complement to the Monte Carlo code in some parts of the criticality analysis.

  19. Thermal Hydraulic Analysis of Spent Fuel Casks

    Energy Science and Technology Software Center (ESTSC)

    1997-10-08

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codesmore » for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.« less

  20. Radioactivity of spent TRIGA fuel

    NASA Astrophysics Data System (ADS)

    Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-01

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  1. Radioactivity of spent TRIGA fuel

    SciTech Connect

    Usang, M. D. Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  2. Electrometallurgical treatment of oxide spent fuel.

    SciTech Connect

    Karell, E. J.

    1999-06-08

    The Department of Energy (DOE) inventory of spent nuclear fuel contains a wide variety of oxide fuel types that may be unsuitable for direct repository disposal in their current form. The molten-salt electrometallurgical treatment technique developed by Argonne National Laboratory (ANL) has the potential to simplify preparing and qualifying these fuels for disposal by converting them into three uniform product streams: uranium metal, a metal waste form, and a ceramic waste form. This paper describes the major steps in the electrometallurgical treatment process for oxide fuels and provides the results of recent experiments performed to develop and scale up the process.

  3. BR-100 spent fuel shipping cask development

    SciTech Connect

    McGuinn, E.J.; Childress, P.C.

    1990-01-01

    Continued public acceptance of commercial nuclear power is contingent to a large degree on the US Department of Energy (DOE) establishing an integrated waste management system for spent nuclear fuel. As part of the from-reactor transportation segment of this system, the B W Fuel Company (BWFC) is under contract to the DOE to develop a spent-fuel cask that is compatible with both rail and barge modes of transportation. Innovative design approaches were the keys to achieving a cask design that maximizes payload capacity and cask performance. The result is the BR-100, a 100-ton rail/barge cask with a capacity of 21 PWR or 52 BWR ten-year cooled, intact fuel assemblies. 3 figs.

  4. Spent nuclear fuel project product specification

    SciTech Connect

    Pajunen, A.L.

    1998-01-30

    Product specifications are limits and controls established for each significant parameter that potentially affects safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for transport to dry storage. The product specifications in this document cover the spent fuel packaged in MultiCanister Overpacks (MCOs) to be transported throughout the SNF Project. The SNF includes N Reactor fuel and single-pass reactor fuel. The FRS removes the SNF from the storage canisters, cleans it, and places it into baskets. The MCO loading system places the baskets into MCO/Cask assembly packages. These packages are then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the MCO cask packages are transferred to the Canister Storage Building (CSB), where the MCOs are removed from the casks, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The key criteria necessary to achieve these goals are documented in this specification.

  5. Spent nuclear fuel project integrated schedule plan

    SciTech Connect

    Squires, K.G.

    1995-03-06

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

  6. Temperature for Spent Fuel Dry Storage

    Energy Science and Technology Software Center (ESTSC)

    1992-07-13

    DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) calculates allowable initial temperatures for dry storage of light-water-reactor spent fuel and the cumulative damage fraction of Zircaloy cladding for specified initial storage temperature and stress and cooling histories. It is made available to ensure compliance with NUREG 10CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI). Although the program''s principal purpose is to calculate estimatesmore » of allowable temperature limits, estimates for creep strain, annealing fraction, and life fraction as a function of storage time are also provided. Equations for the temperature of spent fuel in inert and nitrogen gas storage are included explicitly in the code; in addition, an option is included for a user-specified cooling history in tabular form, and tables of the temperature and stress dependencies of creep-strain rate and creep-rupture time for Zircaloy at constant temperature and constant stress or constant ratio of stress/modulus can be created. DATING includes the GEAR package for the numerical solution of the rate equations and DPLOT for plotting the time-dependence of the calculated cumulative damage-fraction, creep strain, radiation damage recovery, and temperature decay.« less

  7. Temperature for Spent Fuel Dry Storage

    SciTech Connect

    1992-07-13

    DATING (Determining Allowable Temperatures in Inert and Nitrogen Gases) calculates allowable initial temperatures for dry storage of light-water-reactor spent fuel and the cumulative damage fraction of Zircaloy cladding for specified initial storage temperature and stress and cooling histories. It is made available to ensure compliance with NUREG 10CFR Part 72, Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI). Although the program''s principal purpose is to calculate estimates of allowable temperature limits, estimates for creep strain, annealing fraction, and life fraction as a function of storage time are also provided. Equations for the temperature of spent fuel in inert and nitrogen gas storage are included explicitly in the code; in addition, an option is included for a user-specified cooling history in tabular form, and tables of the temperature and stress dependencies of creep-strain rate and creep-rupture time for Zircaloy at constant temperature and constant stress or constant ratio of stress/modulus can be created. DATING includes the GEAR package for the numerical solution of the rate equations and DPLOT for plotting the time-dependence of the calculated cumulative damage-fraction, creep strain, radiation damage recovery, and temperature decay.

  8. Spent nuclear fuel project product specification

    SciTech Connect

    PAJUNEN, A.L.

    1999-02-25

    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.

  9. Spent Nuclear Fuel (SNF) Project Product Specification

    SciTech Connect

    PAJUNEN, A.L.

    2000-01-20

    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.

  10. Numerical Estimation of the Spent Fuel Ratio

    SciTech Connect

    Lindgren, Eric R.; Durbin, Samuel; Wilke, Jason; Margraf, J.; Dunn, T. A.

    2016-01-01

    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 (DUO 2 ), have been conducted in the interim to more definitively determine the source term from these postulated events. 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 surrogate. Previous attempts to define the SFR in the 1980's have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Because of the large uncertainty surrounding the SFR, estimates of releases from security-related events may be unnecessarily conservative. Credible 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 storage of spent nuclear fuel in dry cask systems. In the present work, the shock physics codes CTH and ALE3D were used to simulate spent nuclear fuel (SNF) and DUO 2 targets impacted by a high-velocity jet at an ambient temperature condition. These preliminary results are used to illustrate an approach to estimate the respirable release fraction for each type of material and ultimately, an estimate of the SFR. This page intentionally blank

  11. Stressmeter placement at spent fuel test in climax granite

    SciTech Connect

    Abey, A.E.; Washington, H.R.

    1980-05-20

    Vibrating wire stressmeters were installed in the Spent Fuel Facility at the Nevada Test Site. These stressmeters will measure the changes in in situ stress during the five-year spent fuel test. Before installation, laboratory tests were conducted to study reproducibility of placement and to develop a program hopefully to reduce corrosion of the stressmeters while in place at the Spent Fuel Facility. These laboratory tests are discussed along with the installation of the stressmeters at the Spent Fuel Facility.

  12. Report on interim storage of spent nuclear fuel

    SciTech Connect

    Not Available

    1993-04-01

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

  13. Historical overview of domestic spent fuel shipments

    SciTech Connect

    Pope, R.B.; Wankerl, M.W. ); Armstrong, S.; Hamberger, C., Schmid, S. )

    1991-01-01

    The purpose of this paper is to provide available historical data on most commercial and research reactor spent fuel shipments that have been completed in the United States between 1964 and 1989. This information includes data on the sources of spent fuel that has been shipped, the types of shipping casks used, the number of fuel assemblies that have been shipped, and the number of shipments that have been made. The data are updated periodically to keep abreast of changes. Information on shipments is provided for planning purposes; to support program decisions of the US Department of Energy's (DOE's) Office of Civilian Radioactive Waste Management (OCRWM); and to inform interested members of the public, federal, state, and local government, Indian tribes, and the transportation community. 5 refs., 7 figs., 2 tabs.

  14. 77 FR 75065 - Rescinding Spent Fuel Pool Exclusion Regulations

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-19

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 51 Rescinding Spent Fuel Pool Exclusion Regulations AGENCY... consideration of spent fuel pool storage impacts from license renewal environmental review. The petition was... for a waiver of the NRC's spent fuel pool exclusion regulations. The petitioner requested that, if...

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

    SciTech Connect

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

    2014-06-10

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

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

    DOEpatents

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

    2013-11-05

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

  17. Spent Fuel Working Group Report. Volume 1

    SciTech Connect

    O`Toole, T.

    1993-11-01

    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 pool analysis using TRACE code

    SciTech Connect

    Sanchez-Saez, F.; Carlos, S.; Villanueva, J. F.; Martorell, S.

    2012-07-01

    The storage requirements of Spent Fuel Pools have been analyzed with the purpose to increase their rack capacities. In the past, the thermal limits have been mainly evaluated with conservative codes developed for this purpose, although some works can be found in which a best estimate code is used. The use of best estimate codes is interesting as they provide more realistic calculations and they have the capability of analyzing a wide range of transients that could affect the Spent Fuel Pool. Two of the most representative thermal-hydraulic codes are RELAP-5 and TRAC. Nowadays, TRACE code is being developed to make use of the more favorable characteristics of RELAP-5 and TRAC codes. Among the components coded in TRACE that can be used to construct the model, it is interesting to use the VESSEL component, which has the capacity of reproducing three dimensional phenomena. In this work, a thermal-hydraulic model of the Maine Yankee spent fuel pool using the TRACE code is developed. Such model has been used to perform a licensing calculation and the results obtained have been compared with experimental measurements made at the pool, showing a good agreement between the calculations predicted by TRACE and the experimental data. (authors)

  19. Spent Nuclear Fuel Alternative Technology Risk Assessment

    SciTech Connect

    Perella, V.F.

    1999-11-29

    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.

  20. Regeneration and recovery of spent Claus alumina catalyst

    SciTech Connect

    George, Z.M.

    1982-09-01

    Alberta, Canada recovers about seven million tons of elemental sulfur each year as a by-product of sour natural gas processing. Hydrogen sulfide is separated from the sour gas and subject to a sub-stoichiometric combustion in a reaction furnace at around 1000 degrees C, where about 60% of the sulfur present in the H/sub 2/S is recovered as sulfur. The residue, which is basically a mixture of H/sub 2/S and SO/sub 2/ in the stoichiometric ration of 2:1 together with significant quantities of water vapor and nitrogen is passed through a series of Claus adiabatic catalytic converters containing activated alumina or bauxite at around 250 degrees whereby the Claus reaction takes place: 2H/sub 2/S + SO/sub 2/ in equilibrium 2H/sub 2/O + 3/x s/sub x/ where x refers to the sulfur species at equilibrium. Employing four catalytic convertors in series, equilibrium conversions of over 98% are possible. Since the H/sub 2/S contains small quantities of heavier hydrocarbons, these undergo cracking and polymerization leading to carbon deposits on the catalyst and hence significant decrease in Claus catalytic activitiy. Presented is the results of research at the Alberta Research Council to regenerate spent Claus alumina catalysts. The process involves removal of the water soluble sulfates followed by an oxidate burn off to remove carbon and sulfur deposits. (JMT)

  1. Storage assembly for spent nuclear fuel

    SciTech Connect

    Lapides, M.E.

    1982-04-27

    A technique for storing spent fuel rods from a nuclear reactor is disclosed herein. This technique utilizes a housing including a closed inner chamber for containing the fuel rods and a thermally conductive member located partially within the housing chamber and partially outside the housing for transferring heat generated by the fuel rods from the chamber to the ambient surroundings. Particulate material is located within the chamber and surrounds the fuel rods contained therein. This material is selected to serve as a heat transfer media between the contained cells and the heat transferring member and, at the same time, stand ready to fuse into a solid mass around the contained cells if the heat transferring member malfunctions or otherwise fails to transfer the generated heat out of the housing chamber in a predetermined way.

  2. Radionuclide release from research reactor spent fuel

    NASA Astrophysics Data System (ADS)

    Curtius, H.; Kaiser, G.; Müller, E.; Bosbach, D.

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO 2 fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in 235U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO 2-fuel (LWR fuel, enrichment in 235U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Jülich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl 2-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl x-Al and U 3Si 2-Al) was studied in 400 mL MgCl 2-rich salt brine in the presence of Fe 2+ under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH) 3(s) and Eu(OH) 3(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu(IV) oxyhydroxides species due to radiolytic effects cannot

  3. Spent Nuclear Fuel Vibration Integrity Study

    SciTech Connect

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

    2016-01-01

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

  4. Some factors to consider in handling and storing spent fuel

    SciTech Connect

    Bailey, W.J.

    1985-11-01

    This report includes information from various studies performed under the Wet Storage Task of the Behavior of Spent Fuel in Storage Project of the Commercial Spent Fuel Management (CSFM) Program at Pacific Northwest Laboratory. Wet storage experience has been summarized earlier in several other reports. This report summarizes pertinent items noted during FY 1985 concerning recent developments in the handling and storage of spent fuel and associated considerations. The subjects discussed include recent publications, findings, and developments associated with: (1) storage of water reactor spent fuel in water pools, (2) extended-burnup fuel, (3) fuel assembly reconstitution and reinsertion, (4) rod consolidation, (5) variations in the US Nuclear Regulatory Commission's definition of failed fuel, (6) detection of failed fuel rods, and (7) extended integrity of spent fuel. A list of pertinent publications is included.

  5. LAB STUDY ON REGENERATION OF SPENT DOWEX 21K 16-20 MESH ION EXCHANGE RESIN

    SciTech Connect

    DUNCAN, J.B.

    2007-01-24

    Currently the effort to remove chromate from groundwater in the 100K and 100H Areas uses DOWEX 21K 16-20. This report addresses the procedure and results of a laboratory study for regeneration of the spent resin by sodium hydroxide, sulfuric acid, or sodium sulfate to determine if onsite regeneration by the Effluent Treatment Facility is a feasible option.

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

    SciTech Connect

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

    1994-10-01

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

  7. Particle Size Effects on Fenton Regeneration of MTBE-spent Activated Carbon

    EPA Science Inventory

    Fenton-driven regeneration of spent granular activated carbon (GAC) is a developing technology that may reduce water treatment costs. In this study, the effect of GAC particle size on Fenton-driven oxidation of methyl tert-butyl ether (MTBE)-spent GAC was evaluated. The GAC was...

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

    SciTech Connect

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

    2003-02-25

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

  9. Systems impacts of spent fuel disassembly alternatives

    SciTech Connect

    Not Available

    1984-07-01

    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.

  10. Pyroprocess for processing spent nuclear fuel

    DOEpatents

    Miller, William E.; Tomczuk, Zygmunt

    2002-01-01

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

  11. Hanford spent nuclear fuel project update

    SciTech Connect

    Williams, N.H.

    1997-08-19

    Twenty one hundred metric tons of spent nuclear fuel (SNF) are currently stored in the Hanford Site K Basins near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported to the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building.

  12. Spent fuel management status perspectives in Korea

    SciTech Connect

    Park, H.S.; Lee, J.S.; Kim, B.T. )

    1992-01-01

    Concomitant with steadily increasing nuclear power program in Korea, a national radioactive waste management program has been in initial implementation stage for several years. In late 1990, however, a serious confrontation was witnessed at Anmyon area where residents expressed strong opposition against any possibility to consider that site as a potential candidate for waste disposal by the Authority. As far as spent fuel management is concerned, an interim storage policy was adopted by Korean Atomic Energy Commission. A decision to build a centralized wet storage facility was made followed by a conceptual design. Due to the incident at Anmyon site, the public has became more concerned about radioactive wastes management. Parallel efforts are being made to ameliorate public acceptance in regard to radioactive waste management and in particular to spent fuel management. There are substantial uncertainties, however, whether any site could be found given that precarious mood has been prevailing against radioactive wastes throughout the world. In the meantime waiting for successful siting, various research and development for future perspectives are in order. Of particular importance in such endeavor is to provide technological impetus for future perspectives as well as public acceptance through safety demonstrations of certain viable technology alternatives. The dry storage option, for instance, is acclaimed for intrinsic safety and lower cost as prospective alternative. Combined with rod consolidation, dry storage technologies which have not extensively applied in the past, could be considered as a technological basis for longer term management of spent fuel. Conscious of such global trend, some appropriate programs in preparation for such perspectives have been launched by KAERI.

  13. Fenton-driven regeneration of MTBE-spent granular activated carbon - Effects of particle size and Iron Amendment Procedures

    EPA Science Inventory

    Fenton-driven regeneration of spent granular activated carbon (GAC) is a technology being developed to regenerate organic contaminant-spent GAC. Here, the effect of GAC particle size (>2 mm to <0.35 mm) on Fenton-driven oxidation of methyl tert-butyl ether (MTBE)-spent GAC was ev...

  14. DOE SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    SciTech Connect

    F. Habashi

    1998-06-26

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

  15. Surrogate Spent Nuclear Fuel Vibration Integrity Investigation

    SciTech Connect

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

    2014-01-01

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

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

    SciTech Connect

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

    1982-01-01

    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.

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

    SciTech Connect

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

    1981-09-01

    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.

  18. Volume reduction of spent fuel elements for direct disposal

    SciTech Connect

    Wasserfuhr, I.C.

    1995-12-31

    The method of direct disposal of spent fuel elements provides the placing of fuel and non-fuel elements into the POLLUX final disposal casks. It is, however, necessary to disassemble the fuel elements into fuel rods and structural parts. While the fuel rods are condensed, the remaining structure is treated further with a 500-t skeleton press to minimize the volume.

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

    SciTech Connect

    Jardine, Leslie J.

    2012-07-01

    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

  20. HTGR Spent Fuel Treatment Program. HTGR Spent Fuel Treatment Development Program Plan

    SciTech Connect

    Not Available

    1984-12-01

    The spent fuel treatment (SFT) program plan addresses spent fuel volume reduction, packaging, storage, transportation, fuel recovery, and disposal to meet the needs of the HTGR Lead Plant and follow-on plants. In the near term, fuel refabrication will be addressed by following developments in fresh fuel fabrication and will be developed in the long term as decisions on the alternatives dictate. The formulation of this revised program plan considered the implications of the Nuclear Waste Policy Act of 1982 (NWPA) which, for the first time, established a definitive national policy for management and disposal of nuclear wastes. Although the primary intent of the program is to address technical issues, the divergence between commercial and government interests, which arises as a result of certain provisions of the NWPA, must be addressed in the economic assessment of technically feasible alternative paths in the management of spent HTGR fuel and waste. This new SFT program plan also incorporates a significant cooperative research and development program between the United States and the Federal Republic of Germany. The major objective of this international program is to reduce costs by avoiding duplicate efforts.

  1. The TMI regenerable solid oxide fuel cell

    NASA Technical Reports Server (NTRS)

    Cable, Thomas L.

    1995-01-01

    Energy storage and production in space requires rugged, reliable hardware which minimizes weight, volume, and maintenance while maximizing power output and usable energy storage. These systems generally consist of photovoltaic solar arrays which operate during sunlight cycles to provide system power and regenerate fuel (hydrogen) via water electrolysis; during dark cycles, hydrogen is converted by the fuel cell into system. The currently preferred configuration uses two separate systems (fuel cell and electrolyzer) in conjunction with photovoltaic cells. Fuel cell/electrolyzer system simplicity, reliability, and power-to-weight and power-to-volume ratios could be greatly improved if both power production (fuel cell) and power storage (electrolysis) functions can be integrated into a single unit. The Technology Management, Inc. (TMI), solid oxide fuel cell-based system offers the opportunity to both integrate fuel cell and electrolyzer functions into one unit and potentially simplify system requirements. Based an the TMI solid oxide fuel cell (SOPC) technology, the TMI integrated fuel cell/electrolyzer utilizes innovative gas storage and operational concepts and operates like a rechargeable 'hydrogen-oxygen battery'. Preliminary research has been completed on improved H2/H2O electrode (SOFC anode/electrolyzer cathode) materials for solid oxide, regenerative fuel cells. Improved H2/H2O electrode materials showed improved cell performance in both fuel cell and electrolysis modes in reversible cell tests. ln reversible fuel cell/electrolyzer mode, regenerative fuel cell efficiencies (ratio of power out (fuel cell mode) to power in (electrolyzer model)) improved from 50 percent (using conventional electrode materials) to over 80 percent. The new materials will allow the TMI SOFC system to operate as both the electrolyzer and fuel cell in a single unit. Preliminary system designs have also been developed which indicate the technical feasibility of using the TMI SOFC

  2. Buckling analysis of spent fuel basket

    SciTech Connect

    Lee, A.S.; Bumpas, S.E.

    1995-05-01

    The basket for a spent fuel shipping cask is subjected to compressive stresses that may cause global instability of the basket assemblies or local buckling of the individual members. Adopting the common buckling design practice in which the stability capacity of the entire structure is based on the performance of the individual members of the assemblies, the typical spent fuel basket, which is composed of plates and tubular structural members, can be idealized as an assemblage of columns, beam-columns and plates. This report presents the flexural buckling formulas for five load cases that are common in the basket buckling analysis: column under axial loads, column under axial and bending loads, plate under uniaxial loads, plate under biaxial loadings, and plate under biaxial loads and lateral pressure. The acceptance criteria from the ASME Boiler and Pressure Vessel Code are used to determine the adequacy of the basket components. Special acceptance criteria are proposed to address the unique material characteristics of austenitic stainless steel, a material which is frequently used in the basket assemblies.

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

    SciTech Connect

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

    1982-11-01

    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.

  4. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    SciTech Connect

    Bevard, Bruce Balkcom; Mertyurek, Ugur; Belles, Randy; Scaglione, John M.

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  5. Spent-fuel photon and neutron source spectra

    SciTech Connect

    Hermann, O.W.; Alexander, C.W.

    1983-01-01

    Computational activities at Oak Ridge National Laboratory have been performed to develop appropriate data and techniques for computing the photon and neutron source spectra of spent fuel. The methods reviewed here include both the determination of spent-fuel composition and the radiation source spectra associated with these isotopic inventories.

  6. An approach to meeting the spent fuel standard

    SciTech Connect

    Makhijani, A.

    1996-05-01

    The idea of the spent fuel standard is that there should be a high surface gamma radiation to prevent theft. For purposes of preventing theft, containers should be massive, and the plutonium should be difficult to extract. This report discusses issues associated with the spent fuel standard.

  7. Nevada commercial spent nuclear fuel transportation experience

    SciTech Connect

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed.

  8. Characterization plan for Hanford spent nuclear fuel

    SciTech Connect

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

    1994-12-01

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

  9. Spent nuclear fuel project technical databook

    SciTech Connect

    Reilly, M.A.

    1998-07-22

    The Spent Nuclear Fuel (SNF) project technical databook provides project-approved summary tables of selected parameters and derived physical quantities, with nominal design and safety basis values. It contains the parameters necessary for a complete documentation basis of the SNF Project technical and safety baseline. The databook is presented in two volumes. Volume 1 presents K Basins SNF related information. Volume 2 (not yet available) will present selected sludge and water information, as it relates to the sludge and water removal projects. The values, within this databook, shall be used as the foundation for analyses, modeling, assumptions, or other input to SNF project safety analyses or design. All analysis and modeling using a parameter available in this databook are required to use and cite the appropriate associated value, and document any changes to those values (i.e., analysis assumptions, equipment conditions, etc). Characterization and analysis efforts are ongoing to validate, or update these values.

  10. Case histories of West Valley spent fuel shipments: Final report

    SciTech Connect

    Not Available

    1987-01-01

    In 1983, NRC/FC initiated a study on institutional issues related to spent fuel shipments originating at the former spent fuel processing facility in West Valley, New York. FC staff viewed the shipment campaigns as a one-time opportunity to document the institutional issues that may arise with a substantial increase in spent fuel shipping activity. NRC subsequently contracted with the Aerospace Corporation for the West Valley Study. This report contains a detailed description of the events which took place prior to and during the spent fuel shipments. The report also contains a discussion of the shipment issues that arose, and presents general findings. Most of the institutional issues discussed in the report do not fall under NRC's transportation authority. The case histories provide a reference to agencies and other institutions that may be involved in future spent fuel shipping campaigns. 130 refs., 7 figs., 19 tabs.

  11. DEVELOPMENT OF ELECTROCHEMICAL REDUCTION TECHNOLOGY FOR SPENT OXIDE FUELS

    SciTech Connect

    Hur, Jin-Mok; Seo, Chung-Seok; Kim, Ik-Soo; Hong, Sun-Seok; Kang, Dae-Seung; Park, Seong-Won

    2003-02-27

    The Advanced Spent Fuel Conditioning Process (ACP) has been under development at Korea Atomic Energy Research Institute (KAERI) since 1997. The concept is to convert spent oxide fuel into metallic form and to remove high heat-load fission products such as Cs and Sr from the spent fuel. The heat power, volume, and radioactivity of spent fuel can decrease by a factor of a quarter via this process. For the realization of ACP, a concept of electrochemical reduction of spent oxide fuel in Li2O-LiCl molten salt was proposed and several cold tests using fresh uranium oxides have been carried out. In this new electrochemical reduction process, electrolysis of Li2O and reduction of uranium oxide are taking place simultaneously at the cathode part of electrolysis cell. The conversion of uranium oxide to uranium metal can reach more than 99% ensuring the feasibility of this process.

  12. Cerenkov glow observations from spent fuel

    SciTech Connect

    Skalyo, J. Jr.

    1987-07-01

    The observation of Cerenkov glow from a fuel assembly is an attractive method of detecting the presence of radioactive material. The simple, hand-held instrumentation is very easy to use and does not require penetration of the water in the spent fuel pool. An obstacle to routine use of the instrument arises in that the standard night vision devices have a broad band wavelength response which required the pool area to be darkened. Various techniques used to limit the bandwidth of the devices for use in viewing the Cerenkov glow in the presence of facility illumination have furthered implementation. A properly specified, commercially available instrument has been used to make narrow band observations at two power reactors without interference from the facility illumination. Problems of interpretation of the observations persist. The technique has no useful role to play in the verification of an assembly at the rod level. As an item, the assembly can be verified as containing radioactive material in many instances; however some ambiguous situations were encountered.

  13. International status of dry storage of spent fuels

    SciTech Connect

    Schneider, K.J.; Mitchell, S.J.; Johnson, A.B. Jr.

    1992-04-01

    Spent fuel from the world`s nuclear power reactors, or the high-level radioactive wastes from reprocessing of the spent fuels, are planned to be disposed of in national deep geological repositories in the respective countries of origin. The plans for most countries with nuclear power call for spent fuel or high-level waste disposal to start between 2010 and about 2050. Although storage in water pools is the primary method for management of spent nuclear fuels for the first few years after discharge from the reactor, dry storage has been implemented in several countries and is being considered in others. Dry storage is generally planned for an interim period (from 10 to as long as 100 years) until the spent fuel is disposed of or until a final decision is made on reprocessing. Dry storage is also being used to supplement wet storage capacity at some nuclear power stations. This paper summarizes the world-wide status of dry spent fuel storage and information on the expected long-term integrity of the dry-stored spent fuel based on experience, particularly for Zircaloy-clad fuels. The paper also addresses briefly the dry storage of solidified high-level radioactive wastes. This paper is based on work carried out for the US Department of Energy (DOE) by the Pacific Northwest Laboratory.

  14. International status of dry storage of spent fuels

    SciTech Connect

    Schneider, K.J.; Mitchell, S.J.; Johnson, A.B. Jr.

    1992-04-01

    Spent fuel from the world's nuclear power reactors, or the high-level radioactive wastes from reprocessing of the spent fuels, are planned to be disposed of in national deep geological repositories in the respective countries of origin. The plans for most countries with nuclear power call for spent fuel or high-level waste disposal to start between 2010 and about 2050. Although storage in water pools is the primary method for management of spent nuclear fuels for the first few years after discharge from the reactor, dry storage has been implemented in several countries and is being considered in others. Dry storage is generally planned for an interim period (from 10 to as long as 100 years) until the spent fuel is disposed of or until a final decision is made on reprocessing. Dry storage is also being used to supplement wet storage capacity at some nuclear power stations. This paper summarizes the world-wide status of dry spent fuel storage and information on the expected long-term integrity of the dry-stored spent fuel based on experience, particularly for Zircaloy-clad fuels. The paper also addresses briefly the dry storage of solidified high-level radioactive wastes. This paper is based on work carried out for the US Department of Energy (DOE) by the Pacific Northwest Laboratory.

  15. Regeneration of spent three-way catalysts with nano-structured platinum group metals by gas and acid treatments.

    PubMed

    Kim, Sang Chai; Nahm, Seung Won; Wang, Geun Shim; Seo, Seong Gyu; Lee, Jae Wook

    2008-10-01

    The influence of physicochemical treatments on the catalytic activity of the spent nano-structured three way catalysts (TWCs) was examined to evaluate the possibility of using spent TWCs for removing VOCs. Thermal gases and acid aqueous solutions were used to regenerate the spent nano-structured TWCs. The characterization of the spent catalyst and its modified forms was carried out by using XRD, TEM, ICP, and N2 adsorption-desorption isotherms. The catalytic activity tests revealed that the spent nano-structured TWCs have a great potential for removing toxic compounds. The activities of catalysts were also found to be highly dependent on the treatment conditions. The acid aqueous treatments were very useful for improving the catalytic activity because they removed various contaminants such as fuel additives, lubricant oil additives, and metallic compounds. However, the thermal gas treated TWCs were less active than the parent TWCs. Furthermore, the activities of the catalysts treated with acids were closely connected with the remaining Pt/Al ratios. PMID:19198464

  16. Shippingport Spent Fuel Canister System Description

    SciTech Connect

    JOHNSON, D.M.

    2000-03-27

    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.

  17. Equipment designs for the spent LWR fuel dry storage demonstration

    SciTech Connect

    Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

    1980-01-01

    In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations.

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

    SciTech Connect

    1996-02-01

    Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

  19. Thermal Cooling Limits of Sbotaged Spent Fuel Pools

    SciTech Connect

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

    2010-09-10

    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. Loss of spent fuel pool cooling PRA: Model and results

    SciTech Connect

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 {times} 10{sup {minus}5} and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 {times} 10{sup {minus}3}. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible.

  1. Effect of Helium Accumulation on the Spent Fuel Microstructure

    SciTech Connect

    Ferry, Cecile; Piron, Jean-Paul; Stout, Ray

    2007-07-01

    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)

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

    SciTech Connect

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

    1980-05-01

    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.

  3. Fenton-Driven Chemical Regeneration of MTBE-Spent Granular Activated Carbon -- A Pilot Study

    EPA Science Inventory

    MTBE-spent granular activated carbon (GAC) underwent 3 adsorption/oxidation cycles. Pilot-scale columns were intermittently placed on-line at a ground water pump and treat facility, saturated with MTBE, and regenerated with H2O2 under different chemical, physical, and operational...

  4. Spent fuel test project, Climax granitic stock, Nevada Test Site

    SciTech Connect

    Ramspott, L.D.

    1980-10-24

    The Spent Fuel Test-Climax (SFT-C) is a test of dry geologic storage of spent nuclear reactor fuel. The SFT-C is located at a depth of 420 m in the Climax granitic stock at the Nevada Test Site. Eleven canisters of spent commercial PWR fuel assemblies are to be stored for 3 to 5 years. Additional heat is supplied by electrical heaters, and more than 800 channels of technical information are being recorded. The measurements include rock temperature, rock displacement and stress, joint motion, and monitoring of the ventilation air volume, temperature, and dewpoint.

  5. Disposition of ORNL's Spent Nuclear Fuel

    SciTech Connect

    Turner, D. W.; DeMonia, B. C.; Horton, L. L.

    2002-02-26

    This paper describes the process of retrieving, repackaging, and preparing Oak Ridge spent nuclear fuel (SNF) for off-site disposition. The objective of the Oak Ridge SNF Project is to safely, reliably, and efficiently manage SNF that is stored on the Oak Ridge Reservation until it can be shipped off-site. The project required development of several unique processes and the design and fabrication of special equipment to enable the successful retrieval, transfer, and repackaging of Oak Ridge SNF. SNF was retrieved and transferred to a hot cell for repackaging. After retrieval of SNF packages, the storage positions were decontaminated and stainless steel liners were installed to resolve the vulnerability of water infiltration. Each repackaged SNF canister has been transferred from the hot cell back to dry storage until off-site shipments can be made. Three shipments of aluminum-clad SNF were made to the Savannah River Site (SRS), and five shipments of non-aluminum-clad SNF are planned to the Idaho National Engineering and Environmental Laboratory (INEEL). Through the integrated cooperation of several organizations including the U.S. Department of Energy (DOE), Bechtel Jacobs Company LLC (BJC), Oak Ridge National Laboratory (ORNL), and various subcontractors, preparations for the disposition of SNF in Oak Ridge have been performed in a safe and successful manner.

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

    SciTech Connect

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

    1980-08-01

    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.

  7. Spent Nuclear Fuel (SNF) Project Product Specification

    SciTech Connect

    PAJUNEN, A.L.

    2000-12-07

    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

  8. Spent nuclear fuel assembly inspection using neutron computed tomography

    NASA Astrophysics Data System (ADS)

    Pope, Chad Lee

    The research presented here focuses on spent nuclear fuel assembly inspection using neutron computed tomography. Experimental measurements involving neutron beam transmission through a spent nuclear fuel assembly serve as benchmark measurements for an MCNP simulation model. Comparison of measured results to simulation results shows good agreement. Generation of tomography images from MCNP tally results was accomplished using adapted versions of built in MATLAB algorithms. Multiple fuel assembly models were examined to provide a broad set of conclusions. Tomography images revealing assembly geometric information including the fuel element lattice structure and missing elements can be obtained using high energy neutrons. A projection difference technique was developed which reveals the substitution of unirradiated fuel elements for irradiated fuel elements, using high energy neutrons. More subtle material differences such as altering the burnup of individual elements can be identified with lower energy neutrons provided the scattered neutron contribution to the image is limited. The research results show that neutron computed tomography can be used to inspect spent nuclear fuel assemblies for the purpose of identifying anomalies such as missing elements or substituted elements. The ability to identify anomalies in spent fuel assemblies can be used to deter diversion of material by increasing the risk of early detection as well as improve reprocessing facility operations by confirming the spent fuel configuration is as expected or allowing segregation if anomalies are detected.

  9. Scientists warn of 'trillion-dollar' spent-fuel risk

    NASA Astrophysics Data System (ADS)

    Gwynne, Peter

    2016-07-01

    A study by two Princeton University physicists suggests that a major fire in the spent nuclear fuel stored on the sites of US nuclear reactors could “dwarf the horrific consequences of the Fukushima accident”.

  10. Spent fuel storage at Prairie Island: January 1995 status

    SciTech Connect

    Closs, J.; Kress, L.

    1995-12-31

    The disposal of spent nuclear fuel has been an issue for the US since the inception of the commercial nuclear power industry. In the past decade, it has become a critical factor in the continued operation of some nuclear power plants, including the two units at Prairie Island. As the struggles and litigation over storage alternatives wage on, spent fuel pools continue to fill and plants edge closer to premature shutdown. Due to the delays in the construction of a federal repository, many nuclear power plants have had to seek interim storage alternatives. In the case of Prairie Island, the safest and most feasible option is dry cask storage. This paper discusses the current status of the Independent Spent Fuel Storage Installation (ISFSI) Project at Prairie Island. It provides a historical background to the project, discusses the notable developments over the past year, and presents the projected plans of the Northern States Power Company (NSP) in regards to spent fuel storage.

  11. Method for storing spent nuclear fuel in repositories

    DOEpatents

    Schweitzer, D.G.; Sastre, C.; Winsche, W.

    A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

  12. Method for storing spent nuclear fuel in repositories

    DOEpatents

    Schweitzer, Donald G.; Sastre, Cesar; Winsche, Warren

    1981-01-01

    A method for storing radioactive spent fuel in repositories containing sulfur as the storage medium is disclosed. Sulfur is non-corrosive and not subject to radiation damage. Thus, storage periods of up to 100 years are possible.

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

    SciTech Connect

    Dana, W.P.

    1995-12-01

    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

  14. COBRA-SFS. Thermal Analysis Spent Fuel Storage

    SciTech Connect

    Rector, D.R.

    1986-11-01

    COBRA-SFS is used for steady-state and transient thermal hydraulic analysis of spent fuel storage systems as well as other heat transfer and fluid flow problems. It is designed to predict flow and temperature distributions under a wide range of flow conditions, including mixed and natural convection. Two auxiliary programs, RADX1 and RADGEN, generate blackbody view factors and calculate radiation exchange factors for unconsolidated spent fuel assemblies to be supplied as input to COBRA-SFS.

  15. Reactor-specific spent fuel discharge projections: 1986 to 2020

    SciTech Connect

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

    1987-03-01

    The creation of five reactor-specific 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) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended 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. 6 refs., 8 figs., 8 tabs.

  16. An underwater neutron coincidence counter for measurement of spent fuels

    SciTech Connect

    Staples, P.; Halbig, J.; Lestone, J.; Sprinkle, J.

    1999-07-01

    An underwater neutron coincident counter has been designed and constructed for the measurement of spent nuclear fuel--the spent-fuel coincident counter (SFCC). The SFCC is a medium-detection-efficiency design that incorporates an ionization chamber (IC) for gamma-ray dose evaluation from the spent nuclear fuel. The absolute neutron detection efficiency is 14.5% for {sup 252}Cf sources. The SFCC is hermetically sealed, as it is installed {approximately}5 m below water level in a spent-fuel storage pond. There are 20 {sup 3}He tubes arranged within a polyethylene ring in a single band. There is an inner ring of 6.8 cm of lead to provide shielding from the fission product gamma rays. A single IC is primarily used to determine the dose impinging upon the {sup 3}He tubes and to determine the appropriate operational parameters to avoid gamma-ray pile effects in the {sup 3}He tubes. To further minimize gamma-ray pileup effects, each {sup 3}He tube is connected to a PDT110A preamplifier. The single and double neutron count rates, in addition to the IC measurement information from the SFCC, are used to determine the Pu mass of the spent-fuel assemblies and the decay heat and for classification of the assembly type. This information is required such that safety criteria are met for the safe packaging of the spent-fuel assemblies.

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

    SciTech Connect

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

    1988-03-01

    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.

  18. Microbiology of spent nuclear fuel storage basins.

    PubMed

    Santo Domingo, J W; Berry, C J; Summer, M; Fliermans, C B

    1998-12-01

    Microbiological studies of spent nuclear fuel storage basins at Savannah River Site (SRS) were performed as a preliminary step to elucidate the potential for microbial-influenced corrosion (MIC) in these facilities. Total direct counts and culturable counts performed during a 2-year period indicated microbial densities of 10(4) to 10(7) cells/ml in water samples and on submerged metal coupons collected from these basins. Bacterial communities present in the basin transformed between 15% and 89% of the compounds present in Biologtrade mark plates. Additionally, the presence of several biocorrosion-relevant microbial groups (i.e., sulfate-reducing bacteria and acid-producing bacteria) was detected with commercially available test kits. Scanning electron microscopy and X-ray spectra analysis of osmium tetroxide-stained coupons demonstrated the development of microbial biofilm communities on some metal coupons submerged for 3 weeks in storage basins. After 12 months, coupons were fully covered by biofilms, with some deterioration of the coupon surface evident at the microscopical level. These results suggest that, despite the oligotrophic and radiological environment of the SRS storage basins and the active water deionization treatments commonly applied to prevent electrochemical corrosion in these facilities, these conditions do not prevent microbial colonization and survival. Such microbial densities and wide diversity of carbon source utilization reflect the ability of the microbial populations to adapt to these environments. The presumptive presence of sulfate-reducing bacteria and acid-producing bacteria and the development of biofilms on submerged coupons indicated that an environment for MIC of metal components in the storage basins may occur. However, to date, there has been no indication or evidence of MIC in the basins. Basin chemistry control and corrosion surveillance programs instituted several years ago have substantially abated all corrosion mechanisms

  19. Neutron Generators for Spent Fuel Assay

    SciTech Connect

    Ludewigt, Bernhard A

    2010-12-30

    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.

  20. Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

    SciTech Connect

    McKinnon, M.A.; Creer, J.M.

    1988-10-01

    REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs.

  1. Safeguarding spent fuel storage and final disposal activities

    SciTech Connect

    Weh, R.; Wogatzki, E. )

    1991-01-01

    In Germany, the Atomic Energy Act provides for the spent fuel generated by nuclear power reactors to be reprocessed, if this is technically safe and economically viable. Thus the major share of used fuel from the German reactors is brought to reprocessing. The fuel recovered in this process is intended to be recycled into suitable reactors. The actual reprocessing is carried out abroad, in preference to a domestic solution, and the residues returned to Germany. This paper describes safeguarding measures for spent fuel storage and final disposal activities that are employed in Germany.

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

    SciTech Connect

    Botsch, W.; Smalian, S.; Hinterding, P.

    2013-07-01

    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

  3. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ...-only status and has spent fuel onsite, and for each independent spent fuel storage 10 CFR part 72... possession-only status that has spent fuel onsite, and to each independent spent fuel storage 10 CFR part 72... storage licenses. 171.15 Section 171.15 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ANNUAL FEES...

  4. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ...-only status and has spent fuel onsite, and for each independent spent fuel storage 10 CFR part 72... that has spent fuel onsite, and to each independent spent fuel storage 10 CFR part 72 licensee who does... storage licenses. 171.15 Section 171.15 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ANNUAL FEES...

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

    SciTech Connect

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

    1981-09-01

    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.

  6. Spent nuclear fuel discharges from US reactors 1993

    SciTech Connect

    Not Available

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  7. Historical overview of domestic spent fuel shipments: Update

    SciTech Connect

    Not Available

    1991-07-01

    This report presents available historic data on most commercial and research reactor spent fuel shipments in the United States from 1964 through 1989. Data include sources of the spent fuel shipped, types of shipping casks used, number of fuel assemblies shipped, and number of shipments made. This report also addresses the shipment of spent research reactor fuel. These shipments have not been documented as well as commercial power reactor spent fuel shipment activity. Available data indicate that the greatest number of research reactor fuel shipments occurred in 1986. The largest campaigns in 1986 were from the Brookhaven National Laboratory, Brooklyn, New York, to the Idaho Chemical Processing Plant (ICPP) and from the Oak Ridge National Laboratory's High Flux Isotope Reactor (HFIR) in Tennessee and the Rockwell International Reactor in California to the Savannah River Plant near Aiken, South Carolina. For all years addressed in this report, DOE facilities in Idaho Falls and Savannah River were the major recipients of research reactor spent fuel. In 1989, 10 shipments were received at the Idaho facilities. These originated from universities in California, Michigan, and Missouri. 9 refs., 12 figs., 7 tabs.

  8. COBRA-SFS. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Michener, T.E.; Rector, D.R.; Cuta, J.M.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form the latest release of the code, Cycle 2.

  9. Foreign experience in extended dry storage of spent nuclear fuel

    SciTech Connect

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

    1992-06-01

    Most countries with nuclear power are planning for spent nuclear fuel (or high-level waste from reprocessing of spent fuel) to be disposed of in national deep geological repositories starting in the time period of about 2010 to 2050. While spent fuel has been stored in water basins for the early years after discharge from the reactors, interim dry storage for extended periods (i.e., several tens of years) is being implemented or considered in an increasing number of countries. Dry storage technology is generally considered to be developed on a world-wide basis, and is being initiated and/ or expanded in a number of countries. This paper presents a summary of status and experience in dry storage of spent fuel in other countries, with emphasis on zirconium-clad fuels. Past activities, current status, future plans, research and development, and experience in dry storage are summarized for Argentina, Canada, France, former West Germany, former East Germany, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former Soviet Union. Conclusions from their experience are presented. Their experience to date supports the expectations that proper dry storage should provide for safe extended dry storage of spent fuel.

  10. Mobile Melt-Dilute Treatment for Russian Spent Nuclear Fuel

    SciTech Connect

    Peacock, H.

    2002-09-17

    Treatment of spent Russian fuel using a Melt-Dilute (MD) process is proposed to consolidate fuel assemblies into a form that is proliferation resistant and provides critically safety under storage and disposal configurations. Russian fuel elements contain a variety of fuel meat and cladding materials. The Melt-Dilute treatment process was initially developed for aluminum-based fuels so additional development is needed for several cladding and fuel meat combinations in the Russian fuel inventory (e.g. zirconium-clad, uranium-zirconium alloy fuel). A Mobile Melt-Dilute facility (MMD) is being proposed for treatment of spent fuels at reactor site storage locations in Russia; thereby, avoiding the costs of building separate treatment facilities at each site and avoiding shipment of enriched fuel assemblies over the road. The MMD facility concept is based on laboratory tests conducted at the Savannah River Technology Center (SRTC), and modular pilot-scale facilities constructed at the Savannah River Site for treatment of US spent fuel. SRTC laboratory tests have shown the feasibility of operating a Melt-Dilute treatment process with either a closed system or a filtered off-gas system. The proposed Mobile Melt-Dilute process is presented in this paper.

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

    SciTech Connect

    JOHNSON, D.M.

    2000-01-27

    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.

  12. 77 FR 76952 - Rescinding Spent Fuel Pool Exclusion Regulations

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-31

    ... receipt that appeared in the Federal Register (77 FR 75065; December 19, 2012). In particular, the NRC is...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 51 Rescinding Spent Fuel Pool Exclusion Regulations AGENCY... fuel pool storage impacts from license renewal environmental reviews. This action is necessary...

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

    DOEpatents

    Bucholz, James A.

    1983-01-01

    A separator assembly for use in a spent nuclear fuel shipping cask has a honeycomb-type wall structure defining parallel cavities for holding nuclear fuel assemblies. Tubes formed of an effective neutron-absorbing material are embedded in the wall structure around each of the cavities and provide neutron flux traps when filled with water.

  14. A metallic fuel cycle concept from spent oxide fuel to metallic fuel

    SciTech Connect

    Fujita, Reiko; Kawashima, Masatoshi; Yamaoka, Mitsuaki; Arie, Kazuo; Koyama, Tadafumi

    2007-07-01

    A Metallic fuel cycle concept for Self-Consistent Nuclear Energy System (SCNES) has been proposed in a companion papers. The ultimate goal of the SCNES is to realize sustainable energy supply without endangering the environment and humans. For future transition period from LWR era to SCNES era, a new metallic fuel recycle concept from LWR spent fuel has been proposed in this paper. Combining the technology for electro-reduction of oxide fuels and zirconium recovery by electrorefining in molten salts in the nuclear recycling schemes, the amount of radioactive waste reduced in a proposed metallic fuel cycle concept. If the recovery ratio of zirconium metal from the spent zirconium waste is 95%, the cost estimation in zirconium recycle to the metallic fuel materials has been estimated to be less than 1/25. (authors)

  15. Remote inspection of the IFSF spent fuel storage rack

    SciTech Connect

    Uldrich, E.D.

    1996-05-01

    The Irradiated Fuel Storage Facility (IFSF) is a dry storage facility for spent nuclear fuels located at the Idaho Chemical Processing Plant; it was constructed in the 1970`s specifically for the Fort Saint Vrain spent reactor fuels. Currently, it is being used for various spent fuels. It was not known if IFSF would met current DOE seismic criteria, so re-analysis was started, with the rack being analyzed first. The rack was inspected to determine the as-built condition. LazrLyne and VideoRuler were used in lieu of using a tape measure with the camera. It was concluded that when a visual inspection shows widely varying weld sizes, the engineer has to use all resources available to determine the most probable specified weld sizes.

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

    SciTech Connect

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

    1991-12-01

    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.

  17. Modeling of Spent Fuel Oxidation at Low Temperature

    SciTech Connect

    Poulesquen, Arnaud; Ferry, Cecile; Desgranges, Lionel

    2007-07-01

    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)

  18. Microbial biofilm growth on irradiated, spent nuclear fuel cladding

    NASA Astrophysics Data System (ADS)

    Bruhn, D. F.; Frank, S. M.; Roberto, F. F.; Pinhero, P. J.; Johnson, S. G.

    2009-02-01

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 10 3 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.

  19. Thermal-hydraulic analysis of spent fuel storage systems

    SciTech Connect

    Rector, D.R.; Wheeler, C.L.; Lombardo, N.J.

    1987-01-01

    This paper describes the COBRA-SFS (Spent Fuel Storage) computer code, which is designed to predict flow and temperature distributions in spent nuclear fuel storage and transportation systems. The decay heat generated by spent fuel in a dry storage cask is removed through a combination of conduction, natural convection, and thermal radiation. One major advantage of COBRA-SFS is that fluid recirculation within the cask is computed directly by solving the mass and momentum conservation equations. In addition, thermal radiation heat transfer is modeled using detailed radiation exchange factors based on quarter-rod segments. The equations governing mass, momentum, and energy conservation for incompressible flows are presented, and the semi-implicit solution method is described. COBRA-SFS predictions are compared to temperature data from a spent fuel storage cask test and the effect of different fill media on the cladding temperature distribution is discussed. The effect of spent fuel consolidation on cask thermal performance is also investigated. 16 refs., 6 figs., 2 tabs.

  20. Integrated approach to trailer design for spent fuel casks

    SciTech Connect

    Osborne, D.M.; Burgoyne, R.M.; Grenier, R.M.; Meyer, R.J.

    1989-02-01

    General Atomics (GA) is developing the GA-4 and GA-9 spent fuel transportation systems. The scope of our contract includes spent fuel casks, legal weight trailers, and ancillary equipment. Recent structural failures of spent fuel trailers have focused attention on trailer design. As a major element of spent fuel transportation systems, the concerns address the adequacy of trailer performance requirements, structural design and analysis, and in-service inspection and maintenance procedures. In response to these concerns, GA has applied an integrated approach to the design of the GA-4 and GA-9 transportation systems. The objectives are to design reliable, high-integrity trailers and to demonstrate their performance by test. Once the design is complete, a prototype trailer will be fabricated and a performance test program conducted in accordance with a comprehensive test program. GA`s trailer test program will include both design and operations elements, and will be used to optimize the operations and maintenance plan. The results of this program will provide positive public and regulatory perception of trailer durability and will support the development of industry standards for both legal weight and overweight trailers for spent fuel applications. 2 figs.

  1. Status of spent-fuel shipping cask development

    SciTech Connect

    Hall, I.K.; Hinschberger, T.S.

    1989-01-01

    The purpose of the Cask Systems Development Program is to develop a variety of cask systems that can safely and economically transport commercial spent fuel and high-level waste from the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. This paper is limited to a discussion of the status of from-reactor spent-fuel cask development; future cask development plans include MRS-to-repository casks, specialty casks for nonstandard spent fuel and nonfuel materials, and defense high-level waste casks. Spent-fuel casks must be available in the late 1990s to support the U.S. Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM) shipments from utilities. DOE-Idaho, with the support of EG G Idaho, Inc., Sandia National Laboratories, and selected cask developing contractors, has been assigned the responsibility for developing a new generation of cask systems. Four categories of spent fuel casks were initially proposed: (1) legal weight truck (LWT) casks (2) overweight truck (OWT) casks (3) rail/barge (R/B) casks (4) dual purpose (DP) storage/transport casks. Casks are being designed for reduced occupational radiation exposure at the receiving facility by facilitating the use of remote handling equipment. Automation of remote handling systems may be used to reduce cask turnaround time. Reducing turnaround time promotes reduced radiation exposure to occupational workers and improves cask utilization efficiency.

  2. Integrated approach to trailer design for spent fuel casks

    SciTech Connect

    Osborne, D.M.; Burgoyne, R.M.; Grenier, R.M.; Meyer, R.J.

    1989-02-01

    General Atomics (GA) is developing the GA-4 and GA-9 spent fuel transportation systems. The scope of our contract includes spent fuel casks, legal weight trailers, and ancillary equipment. Recent structural failures of spent fuel trailers have focused attention on trailer design. As a major element of spent fuel transportation systems, the concerns address the adequacy of trailer performance requirements, structural design and analysis, and in-service inspection and maintenance procedures. In response to these concerns, GA has applied an integrated approach to the design of the GA-4 and GA-9 transportation systems. The objectives are to design reliable, high-integrity trailers and to demonstrate their performance by test. Once the design is complete, a prototype trailer will be fabricated and a performance test program conducted in accordance with a comprehensive test program. GA's trailer test program will include both design and operations elements, and will be used to optimize the operations and maintenance plan. The results of this program will provide positive public and regulatory perception of trailer durability and will support the development of industry standards for both legal weight and overweight trailers for spent fuel applications. 2 figs.

  3. Microbial Biofilm Growth on Irradiated, Spent Nuclear Fuel Cladding

    SciTech Connect

    S.M. Frank

    2009-02-01

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 103 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.

  4. Separation of actinides from spent nuclear fuel: A review.

    PubMed

    Veliscek-Carolan, Jessica

    2016-11-15

    This review summarises the methods currently available to extract radioactive actinide elements from solutions of spent nuclear fuel. This separation of actinides reduces the hazards associated with spent nuclear fuel, such as its radiotoxicity, volume and the amount of time required for its' radioactivity to return to naturally occurring levels. Separation of actinides from environmental water systems is also briefly discussed. The actinide elements typically found in spent nuclear fuel include uranium, plutonium and the minor actinides (americium, neptunium and curium). Separation methods for uranium and plutonium are reasonably well established. On the other hand separation of the minor actinides from lanthanide fission products also present in spent nuclear fuel is an ongoing challenge and an area of active research. Several separation methods for selective removal of these actinides from spent nuclear fuel will be described. These separation methods include solvent extraction, which is the most commonly used method for radiochemical separations, as well as the less developed but promising use of adsorption and ion-exchange materials. PMID:27427893

  5. Methodology for determining criteria for storing spent fuel in air

    SciTech Connect

    Reid, C.R.; Gilbert, E.R.

    1986-11-01

    Dry storage in an air atmosphere is a method being considered for spent light water reactor (LWR) fuel as an alternative to storage in an inert gas environment. However, methods to predict fuel integrity based on oxidation behavior of the fuel first must be evaluated. The linear cumulative damage method has been proposed as a technique for defining storage criteria. Analysis of limited nonconstant temperature data on nonirradiated fuel samples indicates that this approach yields conservative results for a strictly decreasing-temperature history. On the other hand, the description of damage accumulation in terms of remaining life concepts provides a more general framework for making predictions of failure. Accordingly, a methodology for adapting remaining life concepts to UO/sub 2/ oxidation has been developed at Pacific Northwest Laboratory. Both the linear cumulative damage and the remaining life methods were used to predict oxidation results for spent fuel in which the temperature was decreased with time to simulate the temperature history in a dry storage cask. The numerical input to the methods was based on oxidation data generated with nonirradiated UO/sub 2/ pellets. The calculated maximum allowable storage temperatures are strongly dependent on the temperature-time profile and emphasize the conservatism inherent in the linear cumulative damage model. Additional nonconstant temperature data for spent fuel are needed to both validate the proposed methods and to predict temperatures applicable to actual spent fuel storage.

  6. Hot startup experience with electrometallurgical treatment of spent nuclear fuel

    SciTech Connect

    Benedict, R.W.; Lineberry, M.J.; McFarlane, H.F.; Rigg, R.H.

    1997-10-01

    The treatment of spent metal fuel from the EBR-II fast reactor commenced in June of 1996 at the Fuel Conditioning Facility on the Argonne-West site in Idaho, USA. During the first year of hot operations, 20 fuel assemblies entered processing and 6 low enrichment uranium product ingots were produced. Results are presented for the various process steps with decontamination factors achieved and equipment operational history reported.

  7. Cosmic-ray imaging of spent fuel casks

    NASA Astrophysics Data System (ADS)

    Guardincerri, Elena; Durham, J. Matthew; Morris, Christopher; Poulson, Daniel; Plaud-Ramos, Kenie; Fabritius, Joseph; Bacon, Jeffrey; Winston, Philip; Chichester, David

    2015-10-01

    Muon radiography was used to image the inside of a partially loaded Westinghouse MC-10 dry cask containing spent nuclear fuel at Idaho National Laboratory. We present here the results of a 100 hours long measurement taken in May 2015 with two muon trackers placed outside the cask. The data clearly show the location of the missing fuel bundles and demonstrate the feasibility of using cosmic rays to monitor fuel casks against illicit diversion of their content.

  8. The burnup dependence of light water reactor spent fuel oxidation

    SciTech Connect

    Hanson, B.D.

    1998-07-01

    Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies

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

    SciTech Connect

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

    1991-12-01

    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. Accelerated Closure of the Spent Nuclear Fuel (SNF) project

    SciTech Connect

    RUTHERFORD, W.W.

    2001-02-01

    The K East and K West Basins, built in the early 1950s, have been used to store irradiated nuclear fuel from the Hanford N Reactor. This fuel, which is referred to as spent nuclear fuel (SNF), has been stored underwater since 1975 in KE Basin and since 1981 in KW Basin. There are 54,000 N Reactor fuel assemblies in 3,800 canisters in the K West Basin, and 51,000 fuel assemblies in 3,700 canisters in the K East Basin that total 2,100 metric tons of SNF.

  11. Information handbook on independent spent fuel storage installations

    SciTech Connect

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

    1996-12-01

    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.

  12. Mission Need Statement: Idaho Spent Fuel Facility Project

    SciTech Connect

    Barbara Beller

    2007-09-01

    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.

  13. Application of spent fuel treatment technology to plutonium immobilization

    SciTech Connect

    McPheeters, C.C; Ackerman, J.P.; Gay, E.C., Johnson, G.K.

    1996-05-01

    The purpose of the electrometallurgical treatment technology being developed at Argonne National Laboratory (ANL) is to convert certain spent nuclear fuels into waste forms that are suitable for disposal in a geological repository for nuclear waste. The spent fuels of interest are those that cannot be safely stored for a long time in their current condition, and those that cannot be qualified for repository disposal. This paper explores the possibility of applying this electrometallurgical treatment technology to immobilization of surplus fissile materials, primarily plutonium. Immobilization of surplus fissile materials by electrometallurgical treatment could be done in the same facilities, at the same time. and in the same equipment as the proposed treatment of the present inventory of spent nuclear fuel. The cost and schedule savings of this simultaneous treatment scheme would be significant.

  14. Status of Proposed Repository for Latin-American Spent Fuel

    SciTech Connect

    Ferrada, J.J.

    2004-10-04

    This report compiles preliminary information that supports the premise that a repository is needed in Latin America and analyzes the nuclear situation (mainly in Argentina and Brazil) in terms of nuclear capabilities, inventories, and regional spent-fuel repositories. The report is based on several sources and summarizes (1) the nuclear capabilities in Latin America and establishes the framework for the need of a permanent repository, (2) the International Atomic Energy Agency (IAEA) approach for a regional spent-fuel repository and describes the support that international institutions are lending to this issue, (3) the current situation in Argentina in order to analyze the Argentinean willingness to find a location for a deep geological repository, and (4) the issues involved in selecting a location for the repository and identifies a potential location. This report then draws conclusions based on an analysis of this information. The focus of this report is mainly on spent fuel and does not elaborate on other radiological waste sources.

  15. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program.

    SciTech Connect

    Gregson, Michael Warren; Mo, Tin; Sorenson, Ken Bryce; Loiseau, Olivier; Nolte, Oliver; Hibbs, Russell S.; Molecke, Martin Alan; Slater-Thompson, Nancy; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Tsai, Han-Chung; Billone, Michael C.; Lange, Florentin; Young, Francis I.

    2004-08-01

    The authors provide a detailed overview of an on-going, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high-energy-density device. The program participants in the United States plus Germany, France and the United Kingdom, part of the international Working Group for Sabotage Concerns of Transport and Storage Casks (WGSTSC) have strongly supported and coordinated this research program. Sandia National Laboratories has the lead role for conducting this research program; test program support is provided by both the US Department of Energy and the US Nuclear Regulatory Commission. The authors provide a summary of the overall, multiphase test design and a description of all explosive containment and aerosol collection test components used. They focus on the recently initiated tests on 'surrogate' spent fuel, unirradiated depleted uranium oxide and forthcoming actual spent fuel tests, and briefly summarize similar results from completed surrogate tests that used non-radioactive, sintered cerium oxide ceramic pellets in test rods.

  16. Spent fuel behavior under abnormal thermal transients during dry storage

    SciTech Connect

    Stahl, D.; Landow, M.P.; Burian, R.J.; Pasupathi, V.

    1986-01-01

    This study was performed to determine the effects of abnormally high temperatures on spent fuel behavior. Prior to testing, calculations using the CIRFI3 code were used to determine the steady-state fuel and cask component temperatures. The TRUMP code was used to determine transient heating rates under postulated abnormal events during which convection cooling of the cask surfaces was obstructed by a debris bed covering the cask. The peak rate of temperature rise during the first 6 h was calculated to be about 15/sup 0/C/h, followed by a rate of about 1/sup 0/C/h. A Turkey Point spent fuel rod segment was heated to approx. 800/sup 0/C. The segment deformed uniformly with an average strain of 17% at failure and a local strain of 60%. Pretest characterization of the spent fuel consisted of visual examination, profilometry, eddy-current examination, gamma scanning, fission gas collection, void volume measurement, fission gas analysis, hydrogen analysis of the cladding, burnup analysis, cladding metallography, and fuel ceramography. Post-test characterization showed that the failure was a pinhole cladding breach. The results of the tests showed that spent fuel temperatures in excess of 700/sup 0/C are required to produce a cladding breach in fuel rods pressurized to 500 psing (3.45 MPa) under postulated abnormal thermal transient cask conditions. The pinhole cladding breach that developed would be too small to compromise the confinement of spent fuel particles during an abnormal event or after normal cooling conditions are restored. This behavior is similar to that found in other slow ramp tests with irradiated and nonirradiated rod sections and nonirradiated whole rods under conditions that bracketed postulated abnormal heating rates. This similarity is attributed to annealing of the irradiation-strengthened Zircaloy cladding during heating. In both cases, the failure was a benign, ductile pinhole rupture.

  17. Systems for the Intermodal Routing of Spent Nuclear Fuel

    SciTech Connect

    Peterson, Steven K; Liu, Cheng

    2015-01-01

    The safe and secure movement of spent nuclear fuel from shutdown and active reactor facilities to intermediate or long term storage sites may, in some instances, require the use of several modes of transportation to accomplish the move. To that end, a fully operable multi-modal routing system is being developed within Oak Ridge National Laboratory s (ORNL) WebTRAGIS (Transportation Routing Analysis Geographic Information System). This study aims to provide an overview of multi-modal routing, the existing state of the TRAGIS networks, the source data needs, and the requirements for developing structural relationships between various modes to create a suitable system for modeling the transport of spent nuclear fuel via a multimodal network. Modern transportation systems are comprised of interconnected, yet separate, modal networks. Efficient transportation networks rely upon the smooth transfer of cargoes at junction points that serve as connectors between modes. A key logistical impediment to the shipment of spent nuclear fuel is the absence of identified or designated transfer locations between transport modes. Understanding the potential network impacts on intermodal transportation of spent nuclear fuel is vital for planning transportation routes from origin to destination. By identifying key locations where modes intersect, routing decisions can be made to prioritize cost savings, optimize transport times and minimize potential risks to the population and environment. In order to facilitate such a process, ORNL began the development of a base intermodal network and associated routing code. The network was developed using previous intermodal networks and information from publicly available data sources to construct a database of potential intermodal transfer locations with likely capability to handle spent nuclear fuel casks. The coding development focused on modifying the existing WebTRAGIS routing code to accommodate intermodal transfers and the selection of

  18. Estimating Source Terms for Diverse Spent Nuclear Fuel Types

    SciTech Connect

    Brett Carlsen; Layne Pincock

    2004-11-01

    The U.S. Department of Energy (DOE) National Spent Nuclear Fuel Program is responsible for developing a defensible methodology for determining the radionuclide inventory for the DOE spent nuclear fuel (SNF) to be dispositioned at the proposed Monitored Geologic Repository at the Yucca Mountain Site. SNF owned by DOE includes diverse fuels from various experimental, research, and production reactors. These fuels currently reside at several DOE sites, universities, and foreign research reactor sites. Safe storage, transportation, and ultimate disposal of these fuels will require radiological source terms as inputs to safety analyses that support design and licensing of the necessary equipment and facilities. This paper summarizes the methodology developed for estimating radionuclide inventories associated with DOE-owned SNF. The results will support development of design and administrative controls to manage radiological risks and may later be used to demonstrate conformance with repository acceptance criteria.

  19. Dissolution of Spent Nuclear Fuel in Carbonate-Peroxide Solution

    SciTech Connect

    Soderquist, Chuck Z.; Hanson, Brady D.

    2010-01-31

    This study shows that spent UO2 fuel can be completely dissolved in a carbonate-peroxide solution apparently without attacking the metallic Mo-Tc-Ru-Rh-Pd fission product phase. Samples of spent nuclear fuel were pulverized and sieved to a uniform size, then duplicate aliquots were weighed into beakers for analysis. One set was dissolved in near-boiling 10M nitric acid, and the other set was dissolved in a solution of ammonium carbonate and hydrogen peroxide at room temperature. All the resulting fuel solutions were then analyzed for Sr-90, Tc-99, Cs-137, plutonium, and Am-241. For all the samples, the concentrations of Cs-137, Sr-90, plutonium, and Am-241 were the same for both the nitric acid dissolution and the ammonium carbonate-hydrogen peroxide dissolution, but the technetium concentration of the ammonium carbonate-hydrogen peroxide fuel solution was only about 25% of the same fuels dissolved in hot nitric acid.

  20. Receiving Basin for Offsite Fuels and the Resin Regeneration Facility Safety Analysis Report, Executive Summary

    SciTech Connect

    Shedrow, C.B.

    1999-11-29

    The Safety Analysis Report documents the safety authorization basis for the Receiving Basin for Offsite Fuels (RBOF) and the Resin Regeneration Facility (RRF) at the Savannah River Site (SRS). The present mission of the RBOF and RRF is to continue in providing a facility for the safe receipt, storage, handling, and shipping of spent nuclear fuel assemblies from power and research reactors in the United States, fuel from SRS and other Department of Energy (DOE) reactors, and foreign research reactors fuel, in support of the nonproliferation policy. The RBOF and RRF provide the capability to handle, separate, and transfer wastes generated from nuclear fuel element storage. The DOE and Westinghouse Savannah River Company, the prime operating contractor, are committed to managing these activities in such a manner that the health and safety of the offsite general public, the site worker, the facility worker, and the environment are protected.

  1. K Basin spent nuclear fuel characterization

    SciTech Connect

    LAWRENCE, L.A.

    1999-02-10

    The results of the characterization efforts completed for the N Reactor fuel stored in the Hanford K Basins were Collected and summarized in this single referencable document. This summary provides a ''road map'' for what was done and the results obtained for the fuel characterization program initiated in 1994 and scheduled for completion in 1999 with the fuel oxidation rate measurement under moist inert atmospheres.

  2. Spent fuel management at the Northern States Power Company

    SciTech Connect

    Closs, J.; Kress, L.

    1996-05-01

    The disposal of nuclear waste has become a critical factor in the continued operation of several commercial nuclear power plants in the US. As the debate continues, spent fuel pools continue to fill and plants edge closer to premature shutdown. This article describes the extensive spent fuel management program (ISFSI) of the Northern States Power Company (NSP). NSP owns and operates two nuclear power plants (Monticello and Prairie Island). Despite political opposition and legal restraint the program has successfully resolved storage limitations at its nuclear plants and has provided for the continuation of power at NSP. Projected plans for the future are also discussed.

  3. Automated shielding analysis sequences for spent fuel casks

    SciTech Connect

    Tang, J.S.; Parks, C.V.; Hermann, O.W.

    1987-01-01

    Two important Shielding Analysis Sequences (SAS) have recently been developed within the SCALE computational system. These sequences significantly enhance the existing SCALE system capabilities for evaluating radiation doses exterior to spent fuel casks. These new control module sequences (SAS1 and SAS4) and their capabilities are discussed and demonstrated, together with the existing SAS2 sequence that is used to generate radiation sources for spent fuel. Particular attention is given to the new SAS4 sequence which provides an automated scheme for generating and using biasing parameters in a subsequent Monte Carlo analysis of a cask.

  4. Spent fuel dissolution studies FY 1991 to 1994

    SciTech Connect

    Gray, W.J.; Wilson, C.N.

    1995-12-01

    Dissolution and transport as a result of groundwater flow are generally accepted as the primary mechanisms by which radionuclides from spent fuel placed in a geologic repository could be released to the biosphere. To help provide a source term for performance assessment calculations, dissolution studies on spent fuel and unirradiated uranium oxides have been conducted over the past few years at Pacific Northwest National Laboratory (PNNL) in support of the Yucca Mountain Site Characterization Project. This report describes work for fiscal years 1991 through 1994. The objectives of these studies and the associated conclusions, which were based on the limited number of tests conducted so far, are described in the following subsections.

  5. Department of Energy study on spent nuclear fuel storage

    SciTech Connect

    1980-03-01

    This report defines the needs for storage facilities and identifies possible sites in three regions of the US where such facilities could be located. The three sites are: Barnwell, South Carolina; Morris, Illinois; and West Valley, New York. This report includes consideration of the technical, economic, and regulatory factors associated with providing spent fuel storage in existing or potential at-reactor storage pools, and in AFR storage pools. This determination was based on specific data regarding the storage capacity needed to accommodate spent fuel from reactor pools by January 1, 1985, that the utilities would be unable to provide for themselves.

  6. Subcritical transmutation of spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Sommer, Christopher M.

    2011-07-01

    A series of fuel cycle simulations were performed using CEA's reactor physics code ERANOS 2.0 to analyze the transmutation performance of the Subcritical Advanced Burner Reactor (SABR). SABR is a fusion-fission hybrid reactor that combines the leading sodium cooled fast reactor technology with the leading tokamak plasma technology based on ITER physics. Two general fuel cycles were considered for the SABR system. The first fuel cycle is one in which all of the transuranics from light water reactors are burned in SABR. The second fuel cycle is a minor actinide burning fuel cycle in which all of the minor actinides and some of the plutonium produced in light water reactors are burned in SABR, with the excess plutonium being set aside for starting up fast reactors in the future. The minor actinide burning fuel cycle is being considered in European Scenario Studies. The fuel cycles were evaluated on the basis of TRU/MA transmutation rate, power profile, accumulated radiation damage, and decay heat to the repository. Each of the fuel cycles are compared against each other, and the minor actinide burning fuel cycles are compared against the EFIT transmutation system, and a low conversion ratio fast reactor.

  7. Reactor-specific spent fuel discharge projections, 1984 to 2020

    SciTech Connect

    Heeb, C.M.; Libby, R.A.; Holter, G.M.

    1985-04-01

    The original spent fuel utility data base (SFDB) has been adjusted to produce agreement with the EIA nuclear energy generation forecast. The procedure developed allows the detail of the utility data base to remain intact, while the overall nuclear generation is changed to match any uniform nuclear generation forecast. This procedure adjusts the weight of the reactor discharges as reported on the SFDB and makes a minimal (less than 10%) change in the original discharge exposures in order to preserve discharges of an integral number of fuel assemblies. The procedure used in developing the reactor-specific spent fuel discharge projections, as well as the resulting data bases themselves, are described in detail in this report. Discussions of the procedure cover the following topics: a description of the data base; data base adjustment procedures; addition of generic power reactors; and accuracy of the data base adjustments. Reactor-specific discharge and storage requirements are presented. Annual and cumulative discharge projections are provided. Annual and cumulative requirements for additional storage are shown for the maximum at-reactor (AR) storage assumption, and for the maximum AR with transshipment assumption. These compare directly to the storage requirements from the utility-supplied data, as reported in the Spent Fuel Storage Requirements Report. The results presented in this report include: the disaggregated spent fuel discharge projections; and disaggregated projections of requirements for additional spent fuel storage capacity prior to 1998. Descriptions of the methodology and the results are included in this report. Details supporting the discussions in the main body of the report, including descriptions of the capacity and fuel discharge projections, are included. 3 refs., 6 figs., 12 tabs.

  8. Anticipating Potential Waste Acceptance Criteria for Defense Spent Nuclear Fuel

    SciTech Connect

    Rechard, R.P.; Lord, M.E.; Stockman, C.T.; McCurley, R.D.

    1997-12-31

    The Office of Environmental Management of the U.S. Department of Energy is responsible for the safe management and disposal of DOE owned defense spent nuclear fuel and high level waste (DSNF/DHLW). A desirable option, direct disposal of the waste in the potential repository at Yucca Mountain, depends on the final waste acceptance criteria, which will be set by DOE`s Office of Civilian Radioactive Waste Management (OCRWM). However, evolving regulations make it difficult to determine what the final acceptance criteria will be. A method of anticipating waste acceptance criteria is to gain an understanding of the DOE owned waste types and their behavior in a disposal system through a performance assessment and contrast such behavior with characteristics of commercial spent fuel. Preliminary results from such an analysis indicate that releases of 99Tc and 237Np from commercial spent fuel exceed those of the DSNF/DHLW; thus, if commercial spent fuel can meet the waste acceptance criteria, then DSNF can also meet the criteria. In large part, these results are caused by the small percentage of total activity of the DSNF in the repository (1.5%) and regulatory mass (4%), and also because commercial fuel cladding was assumed to provide no protection.

  9. SPENT FUEL MANAGEMENT AT THE SAVANNAH RIVER SITE

    SciTech Connect

    Vormelker, P; Robert Sindelar, R; Richard Deible, R

    2007-11-03

    Spent nuclear fuels are received from reactor sites around the world and are being stored in the L-Basin at the Savannah River Site (SRS) in Aiken, South Carolina. The predominant fuel types are research reactor fuel with aluminum-alloy cladding and aluminum-based fuel. Other fuel materials include stainless steel and Zircaloy cladding with uranium oxide fuel. Chemistry control and corrosion surveillance programs have been established and upgraded since the early 1990's to minimize corrosion degradation of the aluminum cladding materials, so as to maintain fuel integrity and minimize personnel exposure from radioactivity in the basin water. Recent activities have been initiated to support additional decades of wet storage which include fuel inspection and corrosion testing to evaluate the effects of specific water impurity species on corrosion attack.

  10. Treatment of oxide spent fuel using the lithium reduction process

    SciTech Connect

    Karell, E.J.; Pierce, R.D.; Mulcahey, T.P.

    1996-05-01

    The wide variety in the composition of DOE spent nuclear fuel complicates its long-term disposition because of the potential requirement to individually qualify each type of fuel for repository disposal. Argonne National Laboratory (ANL) has developed the electrometallurgical treatment technique to convert all of these spent fuel types into a single set of disposal forms, simplifying the qualification process. While metallic fuels can be directly processed using the electrometallurgical treatment technique, oxide fuels must first be reduced to the metallic form. The lithium reduction process accomplishes this pretreatment. In the lithium process the oxide components of the fuel are reduced using lithium at 650 C in the presence of molten LiCl, yielding the corresponding metals and Li{sub 2}O. The reduced metal components are then separated from the LiCl salt phase and become the feed material for electrometallurgical treatment. A demonstration test of the lithium reduction process was successfully conducted using a 10-kg batch of simulated oxide spent fuel and engineering-scale equipment specifically constructed for that purpose. This paper describes the lithium process, the equipment used in the demonstration test, and the results of the demonstration test.

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

    SciTech Connect

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

    1980-05-01

    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.

  12. Nuclear reactor spent fuel storage rack

    SciTech Connect

    Machado, O.J.; Flynn, W.M.; Flanders, H.E. Jr.; Booker, L.W.

    1989-04-11

    A fuel rack is described for use in storing nuclear fuel assemblies in a nuclear fuel storage pool having a floor on which an upwardly projecting stud is mounted; the fuel rack comprising: a base structure at the lower end of the fuel rack including base-plate means having flow openings therein, the base-plate means supporting a first network of interlaced beams which form a multiplicity of polygonal openings; a second network of interlaced beams forming polygonal openings positioned in spaced vertical alignment with corresponding polygonal openings in the first network of beams; a plurality of cells, each cell having sides bounded by inner and outer surfaces and being of a size and configuration designed to hold therein a fuel assembly, each cell positioned in a corresponding pair of the aligned polygonal openings, each cell being open at both ends with a guiding funnel at the upper end, and the cells being positioned over the flow openings in the base-plate to permit flow of coolant through the cells; spaced, outwardly directed, projections on the outer surfaces of the sides of the cells near the tops and bottoms of the sides thereof, each cell being sized to be received within a corresponding of the pair of aligned polygonal openings in which the cells are respectively positioned; and means fixedly securing the projections to the beams in the first and second networks of beams thereby to provide a substantially rigid fuel rack of modular design.

  13. Electro-regeneration of Ce(IV) in real spent Cr-etching solutions.

    PubMed

    Chen, Te-San; Huang, Kuo-Lin

    2013-11-15

    This paper presents the electro-regeneration of Ce(IV) in real (hazardous) spent thin-film transistor liquid-crystal display (TFT-LCD) Cr-etching solutions. In addition to Ce(III)>Ce(IV) in diffusivity, a quasi-reversible behavior of Ce(III)/Ce(IV) was observed at both boron-doped diamond (BDD) and Pt disk electrodes. The Ce(IV) yield on Pt increased with increasing current density, and the best current efficiency (CE) was obtained at 2A/2.25 cm(2). The performance in terms of Ce(IV) yield and CE of tested anodes was in order BDD>Pt>dimensional stable anode (DSA). At 2A/2.25 cm(2) on Pt and 40 °C for 90 min, the Ce(IV) yield, CE and apparent rate constant (k) for Ce(III) oxidation were 81.4%, 21.8% and 3.17 × 10(-4) s(-1), respectively. With the increase of temperature, the Ce(IV) yield, CE, and k increased (activation energy = 10.7 kJ/mol), but the specific electricity consumption decreased. The Neosepta CMX membrane was more suitable than Nafion-117 and Nafion-212 to be used as the separator of the Ce(IV) regeneration process. The obtained parameters are useful to design divided batch reactors for the Ce(IV) electro-regeneration in real spent Cr-etching solutions. PMID:24140527

  14. K-Basin spent nuclear fuel characterization data report

    SciTech Connect

    Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; Thornton, T.A.

    1995-11-01

    The spent nuclear fuel (SNF) project characterization activities will be furnishing technical data on SNF stored at the K Basins in support of a pathway for placement of a ``stabilized`` form of SNF into an interim storage facility. This report summarizes the results so far of visual inspection of the fuel samples, physical characterization (e.g., weight and immersion density measurements), metallographic examinations, and controlled atmosphere furnace testing of three fuel samples shipped from the KW Basin to the Postirradiation Testing Laboratory (PTL). Data on sludge material collected by filtering the single fuel element canister (SFEC) water are also discussed in this report.

  15. Gamma Ray Mirrors for Direct Measurement of Spent Nuclear Fuel

    SciTech Connect

    Pivovaroff, Dr. Michael J.; Ziock, Klaus-Peter; Harrison, Mark J; Soufli, Regina

    2014-01-01

    Direct measurement of the amount of Pu and U in spent nuclear fuel represents a challenge for the safeguards community. Ideally, the characteristic gamma-ray emission lines from different isotopes provide an observable suitable for this task. However, these lines are generally lost in the fierce flux of radiation emitted by the fuel. The rates are so high that detector dead times limit measurements to only very small solid angles of the fuel. Only through the use of carefully designed view ports and long dwell times are such measurements possible. Recent advances in multilayer grazing-incidence gamma-ray optics provide one possible means of overcoming this difficulty. With a proper optical and coating design, such optics can serve as a notch filter, passing only narrow regions of the overall spectrum to a fully shielded detector that does not view the spent fuel directly. We report on the design of a mirror system and a number of experimental measurements.

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

    SciTech Connect

    Einziger, R.E.

    1988-12-01

    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.

  17. Spent fuel utilization in a compact traveling wave reactor

    SciTech Connect

    Hartanto, Donny; Kim, Yonghee

    2012-06-06

    In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

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

    SciTech Connect

    Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

    2010-10-01

    Romania successfully completed the world’s first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

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

    SciTech Connect

    CARLSON, A.B.

    1999-02-24

    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.

  20. Analysis of DOE Spent Nuclear Fuels for Repository Disposal

    SciTech Connect

    L.F. Pincock; W.D. Hintze; J. Duguid

    2006-02-07

    U.S. Department of Energy (DOE) spent nuclear fuel (SNF) consists of hundreds of different fuel types in various conditions. In order to analyze and model the DOE SNF for its suitability for repository disposal, several generalizations and simplifications were necessary. This paper describes the methodology used to arrive at a suitable DOE SNF surrogate and summarizes the proposed analysis of this DOE SNF surrogate for its appropriateness as a representative SNF.

  1. ADVANCED TECHNOLOGIES FOR THE SIMULTANEOUS SEPARATION OF CESIUM AND STRONTIUM FROM SPENT NUCLEAR FUEL

    SciTech Connect

    Jack D. Law; Terry A. Todd; R. Scott Herbst; David H. Meikrantz; Dean R. Peterman; Catherine L. Riddle; Richard D. Tillotson

    2005-02-01

    Two new solvent extraction technologies have been recently developed to simultaneously separate cesium and strontium from spent nuclear fuel, following dissolution in nitric acid. The first process utilizes a solvent consisting of chlorinated cobalt dicarbollide and polyethylene glycol extractants in a phenyltrifluoromethyl sulfone diluent. Recent improvements to the process include development of a new, non-nitroaromatic diluent and development of new stripping reagents, including a regenerable strip reagent that can be recovered and recycled. This new strip reagent reduces product volume by a factor of 20, over the baseline process. Countercurrent flowsheet tests on simulated spent nuclear fuel feed streams have been performed with both cesium and strontium removal efficiencies of greater than 99 %. The second process developed to simultaneously separate cesium and strontium from spent nuclear fuel is based on two highly-specific extractants: 4',4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. A solvent composition has been developed that enables both elements to be removed together and, in fact, a synergistic effect was observed with strontium distributions in the combined solvent that are much higher that in the strontium extraction (SREX) process. Initial laboratory test results of the new combined cesium and strontium extraction process indicate good extraction and stripping performance.

  2. Dry halide method for separating the components of spent nuclear fuels

    DOEpatents

    Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

    1998-06-30

    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.

  3. Dry halide method for separating the components of spent nuclear fuels

    DOEpatents

    Christian, Jerry Dale; Thomas, Thomas Russell; Kessinger, Glen F.

    1998-01-01

    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.

  4. Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel

    SciTech Connect

    J.A> Fortner; A.J. Kropf; R.J. Finch; J.C. Cunnane

    2006-06-20

    Spent uranium oxide nuclear fuel hosts a variety of trace chemical constituents, many of which must be sequestered from the biosphere during fuel storage and disposal. In this paper we present synchrotron x-ray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the plutonium and neptunium 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 U(VI)O{sub 2}{sup 2+} mineral assemblage that is depleted in plutonium and neptunium relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin ({approx}20 micrometer) layer that is enriched in plutonium and neptunium within a predominantly U{sup 4+} environment. Available data for the standard reduction potentials for NpO{sup 2+}/Np{sup 4+} and UO{sub 2}{sup 2+}/U{sup 4+} couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions. Neptunium is an important radionuclide in dose contribution according to performance assessment models of the proposed U. S. repository at Yucca Mountain, Nevada. A scientific understanding of how the UO{sub 2} matrix of spent nuclear fuel impacts the oxidative dissolution and reductive precipitation of neptunium is needed to predict its behavior 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 [1]. In the immediate vicinity of the spent fuel's surface the redox

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

    NASA Astrophysics Data System (ADS)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-01

    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.

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

    SciTech Connect

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-06

    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.

  7. Spent Fuel Test - Climax data acquisition system operations manual

    SciTech Connect

    Nyholm, R.A.

    1983-01-01

    The Spent Fuel Test-Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granite rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the US Department of Energy Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. The multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system (DAS) collects and processes data from more than 900 analog instruments. This report documents the software element of the LLNL developed SFT-C Data Acquisition System. It defines the operating system and hardware interface configurations, the special applications software and data structures, and support software.

  8. Spent fuel test. Climax data acquisition system integration report

    SciTech Connect

    Nyholm, R.A.; Brough, W.G.; Rector, N.L.

    1982-06-01

    The Spent Fuel Test - Climax (SFT-C) is a test of the retrievable, deep geologic storage of commercially generated, spent nuclear reactor fuel in granitic rock. Eleven spent fuel assemblies, together with 6 electrical simulators and 20 guard heaters, are emplaced 420 m below the surface in the Climax granite at the Nevada Test Site. On June 2, 1978, Lawrence Livermore National Laboratory (LLNL) secured funding for the SFT-C, and completed spent fuel emplacement May 28, 1980. This multi-year duration test is located in a remote area and is unattended much of the time. An extensive array of radiological safety and geotechnical instrumentation is deployed to monitor the test performance. A dual minicomputer-based data acquisition system collects and processes data from more than 900 analog instruments. This report documents the design and functions of the hardware and software elements of the Data Acquisition System and describes the supporting facilities which include environmental enclosures, heating/air-conditioning/humidity systems, power distribution systems, fire suppression systems, remote terminal stations, telephone/modem communications, and workshop areas. 9 figures.

  9. Pinhole Breaches in Spent Fuel Containers: Some Modeling Considerations

    SciTech Connect

    Casella, Andrew M.; Loyalka, Sudarsham K.; Hanson, Brady D.

    2006-06-04

    This paper replaces PNNL-SA-48024 and incorporates the ANS reviewer's comments, including the change in the title. Numerical methods to solve the equations for gas diffusion through very small breaches in spent fuel containers are presented and compared with previous literature results.

  10. Status of spent-fuel transportation system development

    SciTech Connect

    Chapman, R.L.; Hall, I.K.

    1988-01-01

    The purpose of the Cask Systems Development Program (CSDP) is to develop a variety of cask systems that can safely and economically be used to move commercial spent fuel and high-level waste from the generator to the federal repository or monitored retrievable storage facility. There are four initiatives to the CSDP, but only the first, from reactor casks, has been activated. This paper is limited to a discussion of the status of that initiative. Schedule objectives for the CSDP include development of spent-fuel cask systems by 1995 to support the Office of Civilian Radioactive Waste Management shipments of spent fuel from utilities beginning in the late 1990s. The US Department of Energy (DOE)-Idaho, with the support of EG G Idaho, Inc., Sandia National Laboratories, and selected cask development contractors, has been assigned the responsibility for developing a family of cask systems that are suitable for the task. Initially, four categories of spent-fuel casks were to be developed. They are legal-weight truck (LWT) casks, overweight truck (OWT) casks, rail/barge (R/B) casks, and dual purpose (DP) (storage/transport) casks. For a variety of reasons, OWT and DP cask development activities have been deferred. Program goals include developing a family of casks that will permit minimizing total system life cycle costs, ensure safety to the general public and to occupational workers, and attain public confidence in the transportation system.

  11. Quality assurance implementation plan for spent nuclear fuel characterization

    SciTech Connect

    Horhota, M.J.; Lawrence, L.A.

    1997-07-10

    A plan was prepared to implement the Quality Assurance requirements of the Office of Civilian Radioactive Waste Management RW-0333P to the Spent Nuclear Fuel Characterization activities. The plan was based on an evaluation of the current characterization activities against the RW-0333P requirements.

  12. Spent fuel storage: Progress with modular vault dry storage

    SciTech Connect

    Bower, C.C.F.

    1995-12-31

    This paper discusses the Modular Vault Dry Store (MVDS) for spent fuels at the Wylfa nuclear power plant in North Wales and at Fort St Vrain in Colorado. It goes on to discuss Scottish Nuclear`s decision not to proceed with MVDS facilities. It concludes by discussing Paks NPP contract with GEC Alsthom for the design and safety case for MDVS.

  13. Spent nuclear fuel project design basis capacity study

    SciTech Connect

    Cleveland, K.J.

    1996-09-09

    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.

  14. Review of Drying Methods for Spent Nuclear Fuel

    SciTech Connect

    Large, W.S.

    1999-10-21

    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.

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

    SciTech Connect

    Unterzuber, R.; Wright, J.B.

    1980-09-01

    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.

  16. Desulfurization sorbent regeneration

    DOEpatents

    Jalan, V.M.; Frost, D.G.

    1982-07-07

    A spent solid sorbent resulting from the removal of hydrogen sulfide from a fuel gas flow is regenerated with a steam-air mixture. The mixture of steam and air may also include additional nitrogen or carbon dioxide. The gas mixture contacts the spent sorbent containing metal sulfide at a temperature above 500/sup 0/C to regenerate the sulfide to metal oxide or carbonate. Various metal species including the period four transition metals and the lanthanides are suitable sorbents that may be regenerated by this method. In addition, the introduction of carbon dioxide gas permits carbonates such as those of strontium, barium and calcium to be regenerated. The steam permits regeneration of spent sorbent without formation of metal sulfate. Moreover, the regeneration will proceed with low oxygen concentrations and will occur without the increase in temperature to minimize the risk of sintering and densification of the sorbent. This method may be used for high-temperature fuel cells.

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

    SciTech Connect

    Fensin, Michael L; Tobin, Stephen J; Swinhoe, Martyn T; Menlove, Howard O; Sandoval, Nathan P

    2009-01-01

    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

  18. Recovery and regeneration of spent MHD seed material by the formate process

    DOEpatents

    Sheth, A.C.; Holt, J.K.; Rasnake, D.G.; Solomon, R.L.; Wilson, G.L.; Herrigel, H.R.

    1991-10-15

    The specification discloses a spent seed recovery and regeneration process for an MHD power plant employing an alkali metal salt seed material such as potassium salt wherein the spent potassium seed in the form of potassium sulfate is collected from the flue gas and reacted with calcium hydroxide and carbon monoxide in an aqueous solution to cause the formation of calcium sulfate and potassium formate. The pH of the solution is adjusted to suppress formation of formic acid and to promote precipitation of any dissolved calcium salts. The solution containing potassium formate is then employed to provide the potassium salt in the form of potassium formate or, optionally, by heating the potassium formate under oxidizing conditions to convert the potassium formate to potassium carbonate. 5 figures.

  19. Recovery and regeneration of spent MHD seed material by the formate process

    DOEpatents

    Sheth, Atul C.; Holt, Jeffrey K.; Rasnake, Darryll G.; Solomon, Robert L.; Wilson, Gregory L.; Herrigel, Howard R.

    1991-01-01

    The specification discloses a spent seed recovery and regeneration process for an MHM power plant employing an alkali metal salt seed material such as potassium salt wherein the spent potassium seed in the form of potassium sulfate is collected from the flue gas and reacted with calcium hydroxide and carbon monoxide in an aqueous solution to cause the formation of calcium sulfate and potassium formate. The pH of the solution is adjusted to supress formation of formic acid and to promote precipitation of any dissolved calcium salts. The solution containing potassium formate is then employed to provide the potassium salt in the form of potassium formate or, optionally, by heating the potassium formate under oxidizing conditions to convert the potassium formate to potassium carbonate.

  20. Removing Al and regenerating caustic soda from the spent washing liquor of Al etching

    NASA Astrophysics Data System (ADS)

    Barakat, M. A.; El-Sheikh, S. M.; Farghly, F. E.

    2005-08-01

    Spent liquor from washing of aluminum section materials after etching with caustic soda (NaOH) has been treated. Aluminum was removed from the liquor and caustic soda was regenerated by adding precipitating agents to hydrolyze sodium aluminate (Na2AlO2), separating the aluminumprecipitate, and concentrating free NaOH in the resulting solution for reuse in the etching process. Four systems were investigated: hydrated lime [Ca(OH)2], hydrogen peroxide (H2O2), H2O2/Ca(OH)2 mixture, and dry lime (CaO). Results revealed that CaO was more efficient in the removal of aluminum from the spent liquor with a higher hydrolyzing rate of Na2AlO2 than Ca(OH)2, H2O2, or their mixture.

  1. Near-term commercial spent fuel shipping cask requirements

    SciTech Connect

    Daling, P.M.

    1984-11-01

    This report describes an analysis of the near-term commercial light water reactor (LWR) spent fuel transportation system. The objective was to determine if the existing commercial spent fuel shipping cask fleet is adequate to provide the needed transportation services for the period of time the US government would be authorized to accept spent fuel for Federal Interim Storage (FIS). A spent fuel shipping cask supply-demand analysis was performed to evaluate the existing fleet size. The results of the shipping cask handling capability study indicated that by weight, 75% of the spent fuel shipments will be by truck (overweight plus legal-weight truck). From the results of the shipping cask supply-demand analysis it was concluded that, if utilities begin large-scale applications for FIS, the five legal-weight truck (LWT) casks currently in service would be inadequate to perform all of the needed shipments as early as 1987. This further assumes that a western site would be selected for the FIS facility. If the FIS site were to be located in the East, the need for additional LWT casks would be delayed by about two years. The overweight truck (OWT) cask fleet (two PWR and two BWR versions) will be adequate through 1992 if some shipments to FIS can be made several years before a reactor is projected to lose full core reserve. This is because OWT cask requirements increase gradually over the next several years. The feasibility of shipping before losing full core reserve has not been evaluated. Cask utilization requirements in later years will be reduced if some shipments can be made prior to the time they are actually needed. The existing three rail casks are adequate to perform near-term shipments. 18 references, 4 figures, 18 tables.

  2. Regulation of spent nuclear fuel shipment: A state perspective

    SciTech Connect

    Halstead, R.J.; Sinderbrand, C.; Woodbury, D.

    1987-01-01

    In 1985, the Wisconsin Department of Natural Resources (WDNR) sought to regulate rail shipments of spent nuclear fuel through the state, because federal regulations did not adequately protect the environmentally sensitive corridor along the route of the shipments. A state interagency working group identified five serious deficiencies in overall federal regulatory scheme: 1) failure to consider the safety or environmental risks associated with selected routes; 2) abscence of route-specific emergency response planning; 3) failure of the NRC to regulate the carrier of spent nuclear fuel or consider its safety record; 4) abscence of requirements for determination of need for, or the propriety of, specific shipments of spent nuclear fuel; and 5) the lack of any opportunity for meaningful public participation with respect to the decision to transport spent nuclear fuel. Pursuant to Wisconsin's hazardous substance statutes, the WDNR issues an order requiring the utility to file a spill prevention and mitigation plan or cease shipping through Wisconsin. A state trial court judge upheld the utility's challenge to Wisconsin's spill plan requirements, based on federal preemption of state authority. The state is now proposing federal legislation which would require: 1) NRC determination of need prior to approval of offsite shipment of spent fuel by the licensees; 2) NRC assessment of the potential environmental impacts of shipments along the proposed route, and comparative evaluation of alternative modes and routes; and 3) NRC approval of a route-specific emergency response and mitigation plan, including local training and periodic exercises. Additionally, the proposed legislation would authorize States and Indian Tribes to establish regulatory programs providing for permits, inspection, contingency plans for monitoring, containments, cleanup and decontamination, surveillance, enforcement and reasonable fees. 15 refs.

  3. Report on interim storage of spent nuclear fuel. Midwestern high-level radioactive waste transportation project

    SciTech Connect

    Not Available

    1993-04-01

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

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

    SciTech Connect

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

    1995-06-01

    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.

  5. Spent-Fuel Test-Climax: a progress report

    SciTech Connect

    Patrick, W.C.; Ballou, L.B.

    1982-09-20

    Both operational and technical objectives are being pursued at the Spent-Fuel Test-Climax (SFT-C). The principal operational objective is to demonstrate the safe and reliable packaging, handling, and storage of spent nuclear reactor fuel in a deep geologic media and to retrieve the fuel afterward. Packaging of the spent fuel at the Engine Maintenance, Assembly and Disassembly (EMAD) facility, initial emplacement 420m below surface in the Climax granitic stock, and three subsequent exchanges of fuel canisters between EMAD and the SFT-C has demonstrated that application of straightforward engineering practices provides a safe and highly reliable system with no significant radiation exposure to the operating personnel. The primary technical objectives of the test are simulation of the thermal effects occurring in a panel of a large repository and comparison of the relative effects on the granitic host rock of heat alone versus heat in combination with ionizing radiation. Other technical objectives direct project activities toward instrument evaluation, ventilation effects, thermal and thermomechanical response of a jointed rock mass, and computer model validation. Recent findings from field measurements and laboratory studies are briefly discussed for: performance of data acquisition system and instrumentation; near-and intermediate-field temperature measurements; ventilation and dewpoint measurements; acoustic emission monitoring of fractures in granites; radiation-dose-to-granite measurements.

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

    SciTech Connect

    Lee, S.Y.

    1999-01-13

    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.

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

    SciTech Connect

    1997-03-01

    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.

  8. Ventilation systems for a spent LWR fuel recycle complex

    SciTech Connect

    Not Available

    1981-01-01

    A conceptual design study has been made of a facility to recycle spent Light Water Reactor fuel. This study was based on coprocessing of plutonium and uranium where plutonium is never available as a separate material. The design of the fuel reprocessing facilities is based on remote operation and remote maintenance. The experience of many years of safe and dependable operation of government fuel processing facilities at Savannah River and Hanford was used in the design. A requirement of the study was that the facilities be licensable under Title 10 and Title 40 of the Code of Federal Regulations.

  9. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGESBeta

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Hu, Jianwei; Schwalbach, P.; Sjoland, A.; et al

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  10. 10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Conditions for spent fuel storage cask renewal. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks...

  11. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel...

  12. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... only status and has spent fuel onsite, and each independent spent fuel storage 10 CFR part 72 licensee... onsite, and to each independent spent fuel storage 10 CFR part 72 licensee who does not hold a 10 CFR... storage licenses. 171.15 Section 171.15 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ANNUAL FEES...

  13. 78 FR 8050 - Spent Fuel Cask Certificate of Compliance Format and Content

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-05

    ... COMMISSION 10 CFR Part 72 Spent Fuel Cask Certificate of Compliance Format and Content AGENCY: Nuclear... that governs the format and content of spent fuel storage cask Certificates of Compliance (CoCs... criteria for the format and content to be included in a spent fuel storage cask Certificate of...

  14. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Approval of Spent Fuel Storage Casks § 72.236 Specific requirements for... storage cask, maximum heat designed to be dissipated, maximum spent fuel loading limit, condition of...

  15. Feasibility of x ray fluorescence for spent fuel safeguards

    SciTech Connect

    Freeman, Corey Ross; Mozin, Vladimir; Tobin, Stephen J; Fensin, Michael L; White, Julia M; Croft, Stephen; Stafford, Alissa; Charlton, William

    2010-01-01

    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

  16. Assessment of spent-fuel waste-form/stabilizer alternatives for geologic disposal

    NASA Astrophysics Data System (ADS)

    Einziger, R. E.; Himes, D. A.

    1982-09-01

    The possibility of burying canisterized unreprocessed spent fuel in a deep geologic repository is studied. One aspect is an assessment of the possible spent fuel waste forms. The fuel performance portion was to evaluate five candidate spent fuel waste forms for postemplacement performance with emphasis on their ability to retard the release of radiouclides to the repository geology. Spent fuel waste forms under general consideration were: (1) unaltered fuel assembly; (2) fuel assembly with end fittings removed to shorten the length; (3) rods vented to remove gases and resealed; (4) disassembled fuel bundles to close pack the rods; and (5) rods chopped and fragments immobilized in a matrix material.

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

    SciTech Connect

    Einziger, R. E.

    1998-12-16

    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

  18. Spent nuclear fuel project operational description

    SciTech Connect

    Duncan, D.R., Westinghouse Hanford

    1996-07-01

    This Operational Description is prepared as input to the Multi- Canister Overpack (MCO) Pressurization Analysis. The MCO Pressurization analysis will include topical studies and analyses on fuel/moisture/has behavior areas, modeling of temperature and pressure, and a culminating integrated gas generation and pressurization analysis providing the expected MCO pressure history for normal operating scenarios and for off-normal events. A basis for the temperature and pressure modeling will be this operational description. Its objective is to provide time,temperature, and MCO material loading envelopes for the modeling efforts, both for normal and off-normal events.

  19. A review on methods of regeneration of spent pickling solutions from steel processing.

    PubMed

    Regel-Rosocka, Magdalena

    2010-05-15

    The review presents various techniques of regeneration of spent pickling solutions, including the methods with acid recovery, such as diffusion dialysis, electrodialysis, membrane electrolysis and membrane distillation, evaporation, precipitation and spray roasting as well as those with acid and metal recovery: ion exchange, retardation, crystallization solvent and membrane extraction. Advantages and disadvantages of the techniques are presented, discussed and confronted with the best available techniques requirements. Most of the methods presented meet the BAT requirements. The best available techniques are electrodialysis, diffusion dialysis and crystallization; however, in practice spray roasting and retardation/ion-exchange are applied most frequently for spent pickling solution regeneration. As "waiting for their chance" solvent extraction, non-dispersive solvent extraction and membrane distillation should be indicated because they are well investigated and developed. Environmental and economic benefits of the methods presented in the review depend on the cost of chemicals and wastewater treatment, legislative regulations and cost of modernization of existing technologies or implementation of new ones. PMID:20056321

  20. Radiochemical analyses of several spent fuel Approved Testing Materials

    SciTech Connect

    Guenther, R.J.; Blahnik, D.E.; Wildung, N.J.

    1994-09-01

    Radiochemical characterization data are described for UO{sub 2} and UO{sub 2} plus 3 wt% Gd{sub 2}O{sub 3} commercial spent nuclear fuel taken from a series of Approved Testing Materials (ATMs). These full-length nuclear fuel rods include MLA091 of ATM-103, MKP070 of ATM-104, NBD095 and NBD131 of ATM-106, and ADN0206 of ATM-108. ATMs 103, 104, and 106 were all irradiated in the Calvert Cliffs Nuclear Power Plant (Reactor No.1), a pressurized-water reactor that used fuel fabricated by Combustion Engineering. ATM-108 was part of the same fuel bundle designed as ATM-105 and came from boiling-water reactor fuel fabricated by General Electric and irradiated in the Cooper Nuclear Power Plant. Rod average burnups and expected fission gas releases ranged from 2,400 to 3,700 GJ/kgM. (25 to 40 Mwd/kgM) and from less than 1% to greater than 10%, respectively, depending on the specific ATM. The radiochemical analyses included uranium and plutonium isotopes in the fuel, selected fission products in the fuel, fuel burnup, cesium and iodine on the inner surfaces of the cladding, {sup 14}C in the fuel and cladding, and analyses of the gases released to the rod plenum. Supporting examinations such as fuel rod design and material descriptions, power histories, and gamma scans used for sectioning diagrams are also included. These ATMs were examined as part of the Materials Characterization Center Program conducted at Pacific Northwest Laboratory provide a source of well-characterized spent fuel for testing in support of the US Department of Energy Office of Civilian Radioactive Waste Management Program.

  1. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 1, Activation measurements and comparison with calculations for spent fuel assembly hardware

    SciTech Connect

    Luksic, A.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel. During consolidation, the fuel pins are removed from an intact fuel assembly and repackaged into a more compact configuration. After repackaging, approximately 30 kg of residual spent fuel assembly hardware per assembly remains that is also radioactive and requires disposal. Understanding the nature of this secondary waste stream is critical to designing a system that will properly handle, package, store, and dispose of the waste. This report presents a methodology for estimating the radionuclide inventory in irradiated spent fuel hardware. Ratios are developed that allow the use of ORIGEN2 computer code calculations to be applied to regions that are outside the fueled region. The ratios are based on the analysis of samples of irradiated hardware from spent fuel assemblies. The results of this research are presented in three volumes. In Volume 1, the development of scaling factors that can be used with ORIGEN2 calculations to estimate activation of spent fuel assembly hardware is documented. The results from laboratory analysis of irradiated spent-fuel hardware samples are also presented in Volume 1. In Volumes 2 and 3, the calculated flux profiles of spent nuclear fuel assemblies are presented for pressurized water reactors and boiling water reactors, respectively. The results presented in Volumes 2 and 3 were used to develop the scaling factors documented in Volume 1. 5 refs., 4 figs., 21 tabs.

  2. Safeguards techniques in a pilot conditioning plant for spent fuel

    SciTech Connect

    Leitner, E.; Rudolf, K.; Weh, R. )

    1991-01-01

    The pilot conditioning plant at Gorleben, Germany, is designed as a multi-purpose plant. Its primary task is the conditioning of spent fuel assemblies into a form suitable for final disposal. As a pilot plant, it allows furthermore for the development and testing of various conditioning techniques. In terms of international safeguards, the pilot conditioning plant is basically considered an item facility. Entire fuel assemblies enter the plant in transport casks, whereas bins filled with fuel rods or canisters containing cut fuel rods leave the facility in final disposal packages (e.g. POLLUX). Each POLLUX final disposal package content is uniquely correlated to a definite number of fuel assemblies which have entered the conditioning process. For this type of facility, containment/surveillance (C/S) should take over the major role in nuclear material safeguards. This paper discusses the safeguards at the Gorleben plant.

  3. Interim Storage of Hanford Spent Fuel & Associated Sludge

    SciTech Connect

    MAKENAS, B.J.

    2002-07-01

    The Hanford site is currently dealing with a number of types of Spent Nuclear Fuel. The route to interim dry storage for the various fuel types branches along two different paths. Fuel types such as metallic N reactor fuel and Shippingport Core 2 Blanket assemblies are being placed in approximately 4 m long canisters which are then stored in tubes below grade in a new canister storage building. Other fuels such as TRIGA{trademark} and Light Water Reactor fuel will be relocated and stored in stand-alone casks on a concrete pad. Varying degrees of sophistication are being applied with respect to the drying and/or evacuation of the fuel interim storage canisters depending on the reactivity of the fuel, the degree of damaged fuel and the previous storage environment. The characterization of sludge from the Hanford K Basins is nearly complete and canisters are being designed to store the sludge (including uranium particles from fuel element cleaning) on an interim basis.

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

    SciTech Connect

    Tobin, S. J.; Sandoval, N. P.; Fensin, M. L.; Lee, S. Y.; Ludewigt, Bernhard A.; Menlovea, H. O.; Quiter, B. J.; Rajasingume, A.; Schearf, M. A.; Smith, L. E.; Swinhoe, M. T.; Thompson, S. J.

    2009-06-30

    There are a variety of reasons for quantifying plutonium (Pu) in spent fuel such as independently verifying the Pu content declared by a regulated facility, making shipper/receiver mass declarations, and quantifying the input mass at a reprocessing facility. As part of the Next Generation Safeguards Initiative, NA-241 has recently funded a multilab/university collaboration to determine the elemental Pu mass in spent fuel assemblies. This research effort is anticipated to be a five year effort: the first part of which is a two years Monte Carlo modeling effort to integrate and down-select among 13 nondestructive assay (NDA) technologies, followed by one year for fabricating instruments and then two years for measuring spent fuel. This paper gives a brief overview of the approach being taken for the Monte Carlo research effort. In addition, preliminary results for the first NDA instrument studied in detail, delayed neutron detection, will be presented. In order to cost effectively and robustly model the performance of several NDA techniques, an"assembly library" was created that contains a diverse range of pressurized water reactor spent fuel assemblies (burnup, enrichment, cooling time) similar to that which exists in spent pools today and in the future, diversion scenarios that capture a range of possible rod removal options, spatial and isotopic detail needed to accurately quantify the capability of all the NDA techniques so as to enable integration. Integration is being designed into this study from the beginning since it is expected that the best performance will be obtained by combining a few NDA techniques. The performance of each instrument will be quantified for the full assembly library in three different media: air, water and borated water. In this paper the preliminary capability of delayed neutron detection will be quantified for the spent fuel library for all three media. The 13 NDA techniques being researched are the following: Delayed Gamma, Delayed

  5. A METHOD FOR REGENERATION OF SPENT ELECTROCHEMICAL DECONTAMINATION SOLUTION AND ITS TREATMENT FOR FINAL DISPOSAL

    SciTech Connect

    Davydov, D.Yu.; Davydov, Yu.P.; Toropov, I.G.; John, J.; Rosikova, K.; Motl, A.; Hudson, M.J.; Prazska, M.

    2003-02-27

    This paper describes the method of regeneration of spent electrochemical decontamination solution. The proposed method allows separation of radionuclides and stable metals from spent decontamination solution in a form suitable for final disposal and repeated use of the remaining solution for electrochemical decontamination. Development of this method was based on the results of the speciation studies which showed that Fe(III) can be precipitated in the presence of organic complexing agents, in a form of iron hydroxide, and Ag-110m, Co-60, Mn-54 radionuclides can be coprecipitated on it. In order to verify the conclusions made as a result of the speciation studies, the experiments with electrochemically prepared simulant solution and real solution were carried out. The test results proved that the proposed method can be applied in practice. Treatment of the ultimately spent decontamination solutions can be also made applying iron precipitation, which allows for removal of the bulk amount of contaminants, as the first step. Then, if necessary the remaining radionuclides can be removed by sorption. A series of novel absorbers has been tested for their potential for the sorption removal of the remaining radionuclides from the supernate. The test results showed that most of them were more effective in neutral or alkaline range of pH, however, the high efficiency of the sorption removal can be achieved only after the removal of the oxalic and citric acids from solution.

  6. Regeneration process for spent SO/sub 2/-NO/sub x/ sorbents

    SciTech Connect

    Nelson, B.W.; Nelson, S.G.

    1989-05-09

    A regeneration process is described for MgO-vermiculite and MgO-perlite sorbents employed to remove nitrogen and sulfur oxides from a flue gas comprising the following steps: (a) Heating the spent sorbents in air to a temperature in the range of 100/sup 0/ to 350/sup 0/C to drive off substantially all free and chemically attached water; (b) Further heating the spent sorbents in an atmosphere containing a reducing gas selected from the group consisting of carbon monoxide, methane and hydrogen to a temperature in the range of 350/sup 0/ to 450/sup 0/C to drive off sorbed nitrogen oxides; (c) Further heating the spent sorbents in the same atmosphere containing a reducing gas to a temperature in the range of 450/sup 0/ to 700/sup 0/C to drive off approximately 90 percent of the sorbed sulfur in the form of sulfur oxides and elemental sulfur and to destroy substantially all nitrogen oxides present in the exit gases; (d) Cooling the sorbents to a temperature below 200/sup 0/C for reuse.

  7. Prototype spent-fuel canister design, analysis, and test

    SciTech Connect

    Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

    1982-03-01

    Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included.

  8. Thermomechanical modeling of the spent fuel test - Climax

    SciTech Connect

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

    1986-12-31

    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.

  9. Thermomechanical modeling of the Spent Fuel Test-Climax

    SciTech Connect

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

    1986-02-01

    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.

  10. Recent developments at the cathode processor for spent fuel treatment.

    SciTech Connect

    Westphal, B. R.; Vaden, D.; Hua, T. Q.; Willit, J. L.; Laug, D. V.

    2002-07-29

    As part of the spent fuel treatment program at Argonne National Laboratory, a vacuum distillation process is being employed for the recovery of uranium following an electrorefining process. Distillation of a molten salt electrolyte, primarily consisting of a eutectic mixture of lithium and potassium chlorides with minor amounts of fission product chlorides, from uranium is achieved by a batch operation called cathode processing. Described in this paper are recent developments, both equipment and process-related, at the cathode processor during the treatment of blanket-type spent fuel. For the equipment developments, the installation of a new induction heating coil has produced significant improvements in equipment performance. The process developments include the elimination of a process step and the study of plutonium in the uranium product.

  11. The Idaho Spent Fuel Project Update-January, 2003

    SciTech Connect

    Roberts, R.; Tulberg, D.; Carter, C.

    2003-02-25

    The Department of Energy awarded a privatized contract to Foster Wheeler Environmental Corporation in May 2000 for the design, licensing, construction and operation of a spent nuclear fuel repackaging and storage facility. The Foster Wheeler Environmental Team consists of Foster Wheeler Environmental Corp. (the primary contractor), Alstec, RWE-Nukem, RIO Technical Services, Winston and Strawn, and Utility Engineering. The Idaho Spent Fuel (ISF) facility is an integral part of the DOE-EM approach to accelerating SNF disposition at the Idaho National Engineering and Environmental Laboratory (INEEL). Construction of this facility is also important in helping DOE to meet the provisions of the Idaho Settlement Agreement. The ISF Facility is a substantial facility with heavy shielding walls in the repackaging and storage bays and state-of-the-art features required to meet the provisions of 10 CFR 72 requirements. The facility is designed for a 40-year life.

  12. Molten tin reprocessing of spent nuclear fuel elements

    DOEpatents

    Heckman, Richard A.

    1983-01-01

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

  13. Process of regenerating spent HF-HNO sub 3 pickle acid containing (ZrF sub 6 )-2

    SciTech Connect

    Walker, R.G.

    1992-01-21

    This patent describes a process for regenerating spent HF-HNO{sub 3} pickle acid containing (ZrF{sub 6}){sup {minus}2}. It comprises NaNO{sub 3} to a spent HF-HNO{sub 3} pickle acid containing (ZrF{sub 6}){sup {minus}2} to precipitate Na{sub 2}ZrF{sub 6}; and separating the HF-HNO{sub 3} pickle acid from the Na{sub 2}ZrF{sub 6} precipitate.

  14. Handling encapsulated spent fuel in a geologic repository environment

    SciTech Connect

    Ballou, L.B.

    1983-02-01

    In support of the Spent Fuel Test-Climate at the U.S. Department of Energy`s Nevada Test Site, a spent-fuel canister handling system has been designed, deployed, and operated successfully during the past five years. This system transports encapsulated commercial spent-fuel assemblies between the packaging facility and the test site ({similar_to}100 km), transfers the canisters 420 m vertically to and from a geologic storage drift, and emplaces or retrieves the canisters from the storage holes in the floor of the drift. The spent-fuel canisters are maintained in a fully shielded configuration at all times during the handling cycle, permitting manned access at any time for response to any abnormal conditions. All normal operations are conducted by remote control, thus assuring as low as reasonably achievable exposures to operators; specifically, we have had no measurable exposure during 30 canister transfer operations. While not intended to be prototypical of repository handling operations, the system embodies a number of concepts, now demonstrated to be safe, reliable, and economical, which may be very useful in evaluating full-scale repository handling alternatives in the future. Among the potentially significant concepts are: Use of an integral shielding plug to minimize radiation streaming at all transfer interfaces. Hydraulically actuated transfer cask jacking and rotation features to reduce excavation headroom requirements. Use of a dedicated small diameter (0.5 m) drilled shaft for transfer between the surface and repository workings. A wire-line hoisting system with positive emergency braking device which travels with the load. Remotely activated grapples - three used in the system - which are insensitive to load orientation. Rail-mounted underground transfer vehicle operated with no personnel underground.

  15. Spent nuclear fuel canister storage building conceptual design report

    SciTech Connect

    Swenson, C.E.

    1996-01-01

    This Conceptual Design Report provides the technical basis for the Spent Nuclear Fuels Project, Canister Storage Building, and as amended by letter (correspondence number 9555700, M.E. Witherspoon to E.B. Sellers, ``Technical Baseline and Updated Cost Estimate for the Canister Storage Building``, dated October 24, 1995), includes the project cost baseline and Criteria to be used as the basis for starting detailed design in fiscal year 1995.

  16. Machine Vision Tests for Spent Fuel Scrap Characteristics

    SciTech Connect

    BERGER, W.W.

    2000-04-27

    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.

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

    SciTech Connect

    KLEM, M.J.

    2000-10-18

    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

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

    DOEpatents

    Miller, William E.; Richmann, Michael K.

    2004-07-27

    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.

  19. SPENT FUEL CASK IMPACT LIMITER ATTACHMENT DESIGN DEFICIENCIES

    SciTech Connect

    Leduc, D; Jeffery England, J

    2007-10-16

    A recent structural analysis of the T-3 Spent Fuel Containment Cask found problems with the design of the attachment system. Assumptions in the original SARP concerning the loading in the attachment bolts were found to be inaccurate in certain drop orientations. Similar weaknesses in the attachment system designs of other casks were also noted. This paper documents the lessons learned and their applicability to impact limiter attachment system designs.

  20. Naval Spent Nuclear Fuel disposal Container System Description Document

    SciTech Connect

    N. E. Pettit

    2001-07-13

    The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the natural barrier will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

  1. CLASSIFICATION OF THE MGR NAVAL SPENT NUCLEAR FUEL DISPOSAL CONTAINER

    SciTech Connect

    J.A. Ziegler

    1999-08-31

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) naval spent nuclear fuel disposal container system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMF 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998).

  2. Characterization of alloy particles extracted from spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Cui, D.; Rondinella, V. V.; Fortner, J. A.; Kropf, A. J.; Eriksson, L.; Wronkiewicz, D. J.; Spahiu, K.

    2012-01-01

    We characterized, for the first time, submicro- and nanosized fission product-alloy particles that were extracted nondestructively from spent nuclear fuel, in terms of noble metal (Mo-Ru-Tc-Rh-Pd-Te) composition, atomic level homogeneity and lattice parameters. The evidences obtained in this work contribute to an improved understanding of the redox chemistry of radionuclides in nuclear waste repository environments and, in particular, of the catalytic properties of these unique metal alloy particles.

  3. Studies and research concerning BNFP: spent fuel dry storage studies at the Barnwell Nuclear Fuel Plant

    SciTech Connect

    Anderson, Kenneth J.

    1980-09-01

    Conceptual designs are presented utilizing the Barnwell Nuclear Fuel Plant for the dry interim storage of spent light water reactor fuel. Studies were conducted to determine feasible approaches to storing spent fuel by methods other than wet pool storage. Fuel that has had an opportunity to cool for several years, or more, after discharge from a reactor is especially adaptable to dry storage since its thermal load is greatly reduced compared to the thermal load immediately following discharge. A thermal analysis was performed to help in determining the feasibility of various spent fuel dry storage concepts. Methods to reject the heat from dry storage are briefly discussed, which include both active and passive cooling systems. The storage modes reviewed include above and below ground caisson-type storage facilities and numerous variations of vault, or hot cell-type, storage facilities.

  4. Thermoelectric powered wireless sensors for spent fuel monitoring

    SciTech Connect

    Carstens, T.; Corradini, M.; Blanchard, J.; Ma, Z.

    2011-07-01

    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)

  5. The Performance of Spent Fuel Casks in Severe Tunnel Fires

    SciTech Connect

    Bajwa, C.S.; Easton, E.P.; Hansen, A.

    2006-07-01

    The Nuclear Regulatory Commission (NRC), working with the National Institute of Standards and Technology (NIST), Pacific Northwest National Laboratory (PNNL), and the National Transportation Safety Board (NTSB), performed analyses to predict the response of various spent fuel transportation cask designs when exposed to a fire similar to that which occurred in the Howard Street railroad tunnel in downtown Baltimore, Maryland on July 18, 2001. The thermal performance of three different spent fuel cask designs (HOLTEC HI-STAR 100, TransNuclear TN-68, and NAC-LWT) was evaluated with the ANSYS{sup R} and COBRA-SFS analysis codes, utilizing boundary conditions for the tunnel fire obtained using NIST's Fire Dynamics Simulator (FDS) code. NRC Staff evaluated the potential for a release of radioactive material from each of the three transportation casks analyzed for the Baltimore tunnel fire scenario. The results of these analyses are described in detail in Spent Fuel Transportation Package Response to the Baltimore Tunnel Fire Scenario, NUREG/CR-6886, published in draft for comment in November 2005. Comments received by the NRC on NUREG/CR-6886 will be addressed in the final version of the report. (authors)

  6. Dosimetry at an interim storage for spent nuclear fuel.

    PubMed

    Králík, M; Kulich, V; Studeny, J; Pokorny, P

    2007-01-01

    The Czech nuclear power plant Dukovany started its operation in 1985. All fuel spent from 1985 up to the end of 2005 is stored at a dry interim storage, which was designed for 60 CASTOR-440/84 casks. Each of these casks can accommodate 84 fuel assemblies from VVER 440 reactors. Neutron-photon mixed fields around the casks were characterized in terms of ambient dose equivalent measured by standard area dosemeters. Except this, neutron spectra were measured by means of a Bonner sphere spectrometer, and the measured spectra were used to derive the corresponding ambient dose equivalent due to neutrons. PMID:17526479

  7. Preliminary Design Report Shippingport Spent Fuel Drying and Inerting System

    SciTech Connect

    JEPPSON, D.W.

    2000-05-18

    A process description and system flow sheets have been prepared to support the design/build package for the Shippingport Spent Fuel Canister drying and inerting process skid. A process flow diagram was prepared to show the general steps to dry and inert the Shippingport fuel loaded into SSFCs for transport and dry storage. Flow sheets have been prepared to show the flows and conditions for the various steps of the drying and inerting process. Calculations and data supporting the development of the flow sheets are included.

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

    SciTech Connect

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

    1993-04-01

    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.

  9. Method for reprocessing and separating spent nuclear fuels

    DOEpatents

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

    1983-01-01

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

  10. Thermal analysis of cold vacuum drying of spent nuclear fuel

    SciTech Connect

    Piepho, M.G.

    1998-07-20

    The thermal analysis examined transient thermal and chemical behavior of the Multi canister Overpack (MCO) container for a broad range of cases that represent the Cold Vacuum Drying (CVD) processes. The cases were defined to consider both normal and off-normal operations at the CVD Facility for an MCO with Mark IV N, Reactor spent fuel in four fuel baskets and one scrap basket. This analysis provides the basis for the MCO thermal behavior at the CVD Facility for its Phase 2 Safety Analysis Report (revision 4).

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

    DOEpatents

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

    1982-01-19

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

  12. DOE-owned spent nuclear fuel program plan

    SciTech Connect

    1995-11-01

    The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines.

  13. The design of the DUPIC spent fuel bundle counter

    SciTech Connect

    Menlove, H.O.; Rinard, P.M.; Kroncke, K.E.; Lee, Y.G.

    1997-05-01

    A neutron coincidence detector had been designed to measure the amount of curium in the fuel bundles and associated process samples used in the direct use of plutonium in Canadian deuterium-uranium (CANDU) fuel cycle. All of the sample categories are highly radioactive from the fission products contained in the pressurized water reactor (PWR) spent fuel feed stock. Substantial shielding is required to protect the He-3 detectors from the intense gamma rays. The Monte Carlo neutron and photon calculational code has been used to design the counter with a uniform response profile along the length of the CANDU-type fuel bundle. Other samples, including cut PWR rods, process powder, waste, and finished rods, can be measured in the system. This report describes the performance characteristics of the counter and support electronics. 3 refs., 23 figs., 6 tabs.

  14. Capabilities for spent fuel characterization at Argonne National Laboratory

    SciTech Connect

    Neimark, L.A.; Strain, R.V.

    1994-10-01

    Summaries of the status of spent nuclear fuel (SNF) owned by the Department of Energy have highlighted the need to obtain a better understanding of the current physical and chemical condition of the SNF as a foundation for establishing a clear path forward for the fuel`s eventual geologic disposal in a long-term repository. To initiate obtaining the required information, DOE has generated an SNF Characterization Plan based on the needs for characterizing the materials stored at the individual major DOE storage sites. The principal focus of the plan is to characterize those fuel attributes that are key to the safe handling, transportation, and storage of the SNF. The drivers for specific attributes are regulatory requirements, resolution of technical issues, or a design need. Argonne National Laboratory`s facilities in Illinois and Idaho possess capabilities that can be used to address many of the characterization issues that have been raised. This paper will describe these capabilities.

  15. Proposed high throughput electrorefining treatment for spent N- Reactor fuel

    SciTech Connect

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

    1996-05-01

    A high-throughput electrorefining process is being adapted to treat spent N-Reactor fuel for ultimate disposal in a geologic repository. Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the type of fragmentation necessary to provide fuel segments suitable for this process. Based on these tests, a conceptual design was produced of a plant-scale electrorefiner. In this design, the diameter of an electrode assembly is about 1.07 m (42 in.). Three of these assemblies in an electrorefiner would accommodate a 3-metric-ton batch of N-Reactor fuel that would be processed at a rate of 42 kg of uranium per hour.

  16. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    SciTech Connect

    A. Alsaed

    2005-07-28

    The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

  17. Commercial Spent Nuclear Fuel Waste Package Misload Analysis

    SciTech Connect

    J.K. Knudson

    2003-10-02

    The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis

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

    SciTech Connect

    Durbin, Samuel G.; Lindgren, Eric Richard

    2014-03-01

    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

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

    SciTech Connect

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

    1997-08-01

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

  20. 10 CFR 72.214 - List of approved spent fuel storage casks.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false List of approved spent fuel storage casks. 72.214 Section 72.214 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General License for Storage of Spent...

  1. Development of Technologies for the Simultaneous Separation of Cesium and Strontium from Spent Nuclear Fuel as Part of an Advanced Fuel Cycle

    SciTech Connect

    Jack D. Law; R. Scott HErbst; David H. Meikrantz; Dean R. Peterman; Catherine L. Riddle; Richard D. Tillotson; Terry A. Todd

    2005-04-01

    As part of the Advanced Fuel Cycle Initiative, two solvent extraction technologies are being developed to simultaneously separate cesium and strontium from dissolved spent nuclear fuel. The first process utilizes a solvent consisting of chlorinated cobalt dicarbollide and polyethylene glycol extractants in a phenyltrifluoromethyl sulfone diluent. Recent improvements to the process include development of a new, non-nitroaromatic diluent and development of new stripping reagents, including a regenerable strip reagent that can be recovered and recycled. Countercurrent flowsheets have been designed and tested on simulated and actual spent nuclear fuel feed streams with both cesium and strontium removal efficiencies of greater than 99 %. The second process developed to simultaneously separate cesium and strontium from spent nuclear fuel is based on two highly-specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. A solvent composition has been developed that enables both elements to be removed together and, in fact, a synergistic effect was observed with strontium distributions in the combined solvent that are much higher that in the strontium extraction (SREX) process. Initial laboratory test results of the new combined cesium and strontium extraction process indicate good extraction and stripping performance. A flowsheet for treatment of spent nuclear fuel is currently being developed.

  2. The synchronous active neutron detection system for spent fuel assay

    SciTech Connect

    Pickrell, M.M.; Kendall, P.K.

    1994-10-01

    The authors have begun to develop a novel technique for active neutron assay of fissile material in spent nuclear fuel. This approach will exploit the unique operating features of a 14-MeV neutron generator developed by Schlumberger. This generator and a novel detection system will be applied to the direct measurement of the fissile material content in spent fuel in place of the indirect measures used at present. The technique they are investigating is termed synchronous active neutron detection (SAND). It closely follows a method that has been used routinely in other branches of physics to detect very small signals in the presence of large backgrounds. Synchronous detection instruments are widely available commercially and are termed {open_quotes}lock-in{close_quotes} amplifiers. The authors have implemented a digital lock-in amplifier in conjunction with the Schlumberger neutron generator to explore the possibility of synchronous detection with active neutrons. This approach is possible because the Schlumberger system can operate at up to a 50% duty factor, in effect, a square wave of neutron yield. The results to date are preliminary but quite promising. The system is capable of resolving the fissile material contained in a small fraction of the fuel rods in a cold fuel assembly. It also appears to be quite resilient to background neutron interference. The interrogating neutrons appear to be nonthermal and penetrating. Although a significant amount of work remains to fully explore the relevant physics and optimize the instrument design, the underlying concept appears sound.

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

    SciTech Connect

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L.; Moore, E.N.

    2013-07-01

    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

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

    SciTech Connect

    Drayer, R.

    2013-06-09

    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

  5. Accelerator-driven transmutation of spent fuel elements

    DOEpatents

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    2002-01-01

    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

  6. Determination of plutonium content in TRR spent fuel by nondestructive neutron counting

    NASA Astrophysics Data System (ADS)

    Chen, Yen-Fu; Sheu, Rong-Jiun; Chiao, Ling-Huan; Yuan, Ming-Chen; Jiang, Shiang-Huei

    2010-07-01

    For the nuclear safeguard purpose, this work aims to nondestructively determine the plutonium content in the Taiwan Research Reactor (TRR) spent fuel rods in the storage pool before the stabilization process, which transforms the metal spent fuel rods into oxide powder. A SPent-fuel-Neutron-Counter (SPNC) system was designed and constructed to carry out underwater scan measurements of neutrons emitting from the spent fuel rod, from which the 240Pu mass in the fuel rod will be determined. The SAS2 H control module of the SCALE 5.1 code package was applied to calculate the 240Pu-to-Pu mass ratio in the TRR spent fuel rod according to the given power history. This paper presents the methodology and design of our detector system as well as the measurements of four TRR spent fuel rods in the storage pool and the comparison of the measured results with the facility declared values.

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

    SciTech Connect

    Parker, Frank L.

    2012-07-01

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

  8. DECONTAMINATION OF ZIRCALOY CLADDING HULLS FROM SPENT NUCLEAR FUEL

    SciTech Connect

    Rudisill, T.

    2010-09-29

    The feasibility of decontaminating spent fuel cladding hulls using hydrofluoric acid (HF) was investigated as part of the Global Energy Nuclear Partnership (GNEP) Separations Campaign. The concentrations of the fission product and transuranic (TRU) isotopes in the decontaminated hulls were compared to the limits for determining the low level waste (LLW) classification in the United States (US). The {sup 90}Sr and {sup 137}Cs concentrations met the disposal criteria for a Class C LLW; although, in a number of experiments the criteria for disposal as a Class B LLW were met. The TRU concentration in the hulls generally exceeded the Class C LLW limit by at least an order of magnitude. The concentration decreased sharply as the initial 30-40 {micro}m of the cladding hull surface were removed. At depths beyond this point, the TRU activity remained relatively constant, well above the Class C limit. Reprocessing of spent nuclear fuel generates a cladding waste which would likely require disposal as a Greater than Class C LLW in the US. If the cladding hulls could be treated to remove a majority of the actinide and fission product contamination, the hulls could potentially meet acceptance criteria for disposal as a LLW or allow recycle of the Zr metal. Discard of the hulls as a LLW would result in significant cost savings compared to disposal as a Greater than Class C waste which currently has no disposition path. During fuel irradiation and reprocessing, radioactive materials are produced and deposited in the Zircaloy cladding. Due to short depths of penetration, the majority of the fission products and actinide elements are located in the ZrO{sub 2} layer which forms on the surface of the cladding during fuel irradiation. Therefore, if the oxide layer is removed, the majority of the contamination should also be removed. It is very difficult, if not impossible to remove all of the activity from spent fuel cladding since traces of U and Th in the unirradiated Zircaloy

  9. Determining plutonium in spent fuel with nondestructive assay techniques

    SciTech Connect

    Tobin, Stephen J; Charlton, William S; Fensin, Michael L; Menlove, Howard O; Hoover, A S; Quiter, B J; Rajasingam, A; Swinhoe, M T; Thompson, S J; Charlton, W S; Ehinger, M H; Sandoval, N P; Saavedra, S F; Strohmeyer, D

    2009-01-01

    There are a variety of motivations for quantifying plutonium in used (spent) fuel assemblies by means of nondestructive assay including the following: shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories or fuel storage facilities. Twelve NDA techniques were identified that provide information about the composition of an assembly. Unfortunately, none of these techniques is capable of determining the Pu mass in an assembly on its own. However, it is expected that the Pu mass can be quantified by combining a few of the techniques. Determining which techniques to combine and estimating the expected performance of such a system is the purpose of the research effort recently begun. The research presented here is a complimentarily experimental effort. This paper will focus on experimental results of one of the twelve non-destructive assay techniques - passive neutron albedo reactivity. The passive neutron albedo reactivity techniques work by changing the multiplication the pin experiences between two separate measurements. Since a single spent fuel pin has very little multiplication, this is a challenging measurement situation for the technique. Singles and Doubles neutron count rate were measured at Oak Ridge National Laboratory for three different burnup pins to test the capability of the passive neutron albedo reactivity technique.

  10. Determination of BWR Spent Nuclear Fuel Assembly Effective Thermal Conductivity

    SciTech Connect

    Matthew D. Hinds

    2001-10-17

    The purpose of this calculation is to provide an effective thermal conductivity for use in predicting peak cladding temperatures in boiling water reactor (BWR) fuel assemblies with 7x7,8x8, and 9x9 rod arrays. The first objective of this calculation is to describe the development and application of a finite element representation that predicts peak spent nuclear fuel temperatures for BWR assemblies. The second objective is to use the discrete representation to develop a basis for determining an effective thermal conductivity (described later) for a BWR assembly with srneared/homogeneous properties and to investigate the thermal behavior of a spent fuel assembly. The scope of this calculation is limited to a steady-state two-dimensional representation of the waste package interior region. This calculation is subject to procedure AP-3.124, Calculations (Ref. 27) and guided by the applicable technical work plan (Ref. 14). While these evaluations were originally developed for the thermal analysis of conceptual waste package designs emplaced in the potential repository at Yucca Mountain, the methodology applies to storage and transportation thermal analyses as well. Note that the waste package sketch in Attachment V depicts a preliminary design, and should not be interpreted otherwise.

  11. Modelling of radiation field around spent fuel container.

    PubMed

    Kryuchkov, E F; Opalovsky, V A; Tikhomirov, G V

    2005-01-01

    Operation of nuclear reactors leads to the production of spent nuclear fuel (SNF). There are two basic strategies of SNF management: ultimate disposal of SNF in geological formations and recycle or repeated utilisation of reprocessed SNF. In both options, there is an urgent necessity to study radiation properties of SNF. Information about SNF radiation properties is required at all stages of SNF management. In order to reach more effective utilisation of nuclear materials, new fuel cycles are under development based on uranium-plutonium, uranium-thorium and some other types of nuclear fuel. These promising types of nuclear fuel are characterised by quite different radiation properties at all the stages of nuclear fuel cycle (NFC) listed above. So, comparative analysis is required for radiation properties of different nuclear fuel types at different NFC stages. The results presented here were obtained from the numerical analysis of the radiation field around transport containers of different SNF types and in SNF storage. The calculations are carried out with the application of the computer code packages SCALE-4.3 and MCNP-4C. Comparison of the dose parameters obtained for different models of the transport container with experimental data allowed us to make certain conclusions about the errors of numerical results caused by the approximate geometrical description of the transport container. PMID:16604702

  12. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    SciTech Connect

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  13. Spent Nuclear Fuel (SNF) Project Safety Basis Implementation Strategy

    SciTech Connect

    TRAWINSKI, B.J.

    2000-02-08

    The objective of the Safety Basis Implementation is to ensure that implementation of activities is accomplished in order to support readiness to move spent fuel from K West Basin. Activities may be performed directly by the Safety Basis Implementation Team or they may be performed by other organizations and tracked by the Team. This strategy will focus on five key elements, (1) Administration of Safety Basis Implementation (general items), (2) Implementing documents, (3) Implementing equipment (including verification of operability), (4) Training, (5) SNF Project Technical Requirements (STRS) database system.

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

    SciTech Connect

    McCormack, R.L.

    1995-02-02

    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.

  15. Contaminated sediment removal from a spent fuel storage canal

    SciTech Connect

    Geber, K R

    1993-01-01

    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.

  16. Closure mechanism and method for spent nuclear fuel canisters

    DOEpatents

    Doman, Marvin J.

    2004-11-23

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

    SciTech Connect

    Rodney C. Ewing

    2004-10-07

    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.

  18. Production of metal waste forms from spent fuel treatment

    SciTech Connect

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

    1995-02-01

    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.

  19. Compact approach to monitored retrievable storage of spent fuel

    SciTech Connect

    Muir, D.W.

    1984-09-01

    Recent federal waste-management legislation has raised national interest in monitored retrievable storage (MRS) of unprocessed spent fuel from civilian nuclear power plants. We have reviewed the current MRS design approaches, and we have examined an alternative concept that is extremely compact in terms of total land use. This approach may offer substantial advantages in the areas of monitoring and in safeguards against theft, as well as in reducing the chances of groundwater contamination. Total facility costs are roughly estimated and found to be generally competitive with other MRS concepts. 4 references, 3 figures, 3 tables.

  20. Closure Mechanism and Method for Spent Nuclear Fuel Canisters

    SciTech Connect

    Doman, Marvin J.

    2004-11-23

    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.

  1. The shutdown reactor: Optimizing spent fuel storage cost

    SciTech Connect

    Pennington, C.W.

    1995-12-31

    Several studies have indicated that the most prudent way to store fuel at a shutdown reactor site safely and economically is through the use of a dry storage facility licensed under 10CFR72. While such storage is certainly safe, is it true that the dry ISFSI represents the safest and most economical approach for the utility? While no one is really able to answer that question definitely, as yet, Holtec has studied this issue for some time and believes that both an economic and safety case can be made for an optimization strategy that calls for the use of both wet and dry ISFSI storage of spent fuel at some plants. For the sake of brevity, this paper summarizes some of Holtec`s findings with respect to the economics of maintaining some fuel in wet storage at a shutdown reactor. The safety issue, or more importantly the perception of safety of spent fuel in wet storage, still varies too much with the eye of the beholder, and until a more rigorous presentation of safety analyses can be made in a regulatory setting, it is not practically useful to argue about how many angels can sit on the head of a safety-related pin. Holtec is prepared to present such analyses, but this does not appear to be the proper venue. Thus, this paper simply looks at certain economic elements of a wet ISFSI at a shutdown reactor to make a prima facie case that wet storage has some attractiveness at a shutdown reactor and should not be rejected out of hand. Indeed, an optimization study at certain plants may well show the economic vitality of keeping some fuel in the pool and converting the NRC licensing coverage from 10CFR50 to 10CFR72. If the economics look attractive, then the safety issue may be confronted with a compelling interest.

  2. Monticello BWR spent fuel assembly decay heat predictions and measurements

    SciTech Connect

    McKinnon, M.A.; Doman, J.W.; Heeb, C.M.; Creer, J.M.

    1986-06-01

    This report compares pre-calorimetry predictions of rates of six 7 x 7 boiling water reactor (BWR) spent fuel assemblies with measured decay heat rates. The assemblies were from Northern States Power Company's Monticello Nuclear Generating Plant and had burnups of 9 to 21 GWd/MTU and cooling times of 9 to 10 years. Conclusions are: The agreement between ORIGEN2 predictions and decay heat measurements of Monticello spent fuel is dependent on the method used to calibrate the calorimeter and to make the decay heat measurements. The agreement between predictions and measurements of decay heat rates of Monticello fuel is the same as that for Cooper and Dresden fuel if the same measurement method is used. The predictions are within a standard deviation of +-15 W of the measurements. Using a different measurement method, ORIGEN2 underpredicts the measured decay heat output of Monticello fuel assemblies by a constant 20 +- 2 W. The 20-W offset appears to be an artifact of the calibration procedure. The constant term in the calibration curve (i.e., q/sub DH/ = mx + b) can account for measurement differences of 40 W based on the 1983, 1984, and 1985 calibration curves. The difference between ORIGEN2 predictions and calorimeter decay heat measurements does not appear to be dependent on the magnitude of decay heat output. Predicted axial decay heat profiles are in good agreement with measured axial gamma radiation profiles. Recommendations are: Predictions using other decay heat codes should be compared to experimental data contained in this report, to evaluate prediction capabilities. The source of the differences that exist among calorimeter calibration curves needs to be determined. Calorimeter operational methods need to be investigated further to determine cause and effect relationships between operational method and calorimeter precision and accuracy.

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

    SciTech Connect

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

    1987-03-01

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

  4. Technical Development on Burn-up Credit for Spent LWR Fuel

    SciTech Connect

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  5. 78 FR 40199 - Draft Spent Fuel Storage and Transportation Interim Staff Guidance

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-03

    ... COMMISSION Draft Spent Fuel Storage and Transportation Interim Staff Guidance AGENCY: Nuclear Regulatory... Regulatory Commission (NRC) requests public comment on Draft Spent Fuel Storage and Transportation Interim... Integrity for Continued Storage of High Burnup Fuel Beyond 20 Years.'' The draft SFST-ISG provides...

  6. 78 FR 66858 - Waste Confidence-Continued Storage of Spent Nuclear Fuel

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-07

    ... storage of spent nuclear fuel beyond a reactor's licensed life for operation and prior to ultimate... generically addresses the environmental impacts of continued storage of spent nuclear fuel beyond the licensed... fuel beyond a reactor's licensed life for operation and prior to ultimate disposal. (78 FR 56776)....

  7. PROGRESS REPORT. CORROSION OF SPENT NUCLEAR FUEL: THE LONG-TERM ASSESSMENT

    EPA Science Inventory

    The successful disposal of spent nuclear fuel (SNF) is one of the most serious challenges to the success of the nuclear fuel cycle and the future of nuclear power generation. Spent nuclear fuel is essentially UO2 with approximately 4-5 atomic percent actinides and fission product...

  8. DIRECT INVESTIGATIONS OF THE IMMOBILIZATION OF RADIONUCLIDES IN THE ALTERATION PHASES OF SPENT NUCLEAR FUEL

    EPA Science Inventory

    DOE is the custodian of several thousand tons of spent nuclear fuel that is intended for geological disposal. The direct disposal of spent nuclear fuel or of mixed oxide fuel (fabricated for the disposal of excess weapons plutonium) requires a careful analysis of the role of spen...

  9. Spent fuel handling system for a geologic storage test at the Nevada Test Site

    SciTech Connect

    Duncan, J.E.; House, P.A.; Wright, G.W.

    1980-05-01

    The Lawrence Livermore Laboratory is conducting a test of the geologic storage of encapsulated spent commercial reactor fuel assemblies in a granitic rock at the Nevada Test Site. The test, known as the Spent Fuel Test-Climax (SFT-C), is sponsored by the US Department of Energy, Nevada Operations Office. Eleven pressurized-water-reactor spent fuel assemblies are stored retrievably for three to five years in a linear array in the Climax stock at a depth of 420 m.

  10. Radiation dose rates from commercial PWR and BWR spent fuel elements

    SciTech Connect

    Willingham, C.E.

    1981-10-01

    Data on measurements of gamma dose rates from commercial reactor spent fuel were collected, and documented calculated gamma dose rates were reviewed. As part of this study, the gamma dose rate from spent fuel was estimated, using computational techniques similar to previous investigations into this problem. Comparison of the measured and calculated dose rates provided a recommended dose rate in air versus distance curve for PWR spent fuel.

  11. Spent Fuel NDA Research Path for the Sweden Encapsulation-Repository

    SciTech Connect

    Tobin, Stephen J.; Trellue, Holly R.; Liljenfeldt, Henrik

    2015-01-22

    This set of slides provides a description of research performed to date on spent fuel NDA: Next Generation Safeguards Initiative Spent Fuel Project, and NDA analysis and research planned for CLINK. The general purpose is strengthening the technical toolkit of safeguard inspectors. Data mining is being applied to determine the optimal mathematical structure to match the complexity of spent fuel NDA signals and to enable a range of quantities to be estimated.

  12. Spent nuclear fuel storage. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect

    1997-07-01

    The bibliography contains citations concerning spent nuclear fuel storage technologies, facilities, sites, and assessment. References review wet and dry storage, spent fuel casks and pools, underground storage, monitored and retrievable storage systems, and aluminum-clad spent fuels. Environmental impact, siting criteria, regulations, and risk assessment are also discussed. Computer codes and models for storage safety are covered. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  13. Regeneration of field-spent activated carbon catalysts for low-temperature selective catalytic reduction of NOx with NH3

    SciTech Connect

    Jeon, Jong Ki; Kim, Hyeonjoo; Park, Young-Kwon; Peden, Charles HF; Kim, Do Heui

    2011-10-15

    In the process of producing liquid crystal displays (LCD), the emitted NOx is removed over an activated carbon catalyst by using selective catalytic reduction (SCR) with NH3 at low temperature. However, the catalyst rapidly deactivates primarily due to the deposition of boron discharged from the process onto the catalyst. Therefore, this study is aimed at developing an optimal regeneration process to remove boron from field-spent carbon catalysts. The spent carbon catalysts were regenerated by washing with a surfactant followed by drying and calcination. The physicochemical properties before and after the regeneration were investigated by using elemental analysis, TG/DTG (thermogravimetric/differential thermogravimetric) analysis, N2 adsorption-desorption and NH3 TPD (temperature programmed desorption). Spent carbon catalysts demonstrated a drastic decrease in DeNOx activity mainly due to heavy deposition of boron. Boron was accumulated to depths of about 50 {mu}m inside the granule surface of the activated carbons, as evidenced by cross-sectional SEM-EDX analysis. However, catalyst activity and surface area were significantly recovered by removing boron in the regeneration process, and the highest NOx conversions were obtained after washing with a non-ionic surfactant in H2O at 70 C, followed by treatment with N2 at 550 C.

  14. Deformation and fracture characteristics of spent Zircaloy fuel cladding

    SciTech Connect

    Chung, H.M.; Yaggee, F.L.

    1982-09-01

    For a better understanding of Zircaloy fuel-rod failure by the pellet-cladding interaction (PCI) phenomenon, a mechanistic study of deformation and fracture behavior of spent power reactor fuel cladding under simulated PCI conditions was conducted. Zircaloy-2 cladding specimens, obtained from fuel assemblies of operating power reactors, were deformed to fracture at 325/sup 0/C by internal gas pressurization in the absence of fission product simulants. Fracture characteristics and microstructures were examined via SEM, TEM, and HVEM. Numerous dislocation tangles and cell structures, observed in TEM specimens of cladding tubes that failed in a ductile manner, were consistent with SEM observations of a limited number of dimples characteristic of microvoid coalescence. A number of brittle-type failures were produced without the influence of fission product simulants. The brittle cracks occurred near the areas compressed by the Swagelok fittings of the internally pressurized tube and propagated from the outer to the inner surface. Since the outer surface was isolated and maintained under a flowing stream of pure helium, it is unlikely that the brittle-type failure was influenced by any fission product traces. SEM fractography of the brittle-type failure revealed a large area of transgranular pseudocleavage with limited areas of ductile fluting, which were similar in appearance to the surfaces produced by in-reactor PCI-type failures. A TEM evaluation of the cladding in the vicinity of the through-wall crack revealed numerous locations that contained an extensive amount of second-phase precipitate (Zr/sub 3/O). We believe that the brittle-type failures of the irradiated spent fuel cladding in the stress rupture experiments are associated with segregation of oxygen, which leads to the formation of the order structure, an immobilization of dislocations, and minimal plastic deformation in the material.

  15. Managing Spent Nuclear Fuel at the Idaho National Laboratory

    SciTech Connect

    Thomas Hill; Denzel L. Fillmore

    2005-10-01

    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. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-12-31

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

  17. Radiation induced corrosion of copper for spent nuclear fuel storage

    NASA Astrophysics Data System (ADS)

    Björkbacka, Åsa; Hosseinpour, Saman; Johnson, Magnus; Leygraf, Christofer; Jonsson, Mats

    2013-11-01

    The long term safety of repositories for radioactive waste is one of the main concerns for countries utilizing nuclear power. The integrity of engineered and natural barriers in such repositories must be carefully evaluated in order to minimize the release of radionuclides to the biosphere. One of the most developed concepts of long term storage of spent nuclear fuel is the Swedish KBS-3 method. According to this method, the spent fuel will be sealed inside copper canisters surrounded by bentonite clay and placed 500 m down in stable bedrock. Despite the importance of the process of radiation induced corrosion of copper, relatively few studies have been reported. In this work the effect of the total gamma dose on radiation induced corrosion of copper in anoxic pure water has been studied experimentally. Copper samples submerged in water were exposed to a series of total doses using three different dose rates. Unirradiated samples were used as reference samples throughout. The copper surfaces were examined qualitatively using IRAS and XPS and quantitatively using cathodic reduction. The concentration of copper in solution after irradiation was measured using ICP-AES. The influence of aqueous radiation chemistry on the corrosion process was evaluated based on numerical simulations. The experiments show that the dissolution as well as the oxide layer thickness increase upon radiation. Interestingly, the evaluation using numerical simulations indicates that aqueous radiation chemistry is not the only process driving the corrosion of copper in these systems.

  18. Thermal hydraulic feasibility assessment for the Spent Nuclear Fuel Project

    SciTech Connect

    Heard, F.J.; Cramer, E.R.; Beaver, T.R.; Thurgood, M.J.

    1996-01-01

    A series of scoping analyses have been completed investigating the thermal-hydraulic performance and feasibility of the Spent Nuclear Fuel Project (SNFP) Integrated Process Strategy (IPS). The SNFP was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy`s Hanford Site in Richland, Washington. The subject efforts focused on independently investigating, quantifying, and establishing the governing heat production and removal mechanisms for each of the IPS operations and configurations, obtaining preliminary results for comparison with and verification of other analyses, and providing technology-based recommendations for consideration and incorporation into the design bases for the SNFP. The goal was to develop a series fo thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the SNFP. A series of sensitivity analyses were also performed to help identify those parameters that have the greatest impact on energy transfer and hence, temperature control. It is anticipated that the subject thermal-hydraulic models will form the basis for a series of advanced and more detailed models that will more accurately reflect the thermal performance of the IPS and alleviate the necessity for some of the more conservative assumptions and oversimplifications, as well as form the basis for the final process and safety analyses.

  19. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect

    Not Available

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  20. Increase of inherent protection level in spent nuclear fuel

    SciTech Connect

    Krasnobaev, A.; Kryuchkov, E.; Glebov, V.

    2006-07-01

    The paper is devoted to upgrading inherent proliferation protection of fissionable nuclear materials (FNM). Some possibilities were investigated to form high radiation barrier inside spent fuel assemblies (SFA) discharged from power reactors of VVER-1000 type and research reactors of IRT type. The radiation barrier is estimated in the terms of rate of equivalent dose (RED) at 30-cm distance from SFA. The values of RED were calculated with application of the computer code package SCALE 4.3. The paper considers the criteria adopted for estimation of FNM proliferation resistance. The paper presents numerical results on a component-wise analysis of the radiation barrier in SFA from reactors of VVER-1000 and IRT type and on capability of various radionuclides to prolong action of the radiation barrier. Isotopic admixtures were selected and amounts of these admixtures were evaluated for significant prolongation of the radiation barrier action at the levels of the radiation standards used for estimation of FNM proliferation resistance. The paper considers vulnerability of the radiation barriers in respect to thermal processing of spent fuel. (authors)

  1. Petite sismique measurements at the Spent Fuel Test - Climax

    SciTech Connect

    Zucca, J.J.

    1984-09-01

    In May 1984, a petite sismique estimate of the deformation modulus (E) was carried out at the Spent Fuel Test - Climax (SFT-C) at the Nevada Test site. The first part of the experiment was to repeat an earlier suite of measurements that were taken before the spent fuel was emplaced to see if any changes had resulted from heating the rock mass. The results of this measurement indicate a decrease in the modulus. However, these results are suspect in view of the findings in the second part of the experiment, which was designed to minimize the effects due to spurious resonances in the source and geophone locations. These effects were thought to bias the earlier measurements. The measurements indicate that the rock acts as a low-pass filter to the propagating wavefield. Furthermore, it is noted that the blow from a hammer is not a purely impulsive source. Therefore, depending on the type of source used and the distance away from the source, a different peak frequency and, hence, E could be measured for the same rock mass. Unless these effects are somehow factored out of a petite sismique survey, the value of E obtained could be severely biased. 20 figures.

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

    SciTech Connect

    Canaan, R.E.

    1995-12-01

    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.

  3. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect

    Not Available

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  4. Spent Fuel and High-Level Radioactive Waste Transportation Report

    SciTech Connect

    Not Available

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  5. X-ray fluorescence analysis of spent reactor fuel solutions

    SciTech Connect

    Berdikov, V.V.; Iokhin, B.S.

    1987-01-01

    The present work examines the possibility of laboratory analytic control by means of a crystal-free x-ray fluorescence analysis (XRFA) installation specifically designed for this purpose with a prior selection of fluorescent emissions by energy. Monitoring by fluorescent emissions from selenium positioned in the installation along with a metering cuvette allows the exclusion of errors associated with the instability of the x-ray tube beam and the effects of the superpositioning of impulses and of dead time. Tests were made on the possibility of using hard standards, consisting of pellets of uranium-aluminum alloy in a fluoroplastic casing. Some characteristics are shown of the crystal-free x-ray fluorescent method of determining uranium concentration in the technological products from reprocessing spent fuel from a water-moderated, water-cooled reactor. The determination of the uranium content in diluted incoming solutions in the reprocessing of spent fuel from such a reactor was also tested by comparison with the results of the method of isotopic dilution concluded using mass-spectrometry.

  6. Spent nuclear fuel recycling with plasma reduction and etching

    DOEpatents

    Kim, Yong Ho

    2012-06-05

    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.

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

    SciTech Connect

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

    2012-04-01

    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.

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

    SciTech Connect

    Chad Pope

    2007-05-01

    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

  9. Dry transfer system for spent fuel: Project report, A system designed to achieve the dry transfer of bare spent fuel between two casks. Final report

    SciTech Connect

    Dawson, D.M.; Guerra, G.; Neider, T.; Shih, P.

    1995-12-01

    This report describes the system developed by EPRI/DOE for the dry transfer of spent fuel assemblies outside the reactor spent fuel pool. The system is designed to allow spent fuel assemblies to be removed from a spent fuel pool in a small cask, transported to the transfer facility, and transferred to a larger cask, either for off-site transportation or on-site storage. With design modifications, this design is capable of transferring single spent fuel assemblies from dry storage casks to transportation casks or visa versa. One incentive for the development of this design is that utilities with limited lifting capacity or other physical or regulatory constraints are limited in their ability to utilize the current, more efficient transportation and storage cask designs. In addition, DOE, in planning to develop and implement the multi-purpose canister (MPC) system for the Civilian Radioactive Waste Management System, included the concept of an on-site dry transfer system to support the implementation of the MPC system at reactors with limitations that preclude the handling of the MPC system transfer casks. This Dry Transfer System can also be used at reactors wi decommissioned spent fuel pools and fuel in dry storage in non-MPC systems to transfer fuel into transportation casks. It can also be used at off-reactor site interim storage facilities for the same purpose.

  10. Stabilization and/or regeneration of spent sorbents from coal gasification. [Quarterly] technical report, December 1, 1991--February 29, 1992

    SciTech Connect

    Abbasian, J.; Hill, A.H.; Wangerow, J.R.

    1992-08-01

    The objective of this investigation is to determine the effects of SO, partial pressure and reaction temperature on the conversion of sulfide containing solid wastes from coal gasifiers to stable and environmentally acceptable calcium-sulfate, while preventing the release of sulfur dioxide through undesirable side reactions during the stabilization step. An additional objective of this program is to investigate the use of the Spent Sorbent Regeneration Process (SSRP) to regenerate spent limestone, from a fluidized-bed gasifier with in-bed sulfur capture, for recycling to the gasifier. To achieve these objectives, selected samples of partially sulfided sorbents will be reacted with oxygen at a variety of operating conditions under sufficient S0{sub 2} partial pressure to prevent release of sulfur from the solids during stabilization that reduces the overall sorbent utilization. Partially sulfided limestone will also be regenerated with water to produce calcium hydroxide and release sulfur as H{sub 2}S. The regenerated sorbent will be dewatered, dried and pelletized. The reactivity of the regenerated sorbent toward H{sub 2}S will also be determined.

  11. 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

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

    2000-09-01

    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

  12. Spent Nuclear Fuel (SNF) Project Acceptance Criteria for Light Water Reactor Spent Fuel Storage System [OCRWM PER REV2

    SciTech Connect

    JOHNSON, D.M.

    2000-12-20

    As part of the decommissioning of the 324 Building Radiochemical Engineering Cells there is a need to remove commercial Light Water Reactor (LWR) spent nuclear fuel (SNF) presently stored in these hot cells. To enable fuel removal from the hot cells, the commercial LWR SNF will be packaged and shipped to the 200 Area Interim Storage Area (ISA) in a manner that satisfies site requirements for SNF interim storage. This document identifies the criteria that the 324 Building Radiochemical Engineering Cell Clean-out Project must satisfy for acceptance of the LWR SNF by the SNF Project at the 200 Area ISA. In addition to the acceptance criteria identified herein, acceptance is contingent on adherence to applicable Project Hanford Management Contract requirements and procedures in place at the time of work execution.

  13. Hanford K Basins spent nuclear fuels project update

    SciTech Connect

    Hudson, F.G.

    1997-10-17

    Twenty one hundred metric tons of spent nuclear fuel are stored in two concrete pools on the Hanford Site, known as the K Basins, near the Columbia River. The deteriorating conditions of the fuel and the basins provide engineering and management challenges to assure safe current and future storage. DE and S Hanford, Inc., part of the Fluor Daniel Hanford, Inc. lead team on the Project Hanford Management Contract, is constructing facilities and systems to move the fuel from current wet pool storage to a dry interim storage facility away from the Columbia River, and to treat and dispose of K Basins sludge, debris and water. The process starts in the K Basins where fuel elements will be removed from existing canisters, washed, and separated from sludge and scrap fuel pieces. Fuel elements will be placed in baskets and loaded into Multi-Canister Overpacks (MCOs) and into transportation casks. The MCO and cask will be transported into the Cold Vacuum Drying Facility, where free water within the MCO will be removed under vacuum at slightly elevated temperatures. The MCOs will be sealed and transported via the transport cask to the Canister Storage Building (CSB) in the 200 Area for staging prior to hot conditioning. The conditioning step to remove chemically bound water is performed by holding the MCO at 300 C under vacuum. This step is necessary to prevent excessive pressure buildup during interim storage that could be caused by corrosion. After conditioning, MCOs will remain in the CSB for interim storage until a national repository is completed.

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

    SciTech Connect

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

    2004-12-27

    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.

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

    SciTech Connect

    Hill, Thomas J

    2005-09-01

    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

  16. The resistance to impact of spent Magnox fuel transport flasks

    SciTech Connect

    Not Available

    1985-01-01

    This book completes the papers of the four-year programme of research and demonstrations embarked upon by the CEGB in 1981, culminating in the spectacular train crash at Old Dalby in July 1984. It explains the CEGB's operations in relation to the transportation of spent Magnox fuel. The public tests described in this book are more effective in improving public understanding and confidence than any amount of explanations could have been, raising the wider question of how best the scientific community can respond to the legitimate concerns of the man and woman in the street about the generating of electricity from nuclear power. The contents are: Taking care; irradiated fuel transport in the UK; programming for flask safety; the use of scale models in impact testing; flask analytical studies; drop test facilities; demonstration drop test; a study of flask transport impact hazards; impact of Magnox irradiated fuel transport flasks into rock and concrete; rail crash demonstration scenarios; horizontal impact testing of quarter scale flasks using masonry targets; horizontal crash testing and analysis of model flatrols; flatrol test; analysis of full scale impact into an abutment; analysis of primary impact forces in the train crash demonstration; horizontal impact tests of quarter scale Magnox flasks and stylised model locomotives; predictive estimates for behaviour in the train crash demonstration; design and organization of the crash; execution of the crash demonstration by British Rail; instrumentation for the train crash demonstration; photography for the crash demonstration; a summary of the CEGB's flask accident impact studies.

  17. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Specific requirements for spent fuel storage cask approval and fabrication. 72.236 Section 72.236 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND...

  18. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... for each power reactor holding a 10 CFR part 50 license that is in a decommissioning or possession-only status and has spent fuel onsite, and for each independent spent fuel storage 10 CFR part 72... storage licenses. 171.15 Section 171.15 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ANNUAL FEES...

  19. 10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Conditions for spent fuel storage cask renewal. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  20. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  1. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Conditions for spent fuel storage cask reapproval. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  2. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Specific requirements for spent fuel storage cask approval and fabrication. 72.236 Section 72.236 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND...

  3. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Specific requirements for spent fuel storage cask approval and fabrication. 72.236 Section 72.236 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE...

  4. 10 CFR 72.240 - Conditions for spent fuel storage cask reapproval.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Conditions for spent fuel storage cask reapproval. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  5. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  6. 10 CFR 72.240 - Conditions for spent fuel storage cask renewal.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Conditions for spent fuel storage cask renewal. 72.240 Section 72.240 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  7. 10 CFR 72.236 - Specific requirements for spent fuel storage cask approval and fabrication.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Specific requirements for spent fuel storage cask approval and fabrication. 72.236 Section 72.236 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND...

  8. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  9. 10 CFR 72.230 - Procedures for spent fuel storage cask submittals.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Procedures for spent fuel storage cask submittals. 72.230 Section 72.230 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  10. RADIOLYTIC AND THERMAL PROCESSES RELEVANT TO DRY STORAGE OF SPENT NUCLEAR FUELS

    EPA Science Inventory

    Thousands of tons of metallic uranium spent-nuclear-fuel (SNF) remain in water storage across the Department of Energy complex. For example, the Hanford Site K-Basins hold 2300 metric tons of spent fuel, much of it severely corroded. Similar situations exist elsewhere in the DOE ...

  11. 76 FR 70331 - List of Approved Spent Fuel Storage Casks: MAGNASTOR ® System, Revision 2

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-14

    ... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR 29181... spent fuel storage cask designs. The NRC subsequently issued a final rule on November 21, 2008 (73 FR... 3, 1997 (62 FR 46517), this rule is classified as Compatibility Category ``NRC.'' Compatibility...

  12. 76 FR 9381 - Notice of Availability of Interim Staff Guidance Documents for Spent Fuel Storage Casks

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-02-17

    ... Gordon, Structural Mechanics and Materials Branch, Division of Spent Fuel Storage and Transportation... ISG-23 should be directed to Matthew Gordon, Structural Mechanics and Materials Branch, Division of.... Michele Sampson, Acting Chief, Structural Mechanics and Materials Branch, Division of Spent Fuel...

  13. 76 FR 30980 - Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-27

    ... entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR... COMMISSION Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation... a modification to License No. SNM-2514 at its Humboldt Bay Independent Spent Fuel...

  14. Fracture mapping at the Spent Fuel Test-Climax

    SciTech Connect

    Wilder, D.G.; Yow, J.L. Jr.

    1981-05-01

    Mapping of geologic discontinuities has been done in several phases at the Spent Fuel Test-Climax (SFT-C) in the granitic Climax stock at the Nevada Test Site. Mapping was carried out in the tail drift, access drift, canister drift, heater drifts, instrumentation alcove, and receiving room. The fractures mapped as intersecting a horizontal datum in the canister and heater drifts are shown on one figure. Fracture sketch maps have been compiled as additional figures. Geologic mapping efforts were scheduled around and significantly impacted by the excavation and construction schedules. Several people were involved in the mapping, and over 2500 geologic discontinuities were mapped, including joints, shears, and faults. Some variance between individuals` mapping efforts was noticed, and the effects of various magnetic influences upon a compass were examined. The examination of compass errors improved the credibility of the data. The compass analysis work is explained in Appendix A. Analysis of the fracture data will be presented in a future report.

  15. The effect of coprecipitation in some key spent fuel elements

    NASA Astrophysics Data System (ADS)

    Quiñones, J.; Serrano, J.; Diaz Arocas, P.

    2001-09-01

    Performance assessment (PA) of high-level waste (HLW) repositories needs to know real aqueous concentrations of key radionuclides under repository conditions for assuring the safety of the emplacement. The scarcity of these values under repository conditions leads to the use, in the PA studies, of the solubility of pure phases, which is a conservative assumption. Coprecipitation experiments are a very useful tool for giving realistic solubilities of key radionuclides. In this work, experimental data obtained from spent fuel (SF) and SIMFUEL coprecipitation tests under granite and saline conditions are presented. The experimental concentrations measured for several elements when equilibrium was achieved were much lower than expected considering only the solubility of pure phases. To explain this discrepancy, a tentative approach for modelling these experimental leaching and precipitation results of uranium, plutonium, americium, and strontium taking into account solid solution formations was made.

  16. ALARA studies on spent fuel and waste casks

    SciTech Connect

    Sutherland, S.H.

    1980-04-01

    In this report, some implications of applying the ALARA concept to cask designs for transporting spent fuel, high-level commercial and defense waste, and remote-handled transuranic waste are investigated. The XSDRNPM, one-dimensional radiation transport code, was used to obtain potential shield designs that would yield total dose rates at 1.8 m from the cask surface of 10, 5, and 2 mrem/h. Gamma shields of depleted uranium, lead, and steel were studied. The capacity of the casks was assumed to be 1, 4, or 7 elements or canisters, and the wastes were 1, 3, 5, and 10 years old. Depending on the dose rate, the cask empty weights and lifetime transportation costs were estimated.

  17. Postulated licensing schedule for an independent spent fuel storage installation

    SciTech Connect

    Ludwick, J.D.

    1982-11-01

    A review of licensing requirements, processes, and anticipated actions for independent spent fuel storage installations (ISFSIs) was conducted in order to develop an estimated schedule and sequence of events for licensing a new ISFSI. This estimate will be useful to potential ISFSI owners in planning for the licensing of their facilities. It is concluded that, although many uncertainties exist with respect to such things as legal appeals, about 29 months are estimated to elapse between license application and license issuance for an ISFSI. This estimate is in reasonable agreement with a previous time estimate for licensing an ISFSI, and, taking into account the special circumstances involved, with the actual licensing schedule for the GE-Morris ISFSI. However, individual portions of the licensing schedule from each case studied sometimes vary significantly.

  18. Borated stainless steel application in spent-fuel storage racks

    SciTech Connect

    Smith, R.J.; Loomis, G.W.; Deltete, C.P.

    1992-06-01

    EPRI is continuing to investigate the application of borated stainless steel products within the commercial nuclear power industry through participation in code development and material testing. This effort provides documentation of the material properties of interest in design applications utilizing the borated stainless steel products as structural elements as well as serving as neutron absorbers. The properties of most concern in the design of spent fuel storage racks, shipping casks, and other containment type applications are the materials' ductility, tensile strength, corrosion resistance and resistance to degradation due to radiation and temperature. The data presented in this report indicate that practical designs can be achieved utilizing borated stainless steels and that the materials can be cost effectively applied.

  19. Climax spent fuel dosimetry. Progress report, September 1980-September 1981

    SciTech Connect

    Quam, W.; DeVore, T.

    1981-10-01

    This progress report covers dosimetry work at the Climax Spent Fuel Test Facility up to September 1981. During this time the gamma calibrations were completed, the temperature-induced fading study was completed, the first set of exposed dosimeters was retrieved, and the second set of dosimeters was placed in the field. These were installed in stainless steel tubes located on the inside wall of five canister emplacement holes (0.61 m in diameter), numbers 1, 3, 4, 7, and 11. Hole 3 also had dosimeters in similar stainless steel tubes placed at radii of 0.51 and 0.66 m from the canister centerline. Data obtained from the first exposure (about 270 days in duration) are reported. Significant neutron exposures were measured; in some cases they were sufficiently high that neutron spectra could be calculated.

  20. Training implementation matrix, Spent Nuclear Fuel Project (SNFP)

    SciTech Connect

    EATON, G.L.

    2000-06-08

    This Training Implementation Matrix (TIM) describes how the Spent Nuclear Fuel Project (SNFP) implements the requirements of DOE Order 5480.20A, Personnel Selection, Qualification, and Training Requirements for Reactor and Non-Reactor Nuclear Facilities. The TIM defines the application of the selection, qualification, and training requirements in DOE Order 5480.20A at the SNFP. The TIM also describes the organization, planning, and administration of the SNFP training and qualification program(s) for which DOE Order 5480.20A applies. Also included is suitable justification for exceptions taken to any requirements contained in DOE Order 5480.20A. The goal of the SNFP training and qualification program is to ensure employees are capable of performing their jobs safely and efficiently.

  1. Air Transport of Spent Nuclear Fuel (SNF) Assemblies

    SciTech Connect

    Haire, M.J.; Moses, S.D.; Shapovalov, V.I.; Morenko, A.

    2007-07-01

    Sometimes the only feasible means of shipping research reactor spent nuclear fuel (SNF) among countries is via air transport because of location or political conditions. The International Atomic Energy Agency (IAEA) has established a regulatory framework to certify air transport Type C casks. However, no such cask has been designed, built, tested, and certified. In lieu of an air transport cask, research reactor SNF has been transported using a Type B cask under an exemption with special arrangements for administrative and security controls. This work indicates that it may be feasible to transport commercial power reactor SNF assemblies via air, and that the cost is only about three times that of shipping it by railway. Optimization (i.e., reduction) of this cost factor has yet to be done. (authors)

  2. Evaluation of improvement potential for spent fuel cask handling

    SciTech Connect

    Franklin, A.L.

    1981-02-01

    This report describes the quantitative analysis of opportunities to improve the loading/unloading operations for spent fuel shipping casks. The improvement potential is defined as a reduction in the time for completion or worker exposure for the complete handling operations. Two casks have been chosen as representative of presently available shipping casks. These are the NAC-1/NFS-4 legal weight truck cask and the IF-300 rail cask. The handling operations for each of these casks are broken down into a series of sequential steps. The time for completion and worker exposure is described by a probability density function for each step. These step descriptions are then combined to form a base case description of the total loading/unloading operation. Potential improvement opportunities are evaluated by modifying the appropriate probability density function descriptors then recombining the steps to form a probabilistic description of the modified operation.

  3. Eddy Current Examination of Spent Nuclear Fuel Canister Closure Welds

    SciTech Connect

    Arthur D. Watkins; Dennis C. Kunerth; Timothy R. McJunkin

    2006-04-01

    The National Spent Nuclear Fuel Program (NSNFP) has developed standardized DOE SNF canisters for handling and interim storage of SNF at various DOE sites as well as SNF transport to and SNF handling and disposal at the repository. The final closure weld of the canister will be produced remotely in a hot cell after loading and must meet American Society of Mechanical Engineers (ASME) Section III, Division 3 code requirements thereby requiring volumetric and surface nondestructive evaluation to verify integrity. This paper discusses the use of eddy current testing (ET) to perform surface examination of the completed welds and repair cavities. Descriptions of integrated remote welding/inspection system and how the equipment is intended function will also be discussed.

  4. Homogeneous versus heterogeneous shielding modeling of spent-fuel casks

    SciTech Connect

    Carbajo, J.J.; Lindner, C.N. )

    1992-01-01

    The design of spent-fuel casks for storage and transport requires modeling the cask for criticality, shielding, thermal, and structural analyses. While some parts of the cask are homogeneous, other regions are heterogeneous with different materials intermixed. For simplicity, some of the heterogeneous regions may be modeled as homogeneous. This paper evaluates the effect of homogenizing some regions of a cask on calculating radiation dose rates outside the cask. The dose rate calculations were performed with the one-dimensional discrete ordinates shielding XSDRNPM code coupled with the XSDOSE code and with the three-dimensional QAD-CGGP code. Dose rates were calculated radially at the midplane of the cask at two locations, cask surface and 2.3 m from the radial surface. The last location corresponds to a point 2 m from the lateral sides of a transport railroad car.

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

    SciTech Connect

    Chung, Kiwhan; Beddingfield, David H.; Geist, William H.; Lee, Sang-Yoon

    2012-07-03

    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.

  6. A study on the expulsion of iodine from spent-fuel solutions

    SciTech Connect

    Sakurai, Tsutomu; Takahashi, Akira; Ishikawa, Niroh

    1995-02-01

    During dissolution of spent nuclear fuels, some radioiodine remains in spent-fuel solutions. Its expulsion to dissolver off-gas is important to minimize iodine escape to the environment. In our current work, the iodine remaining in spent-fuel solutions varied from 0 to 10% after dissolution of spent PWR-fuel specimens (approximately 3 g each). The amount remaining probably was dependent upon the dissolution time required. The cause is ascribable to the increased nitrous acid concentration that results from NOx generated during dissolution. The presence of nitrous acid was confirmed spectrophotometrically in an NO-HNO{sub 3} system at 100{degrees}C. Experiments examining NOx concentration versus the quantity of iodine in a simulated spent-fuel solution indicate that iodine (I{minus}) in spent fuels is subjected to the following three reactions: (1) oxidation into I{sub 2} by nitric acid, (2) oxidation into I{sub 2} by nitrous acid arising from NOx, and (3) formation of colloidal iodine (AgI, PdI{sub 2}), the major iodine species in a spent-fuel solution. Reaction (2) competes with reaction (3) to control the quantity of iodine remaining in solution. The following two-step expulsion process to remove iodine from a spent-fuel solution was derived from these experiments: Step One - Heat spent-fuel solutions without NOx sparging. When aged colloidal iodine is present, an excess amount of iodate should be added to the solution. Step Two - Sparge the fuel solution with NOx while heating. Effect of this new method was confirmed by use of a spent PWR-fuel solution.

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

    DOEpatents

    Reichner, P.; Dollard, W.J.

    1991-01-08

    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.

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

    SciTech Connect

    Allen Nellesen

    2005-01-01

    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.

  9. Structural geology report: Spent Fuel Test - Climax Nevada Test Site

    SciTech Connect

    Wilder, D.G.; Yow, J.L. Jr.

    1984-10-01

    We performed underground mapping and core logging in the Climax Stock, a granitic intrusive at the Nevada Test Site, as part of a major field test to determine the feasibility of using granitic or crystalline rock for the underground storage of spent fuel from a nuclear reactor. This mapping and logging identified more than 2500 fractures, over 1500 of which were described in enough detail to allow statistical analyses and orientation studies to be performed. We identified eight joint sets, three major shear sets, and a fault zone within the Spent Fuel Test - Climax (SFT-C) portion of the Stock. Joint sets identified within the SFT-C and elsewhere in the Stock correlated well. The orientations of joint sets identified by other investigators were consistent with our findings, indicating that the joint sets are persistent and have a relatively uniform orientation throughout a major portion of the Stock. The one joint set not seen elsewhere in the Stock is healed and the wall rock is altered, implying that healed joints were not included in the mapping criteria used by other investigators. The shear sets were distinguished from the joint sets by virtue of crushed minerals, continuous clay infilling, and other evidences of shearing, and from faults by the lack of offsetting. Previous investigators working mainly in the Pile Driver Drifts identified two of the shear sets. The third set, being nearly parallel to these Drifts had not been identified previously. The fault zone identified at the far (Receiving Room) end of the project is oriented approximately N45{sup 0}E-75{sup 0}SE, similar to both the Boundary and Shaft Station Faults. We have, therefore, concluded that the Receiving Room Fault is one of a series of normal faults that occur within the Climax Stock and that are possibly related, in both age and genesis, to the Boundary Fault. 52 refs., 26 figs., 11 tabs.

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

    SciTech Connect

    Quiter, Brian; Ludewigt, Bernhard; Ambers, Scott

    2011-06-30

    In nuclear resonance fluorescence (NRF) measurements, resonances are excited by an external photon beam leading to the emission of gamma rays with specific energies that are characteristic of the emitting isotope. NRF promises the unique capability of directly quantifying a specific isotope without the need for unfolding the combined responses of several fissile isotopes as is required in other measurement techniques. We have analyzed the potential of NRF as a non-destructive analysis technique for quantitative measurements of Pu isotopes in spent nuclear fuel (SNF). Given the low concentrations of 239Pu in SNF and its small integrated NRF cross sections, the main challenge in achieving precise and accurate measurements lies in accruing sufficient counting statistics in a reasonable measurement time. Using analytical modeling, and simulations with the radiation transport code MCNPX that has been experimentally tested recently, the backscatter and transmission methods were quantitatively studied for differing photon sources and radiation detector types. Resonant photon count rates and measurement times were estimated for a range of photon source and detection parameters, which were used to determine photon source and gamma-ray detector requirements. The results indicate that systems based on a bremsstrahlung source and present detector technology are not practical for high-precision measurements of 239Pu in SNF. Measurements that achieve the desired uncertainties within hour-long measurements will either require stronger resonances, which may be expressed by other Pu isotopes, or require quasi-monoenergetic photon sources with intensities that are approximately two orders of magnitude higher than those currently being designed or proposed.This work is part of a larger effort sponsored by the Next Generation Safeguards Initiative to develop an integrated instrument, comprised of individual NDA techniques with complementary features, that is fully capable of

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

    SciTech Connect

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

    1986-10-01

    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.

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

    SciTech Connect

    Ham, Y. S.; Sitaraman, S.

    2011-07-01

    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

  13. In-Situ Safeguards Verification of Low Burn-up Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S; Park, I; Kim, J; Ahn, G

    2008-04-16

    A novel in-situ gross defect verification method for light water reactor spent fuel assemblies was developed and investigated by a Monte Carlo study. This particular method is particularly effective for old pressurized water reactor spent fuel assemblies that have natural uranium in their upper fuel zones. Currently there is no method or instrument that does verification of this type of spent fuel assemblies without moving the spent fuel assemblies from their storage positions. The proposed method uses a tiny neutron detector and a detector guiding system to collect neutron signals inside PWR spent fuel assemblies through guide tubes present in PWR assemblies. The data obtained in such a manner are used for gross defect verification of spent fuel assemblies. The method uses 'calibration curves' which show the expected neutron counts inside one of the guide tubes of spent fuel assemblies as a function of fuel burn-up. By examining the measured data in the 'calibration curves', the consistency of the operator's declaration is verified.

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

    SciTech Connect

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

    1982-09-01

    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).

  15. Spent nuclear fuel project recommended reaction rate constants for corrosion of N-Reactor fuel

    SciTech Connect

    Cooper, T.D.

    1998-06-15

    The US Department of Energy (DOE) established the Spent Nuclear Fuel Project (SNF Project) to address safety and environmental concerns associated with deteriorating spent nuclear fuel presently stored in the Hanford Site`s K Basins. The SNF Project has been tasked by the DOE with moving the spent N-Reactor fuel from wet storage to contained dry storage in order to reduce operating costs and environmental hazards. The chemical reactivity of the fuel must be understood at each process step and during long-term dry storage. Normally, the first step would be to measure the N-fuel reactivity before attempting thermal-hydraulic transfer calculations; however, because of the accelerated project schedule, the initial modeling was performed using literature values for uranium reactivity. These literature values were typically found for unirradiated, uncorroded metal. It was fully recognized from the beginning that irradiation and corrosion effects could cause N-fuel to exhibit quite different reactivities than those commonly found in the literature. Even for unirradiated, uncorroded uranium metal, many independent variables affect uranium metal reactivity resulting in a wide scatter of data. Despite this wide reactivity range, it is necessary to choose a defensible model and estimate the reactivity range of the N-fuel until actual reactivity can be established by characterization activities. McGillivray, Ritchie, and Condon developed data and/or models that apply for certain samples over limited temperature ranges and/or reaction conditions (McGillivray 1994, Ritchie 1981 and 1986, and Condon 1983). These models are based upon small data sets and have relatively large correlation coefficients.

  16. Applying fast calorimetry on a spent nuclear fuel calorimeter

    SciTech Connect

    Liljenfeldt, Henrik

    2015-04-15

    Recently at Los Alamos National Laboratory, sophisticated prediction algorithms have been considered for the use of calorimetry for treaty verification. These algorithms aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy. The algorithms have been implemented in MATLAB and applied on existing equilibrium measurements from a spent nuclear fuel calorimeter located at the Swedish nuclear fuel interim storage facility. The results show significant improvements in measurement time in the order of 15 to 50 compared to equilibrium measurements, but cannot predict the heat accurately in less time than the currently used temperature increase method can. This Is both due to uncertainties in the calibration of the method as well as identified design features of the calorimeter that limits the usefulness of equilibrium type measurements. The conclusions of these findings are discussed, and suggestions of both improvements of the current calorimeter as well as what to keep in mind in a new design are given.

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

    SciTech Connect

    Huang, Fan-Hsiung F.

    1997-08-13

    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.

  18. Spent nuclear fuels project characterization data quality objectives strategy

    SciTech Connect

    Lawrence, L.A.; Thornton, T.A.; Redus, K.S.

    1994-12-01

    A strategy is presented for implementation of the Data Quality Objectives (DQO) process to the Spent Nuclear Fuels Project (SNFP) characterization activities. Westinghouse Hanford Company (WHC) and the Pacific Northwest Laboratory (PNL) are teaming in the characterization of the SNF on the Hanford Site and are committed to the DQO process outlined in this strategy. The SNFP characterization activities will collect and evaluate the required data to support project initiatives and decisions related to interim safe storage and the path forward for disposal. The DQO process is the basis for the activity specific SNF characterization requirements, termed the SNF Characterization DQO for that specific activity, which will be issued by the WHC or PNL organization responsible for the specific activity. The Characterization Plan prepared by PNL defines safety, remediation, and disposal issues. The ongoing Defense Nuclear Facility Safety Board (DNFSB) requirement and plans and the fuel storage and disposition options studies provide the need and direction for the activity specific DQO process. The hierarchy of characterization and DQO related documentation requirements is presented in this strategy. The management of the DQO process and the means of documenting the DQO process are described as well as the tailoring of the DQO process to the specific need of the SNFP characterization activities. This strategy will assure stakeholder and project management that the proper data was collected and evaluated to support programmatic decisions.

  19. Safe Advantage on Dry Interim Spent Nuclear Fuel Storage

    SciTech Connect

    Romanato, L.S.

    2008-07-01

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

  20. High Burn-Up Spent Nuclear Fuel Vibration Integrity Study

    SciTech Connect

    Wang, Jy-An John; Wang, Hong; Jiang, Hao; Bevard, Bruce Balkcom; Howard, Rob L; Scaglione, John M

    2015-01-01

    The Oak Ridge National Laboratory (ORNL) has developed the cyclic integrated reversible-bending fatigue tester (CIRFT) approach to successfully demonstrate the controllable fatigue fracture on high burnup (HBU) spent nuclear fuel (SNF) in a normal vibration mode. CIRFT enables examination of the underlying mechanisms of SNF system dynamic performance. Due to the inhomogeneous composite structure of the SNF system, the detailed mechanisms of the pellet-pellet and pellet-clad interactions and the stress concentration effects at the pellet-pellet interface cannot be readily obtained from a CIRFT system measurement. Therefore, finite element analyses (FEAs) are used to translate the global moment-curvature measurement into local stress-strain profiles for further investigation. The major findings of CIRFT on the HBU SNF are as follows: SNF system interface bonding plays an important role in SNF vibration performance. Fuel structure contributes to SNF system stiffness. There are significant variations in stress and curvature of SNF systems during vibration cycles resulting from segment pellets and clad interactions. SNF failure initiates at the pellet-pellet interface region and appears to be spontaneous.

  1. Impact analysis of stainless steel spent fuel canisters

    SciTech Connect

    Aramayo, G.A.; Turner, D.W.

    1998-04-01

    This paper presents the results of the numerical analysis performed to asses the structural integrity of spent nuclear fuel (SNF) stainless steel canisters when subjected to impact loads associated with free gravity drops from heights not exceeding 20 ft. The SNF canisters are to be used for the Shipment of radioactive material from the Oak Ridge National Laboratory (ORNL) Site to the Idaho National Engineering and Environmental Laboratory (INEEL) for storage. The Idaho chemical Processing Plant Fuel Receipt Criteria Questionnaire requires that the vertical drop accidents from two heights be analyze. These heights are those that are considered to be critical at the time of unloading the canisters from the shipping cask. The configurations analyzed include a maximum payload of 90 lbs dropping from heights of 20 and 3 ft. The nominal weight of the canister is 23.3 lbs. The analysis has been performed using finite element methods. Innovative analysis techniques are used to capture the effects of failure and separation of canister components. The structural integrity is evaluated in terms of physical deformation and separation of the canister components that may result from failure of components at selected interfaces.

  2. Parametric study of radiation dose rates from rail and truck spent fuel transport casks

    SciTech Connect

    Parks, C.V.; Hermann, O.W.; Knight, J.R.

    1985-08-01

    Neutron and gamma dose rates from typical rail and truck spent fuel transport casks are reported for a variety of spent PWR fuel sources and cask conditions. The IF 300 rail cask and NLI 1/2 truck cask were selected for use as appropriate cask models. All calculations (cross section preparation, generation of spent fuel source terms, radiation transport calculations, and dose evaluation) were performed using various modules of the SCALE computational system. Conditions or parameters for which there were variations between cases include: detector distance from cask, spent fuel cooling time, the setting of fuel or neutron shielding cavities to either wet or dry, the cobalt content of assembly materials, normal fuel assemblies and consolidated cannisters, the geometry mesh interval size, and the order of the angular quadrature set. 13 refs., 6 figs., 9 tabs.

  3. Status report on the spent fuel test-Climax, Nevada Test Site: A test of dry storage of spent fuel in a deep granite location

    SciTech Connect

    Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

    1982-12-31

    The Spent Fuel Test-Climax (SFT-C) is located at a depth of 420 m in the Climax granite at the Nevada Test Site. The test array contains 11 canistered PWR fuel assemblies, plus associated electrical simulators and electrical heaters. There are nearly 900 channels of thermal, radiation, stress, displacement, and test control instrumentation.

  4. 75 FR 25120 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-07

    ...The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN), NUHOMS[supreg] HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to Certificate of Compliance (CoC) Number 1030. Amendment No. 1 would modify the CoC to add Combustion Engineering 16x16 class fuel......

  5. 76 FR 2277 - List of Approved Spent Fuel Storage Casks: NUHOMS® HD System Revision 1

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-13

    ...The U.S. Nuclear Regulatory Commission (NRC or the Commission) is proposing to amend its spent fuel storage cask regulations by revising the Transnuclear, Inc. (TN) NUHOMS[supreg] HD System listing within the ``List of Approved Spent Fuel Storage Casks'' to include Amendment No. 1 to Certificate of Compliance (CoC) Number 1030. Amendment No. 1 would revise the definitions for Damaged Fuel......

  6. Interim report spent nuclear fuel retrieval system fuel handling development testing

    SciTech Connect

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  7. Waste characteristics of spent nuclear fuel from a pebble bed reactor

    SciTech Connect

    Owen, P.E.

    1999-06-01

    A preliminary comparative assessment is made of the spent fuel characteristics and disposal aspects between a high temperature, gas cooled, reactor with a pebble bed core (PBR) and a pressurized water reactor (PWR). There are three significant differences which impact the disposal characteristics of PBR spent pebble fuel from PWR spent fuel assemblies. Pebble bed fuel has burnup as high as 100,000 MWD(t)/MTHM and thus, there is significantly less activity and decay heat in the fuel when it is disposed. The large amount of graphite in the waste form leads to a low power density and more waste per unit volume than a typical PWR. Pebble Fuel contains a protective layer of Silicon Carbide. The theoretical spacing of waste packages of spent pebble fuel given its unique characteristics as applied to the conditions of Yucca Mountain is of major concern when determining the cost of disposing of the larger volumes of spent pebble fuel. Graphite is a unique waste form and atypical of waste designated for Yucca Mountain. The interactions of silicon carbide with uranium oxide fuel and its implications to long term storage at the repository are examined. There are three primary conclusions to this thesis. First, the area required to store pebble fuel is less than the area required to store light water reactor spent fuel. Second, graphite has excellent characteristics as a waste form. The waste form of the spent pebble fuel is more robust and will perform better than light water reactor fuel at the United States repository at Yucca Mountain. Third, a secondary phase forms between the layers of silicon carbide and the uranium oxide fuel. The secondary phase retards the release of radionuclides to the environment.

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

    SciTech Connect

    Brinton, S.; Kazimi, M.

    2013-07-01

    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)

  9. Disposal of defense spent fuel and HLW from the Idaho Chemical Processing Plant

    SciTech Connect

    Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

    1992-12-01

    Acid high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the US Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage ate the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal.

  10. Preliminary assessment of costs and risks of transporting spent fuel by barge

    SciTech Connect

    Tobin, R.L.; Meshkov, N.K.; Jones, R.H.

    1985-12-01

    The purpose of this study is to analyze the costs and risks associated with transporting spent fuel by barge. The barge movements would be made in combination with rail movements to transport spent fuel from plants to a repository. For the purpose of this analysis, three candidate repository sites are analyzed: Yucca Mountain, Nevada, Deaf Smith, Texas, and Hanford, Washington. This report complements a report prepared by Sandia National Laboratories in 1984 that analyzes the costs and risks of transporting spent fuel by rail and by truck to nine candidate repository sites.

  11. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.

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

    SciTech Connect

    Bailey, W.J.

    1990-02-01

    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.

  13. Multi-pack Disposal Concepts for Spent Fuel (Rev. 0)

    SciTech Connect

    Hadgu, Teklu; Hardin, Ernest; Matteo, Edward N.

    2015-12-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media (Hardin et al., 2012). Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design (CRWMS M&O, 1999). Thermal analysis showed that, if “enclosed” concepts are constrained by peak package/buffer temperature, waste package capacity is limited to 4 PWR assemblies (or 9-BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems (EnergySolution, 2015). This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  14. Multi-Pack Disposal Concepts for Spent Fuel (Revision 1)

    SciTech Connect

    Hardin, Ernest; Matteo, Edward N.; Hadgu, Teklu

    2016-01-01

    At the initiation of the Used Fuel Disposition (UFD) R&D campaign, international geologic disposal programs and past work in the U.S. were surveyed to identify viable disposal concepts for crystalline, clay/shale, and salt host media. Concepts for disposal of commercial spent nuclear fuel (SNF) and high-level waste (HLW) from reprocessing are relatively advanced in countries such as Finland, France, and Sweden. The UFD work quickly showed that these international concepts are all “enclosed,” whereby waste packages are emplaced in direct or close contact with natural or engineered materials . Alternative “open” modes (emplacement tunnels are kept open after emplacement for extended ventilation) have been limited to the Yucca Mountain License Application Design. Thermal analysis showed that if “enclosed” concepts are constrained by peak package/buffer temperature, that waste package capacity is limited to 4 PWR assemblies (or 9 BWR) in all media except salt. This information motivated separate studies: 1) extend the peak temperature tolerance of backfill materials, which is ongoing; and 2) develop small canisters (up to 4-PWR size) that can be grouped in larger multi-pack units for convenience of storage, transportation, and possibly disposal (should the disposal concept permit larger packages). A recent result from the second line of investigation is the Task Order 18 report: Generic Design for Small Standardized Transportation, Aging and Disposal Canister Systems. This report identifies disposal concepts for the small canisters (4-PWR size) drawing heavily on previous work, and for the multi-pack (16-PWR or 36-BWR).

  15. Multi-Detector Analysis System for Spent Nuclear Fuel Characterization

    SciTech Connect

    Reber, Edward Lawrence; Aryaeinejad, Rahmat; Cole, Jerald Donald; Drigert, Mark William; Jewell, James Keith; Egger, Ann Elizabeth; Cordes, Gail Adele

    1999-09-01

    The Spent Nuclear Fuel (SNF) Non-Destructive Analysis (NDA) program at INEEL is developing a system to characterize SNF for fissile mass, radiation source term, and fissile isotopic content. The system is based on the integration of the Fission Assay Tomography System (FATS) and the Gamma-Neutron Analysis Technique (GNAT) developed under programs supported by the DOE Office of Non-proliferation and National Security. Both FATS and GNAT were developed as separate systems to provide information on the location of special nuclear material in weapons configuration (FATS role), and to measure isotopic ratios of fissile material to determine if the material was from a weapon (GNAT role). FATS is capable of not only determining the presence and location of fissile material but also the quantity of fissile material present to within 50%. GNAT determines the ratios of the fissile and fissionable material by coincidence methods that allow the two prompt (immediately) produced fission fragments to be identified. Therefore, from the combination of FATS and GNAT, MDAS is able to measure the fissile material, radiation source term, and fissile isotopics content.

  16. Cermet Spent Nuclear Fuel Casks and Waste Packages

    SciTech Connect

    Forsberg, Charles W.; Dole, Leslie R.

    2007-07-01

    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)

  17. A NOVEL APPROACH TO SPENT FUEL POOL DECOMMISSIONING

    SciTech Connect

    R. L. Demmer

    2011-04-01

    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.

  18. Shipping and storage cask data for spent nuclear fuel

    SciTech Connect

    Johnson, E.R.; Notz, K.J.

    1988-11-01

    This document is a compilation of data on casks used for the storage and/or transport of commercially generated spent fuel in the US based on publicly available information. In using the information contained in the following data sheets, it should be understood that the data have been assembled from published information, which in some instances was not internally consistent. Moreover, it was sometimes necessary to calculate or infer the values of some attributes from available information. Nor was there always a uniform method of reporting the values of some attributes; for example, an outside surface dose of the loaded cask was sometimes reported to be the maximum acceptable by NRC, while in other cases the maximum actual dose rate expected was reported, and in still other cases the expected average dose rate was reported. A summary comparison of the principal attributes of storage and transportable storage casks is provided and a similar comparison for shipping casks is also shown. References to source data are provided on the individual data sheets for each cask.

  19. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    SciTech Connect

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J.; Brey, R.F.; Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  20. Refinishing contamination floors in Spent Nuclear Fuels storage basins

    SciTech Connect

    Huang, F.F.; Moore, F.W.

    1997-07-11

    The floors of the K Basins at the Hanford Site are refinished to make decontamination easier if spills occur as the spent nuclear fuel (SNF) is being unloaded from the basins for shipment to dry storage. Without removing the contaminated existing coating, the basin floors are to be coated with an epoxy coating material selected on the basis of the results of field tests of several paint products. The floor refinishing activities must be reviewed by a management review board to ensure that work can be performed in a controlled manner. Major documents prepared for management board review include a report on maintaining radiation exposure as low as reasonably achievable, a waste management plan, and reports on hazard classification and unreviewed safety questions. To protect personnel working in the radiation zone, Operational Health Physics prescribed the required minimum protective methods and devices in the radiological work permit. Also, industrial hygiene safety must be analyzed to establish respirator requirements for persons working in the basins. The procedure and requirements for the refinishing work are detailed in a work package approved by all safety engineers. After the refinishing work is completed, waste materials generated from the refinishing work must be disposed of according to the waste management plan.

  1. The Suitable Geological Formations for Spent Fuel Disposal in Romania

    SciTech Connect

    Marunteanu, C.; Ionita, G.; Durdun, I.

    2007-07-01

    Using the experience in the field of advanced countries and formerly Romanian program data, ANDRAD, the agency responsible for the disposal of radioactive wastes, started the program for spent fuel disposal in deep geological formations with a documentary analysis at the national scale. The potential geological formations properly characterized elsewhere in the world: salt, clay, volcanic tuff, granite and crystalline rocks,. are all present in Romania. Using general or specific selection criteria, we presently consider the following two areas for candidate geological formations: 1. Clay formations in two areas in the western part of Romania: (1) The Pannonian basin Socodor - Zarand, where the clay formation is 3000 m thick, with many bentonitic strata and undisturbed structure, and (2) The Eocene Red Clay on the Somes River, extending 1200 m below the surface. They both need a large investigation program in order to establish and select the required homogeneous, dry and undisturbed zones at a suitable depth. 2. Old platform green schist formations, low metamorphosed, quartz and feldspar rich rocks, in the Central Dobrogea structural unit, not far from Cernavoda NPP (30 km average distance), 3000 m thick and including many homogeneous, fine granular, undisturbed, up to 300 m thick layers. (authors)

  2. Estimated consequences from severe spent nuclear fuel transportation accidents

    SciTech Connect

    Arnish, J.J.; Monette, F.; LePoire, D.; Biwer, B.M.

    1996-06-01

    The RISKIND software package is used to estimate radiological consequences of severe accident scenarios involving the transportation of spent nuclear fuel. Radiological risks are estimated for both a collective population and a maximally exposed individual based on representative truck and rail cask designs described in the U.S. Nuclear Regulatory Commission (NRC) modal study. The estimate of collective population risk considers all possible environmental pathways, including acute and long-term exposures, and is presented in terms of the 50-y committed effective dose equivalent. Radiological risks to a maximally exposed individual from acute exposure are estimated and presented in terms of the first year and 50-y committed effective dose equivalent. Consequences are estimated for accidents occurring in rural and urban population areas. The modeled pathways include inhalation during initial passing of the radioactive cloud, external exposure from a reduction of the cask shielding, long-term external exposure. from ground deposition, and ingestion from contaminated food (rural only). The major pathways and contributing radionuclides are identified, and the effects of possible mitigative actions are discussed. The cask accident responses and the radionuclide release fractions are modeled as described in the NRC modal study. Estimates of severe accident probabilities are presented for both truck and rail modes of transport. The assumptions made in this study tend to be conservative; however, a set of multiplicative factors are identified that can be applied to estimate more realistic conditions.

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

    SciTech Connect

    Vinson, D.

    2010-07-11

    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.

  4. Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel

    SciTech Connect

    1994-03-25

    One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

  5. Spent Nuclear Fuel Project (SNFP) gas generation from N-Fuel in multi-canister overpacks

    SciTech Connect

    Cooper, T.D.

    1996-08-01

    During the conversion from wet pool storage for spent nuclear fuel at Hanford, gases will be generated from both radiolysis and chemical reactions. The gas generation phenomenon needs to be understood as it applies to safety and design issues,specifically over pressurization of sealed storage containers,and detonation/deflagration of flammable gases. This study provides an initial basis to predict the implications of gas generation on the proposed functional processes for spent nuclear fuel conversion from wet to dry storage. These projections are based upon examination of the history of fuel manufacture at Hanford, irradiation in the reactors, corrosion during wet pool storage, available fuel characterization data and available information from literature. Gas generation via radiolysis and metal corrosion are addressed. The study examines gas generation, the boundary conditions for low medium and high levels of sludge in SNF storage/processing containers. The functional areas examined include: flooded and drained Multi-Canister Overpacks, cold vacuum drying, shipping and staging and long term storage.

  6. Spent-Fuel Test - Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Executive summary of final results

    SciTech Connect

    Patrick, W.C.

    1986-09-02

    This summary volume outlines results that are covered in more detail in the final report of the Spent-Fuel Test - Climate project. The project was conducted between 1978 and 1983 in the granitic Climax stock at the Nevada Test Site. Results indicate that spent fuel can be safely stored for periods of years in this host medium and that nuclear waste so emplaced can be safely retrieved. We also evaluated the effects of heat and radiation (alone and in combination) on emplacement canisters and the surrounding rock mass. Storage of the spent-fuel affected the surrounding rock mass in measurable ways, but did not threaten the stability or safety of the facility at any time.

  7. Comparison of the radiological hazard of thorium and uranium spent fuels from VVER-1000 reactor

    NASA Astrophysics Data System (ADS)

    Frybort, Jan

    2014-11-01

    Thorium fuel is considered as a viable alternative to the uranium fuel used in the current generation of nuclear power plants. Switch from uranium to thorium means a complete change of composition of the spent nuclear fuel produced as a result of the fuel depletion during operation of a reactor. If the Th-U fuel cycle is implemented, production of minor actinides in the spent fuel is negligible. This is favourable for the spent fuel disposal. On the other hand, thorium fuel utilisation is connected with production of 232U, which decays via several alpha decays into a strong gamma emitter 208Tl. Presence of this nuclide might complicate manipulations with the irradiated thorium fuel. Monte-Carlo computation code MCNPX can be used to simulate thorium fuel depletion in a VVER-1000 reactor. The calculated actinide composition will be analysed and dose rate from produced gamma radiation will be calculated. The results will be compared to the reference uranium fuel. Dependence of the dose rate on time of decay after the end of irradiation in the reactor will be analysed. This study will compare the radiological hazard of the spent thorium and uranium fuel handling.

  8. Dosimetry Modeling for Predicting Radiolytic Production at the Spent Fuel - Water Interface

    SciTech Connect

    Miller, William H.; Kline, Amanda J.; Hanson, Brady D.

    2006-04-30

    Modeling of the alpha, beta, and gamma dose from spent fuel as a function of particle size and fuel to water ratio was examined. These doses will be combined with modeling of G values and interactions to determine the concentration of various species formed at the fuel water interface and their affect on dissolution rates.

  9. Application of the electrometallurgical treatment technique to long-term disposition of DOE spent fuel

    SciTech Connect

    Karell, E.J.; Gourishankar, K.V.; McPheeters, C.C.

    1997-09-01

    The DOE inventory of spent nuclear fuel consists of approximately 2700 tonnes heavy metal (MTHM), containing over 100 different fuel types. The current plan for the disposition of this fuel is to condition it for dry storage until it can be placed in a geological repository. However, the variation in the physical condition and chemical composition of DOE spent fuel complicates the task of qualifying the fuel for repository disposal. Each type or category of fuel must be characterized and certified to meet repository disposal criteria, an expensive and time-consuming process. Some of the fuel types contain chemically reactive components (such as metallic sodium), which must be stabilized prior to long-term storage or disposal. Finally, some of the fuel is damaged or declad, and some has already been altered by its present storage environment, making it difficult to qualify that general type of fuel for disposal. The electrometallurgical (EM) treatment technique developed at Argonne National Laboratory (ANL) has the potential to convert many of these spent fuel types into a uniform set of three product streams (uranium metal, metal waste form, ceramic waste form). This treatment would simplify the process of preparing and qualifying these fuels for repository disposal. This paper reviews work done on evaluating the applicability of the EM technique to the treatment of the types of DOE spent fuels currently being stored at the Idaho National Engineering and Environmental Laboratory (INEEL).

  10. International safeguards relevant to geologic disposal of high-level wastes and spent fuels

    SciTech Connect

    Pillay, K.K.S.; Picard, R.R.

    1989-01-01

    Spent fuels from once-through fuel cycles placed in underground repositories have the potential to become attractive targets for diversion and/or theft because of their valuable material content and decreasing radioactivity. The first geologic repository in the US, as currently designed, will contain approximately 500 Mt of plutonium, 60,000 Mt of uranium and a host of other fissile and strategically important elements. This paper identifies some of the international safeguards issues relevant to the various proposed scenarios for disposing of the spent fuel. In the context of the US program for geologic disposal of spent fuels, this paper highlights several issues that should be addressed in the near term by US industries, the Department of Energy, and the Nuclear Regulatory Commission before the geologic repositories for spent fuels become a reality. Based on US spent fuel discharges, an example is presented to illustrate the enormity of the problem of verifying spent fuel inventories. The geologic disposal scenario for high-level wastes originating from defense facilities produced a practicably irrecoverable'' waste form. Therefore, safeguards issues for geologic disposal of high-level waste now in the US are less pressing. 56 refs. , 2 figs.

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

    SciTech Connect

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

    1998-07-01

    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.

  12. 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

    Not Available

    1984-07-01

    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.

  13. Effective thermal conductivity method for predicting spent nuclear fuel cladding temperatures in a dry fill gas

    SciTech Connect

    Bahney, Robert

    1997-12-19

    This paper summarizes the development of a reliable methodology for the prediction of peak spent nuclear fuel cladding temperature within the waste disposal package. The effective thermal conductivity method replaces other older methodologies.

  14. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    SciTech Connect

    MITCHELL, R.M.

    2000-09-28

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

  15. Preoperational Environmental Survey for the Spent Nuclear Fuel (SNF) Project Facilities

    SciTech Connect

    MITCHELL, R.M.

    2000-10-12

    This document represents the report for environmental sampling of soil, vegetation, litter, cryptograms, and small mammals at the Spent Nuclear Fuel Project facilities located in 100 K and 200 East Areas in support of the preoperational environmental survey.

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

    SciTech Connect

    Koji Shirai

    2006-04-01

    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.

  17. FINAL REPORT (PART 1). RADIOLYTIC AND THERMAL PROCESSES RELEVANT TO DRY STORAGE OF SPENT NUCLEAR FUELS

    EPA Science Inventory

    The scientific and engineering demands of the Department of Energy (DOE) Environmental Restoration and Waste Management tasks are enormous. For example, several thousand metric tons of metallic uranium spent nuclear fuel (SNF) remain in water storage awaiting disposition. Of this...

  18. 324 Building spent fuel segments pieces and fragments removal summary report

    SciTech Connect

    SMITH, C L

    2003-01-09

    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).

  19. 78 FR 61401 - Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-03

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing an exemption in response to a request submitted by Entergy Nuclear Operations, Inc. (ENO) on June 20, 2012, for the Big Rock Point (BRP) Independent Spent Fuel Storage Installation...

  20. Hazard & Operability Study for Removal of Spent Nuclear Fuel from the 324 Building

    SciTech Connect

    VAN KEUREN, J.C.

    2002-05-07

    A hazard and operability (HAZOP) study was conducted to examine the hazards associated with the removal of the spent nuclear fuel from the 324 Building. Fifty-nine potentially hazardous conditions were identified.

  1. 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

    Not Available

    1994-06-01

    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.

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

    SciTech Connect

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

    2009-03-01

    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.

  3. Pilot-scale equipment development for pyrochemical treatment of spent oxide fuel.

    SciTech Connect

    Herrmann, S. D.

    1999-06-08

    Fundamental objectives regarding spent nuclear fuel treatment technologies include, first, the effective distribution of spent fuel constituents among product and stable waste forms and, second, the minimization and standardization of waste form types and volumes. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical treatment of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in an uranium product and two stable waste forms, i.e. ceramic and metallic. Engineering efforts are underway at ANL to develop pilot-scale equipment which would precondition irradiated oxide fuel via pyrochemical processing and subsequently allow for electrometallurgical treatment of such non-metallic fuels into standard product and waste forms. This paper highlights the integration of proposed spent oxide fuel treatment with existing electrometallurgical processes. System designs and technical bases for development of pilot-scale oxide reduction equipment are also described.

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

    SciTech Connect

    1996-07-01

    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.

  5. Development of the ACP safeguards neutron counter for PWR spent fuel rods

    NASA Astrophysics Data System (ADS)

    Lee, Tae-Hoon; Menlove, Howard O.; Lee, Sang-Yoon; Kim, Ho-Dong

    2008-04-01

    An advanced neutron multiplicity counter has been developed for measuring spent fuel in the Advanced spent fuel Conditioning Process (ACP) at the Korea Atomic Energy Research Institute (KAERI). The counter uses passive neutron multiplicity counting to measure the 244Cm content in spent fuel. The input to the ACP process is spent fuel from pressurized water reactors (PWRs), and the high intensity of the gamma-ray exposure from spent fuel requires a careful design of the counter to measure the neutrons without gamma-ray interference. The nuclear safeguards for the ACP facility requires the measurement of the spent fuel input to the process and the Cm/Pu ratio for the plutonium mass accounting. This paper describes the first neutron counter that has been used to measure the neutron multiplicity distribution from spent fuel rods. Using multiple samples of PWR spent fuel rod-cuts, the singles (S), doubles (D), and triples (T) rates of the neutron distribution for the 244Cm nuclide were measured and calibration curves were produced. MCNPX code simulations were also performed to obtain the three counting rates and to compare them with the measurement results. The neutron source term was evaluated by using the ORIGEN-ARP code. The results showed systematic difference of 21-24% in the calibration graphs between the measured and simulation results. A possible source of the difference is that the burnup codes have a 244Cm uncertainty greater than ±15% and it would be systematic for all of the calibration samples. The S/D and D/T ratios are almost constant with an increment of the 244Cm mass, and this indicates that the bias is in the 244Cm neutron source calculation using the ORIGEN-ARP source code. The graphs of S/D and D/T ratios show excellent agreement between measurement and MCNPX simulation results.

  6. Worker exposure for at-reactor management of spent nuclear fuel.

    PubMed

    Weck, Philippe F

    2013-09-01

    The radiological impact on workers associated with spent nuclear fuel dry storage operations at reactor sites is discussed. The resulting doses to workers exposed to external radiation include the dose during dry storage system loading, unloading and handling activities, the dose associated with independent spent fuel storage installation (ISFSI) operations, maintenance and surveillance activities, and the dose associated with additional ISFSI construction. Comprehensive dose estimates are reported based on previous radiation surveys. PMID:23564883

  7. COBRA-SFS (Spent-Fuel Storage) thermal-hydraulic analyses of the CASTOR-1C and REA 2023 BWR storage casks containing consolidated spent fuel

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.

    1986-12-01

    Consolidation of spent nuclear fuel rods is being considered as one option for more efficient and compact storage of reactor spent fuel assemblies. In this concept, rods from two disassembled spent fuel assemblies will be consolidated in a space originally intended to store a single unconsolidated assembly. The thermal performance of consolidated fuel rods in dry storage, especially in multiassembly storage systems, is one of the major issues that must be addressed prior to implementation. In this study, Pacific Northwest Laboratory researchers performed thermal-hydraulic analyses for both the REA 2023 cask and the CASTOR-1C cask containing either unconsolidated or consolidated BWR spent fuel assemblies. The objective was to determine the effect of consolidating spent fuel assemblies on the temperature distributions within both types of casks. Two major conclusions resulted from this study. First, a lumping technique (combining rods and flow channels), which reduces the number of computational nodes required to model complex multiassembly geometries, could be used for both unconsolidated and consolidated rods with negligible effect on prediction accuracies. Second, with a relatively high thermal conductivity backfill gas (e.g., helium), the predicted peak fuel rod temperature in a canister of consolidated rods generating the same amount of heat as an unconsolidated assembly is essentially the same as the peak temperature in the unconsolidated assembly. In contrast, with a relatively low thermal conductivity backfill gas (e.g., nitrogen), the opposite is true and the predicted peak temperature in a consolidated canister is significantly higher than in an unconsolidated assembly. Therefore, when rods are consolidated, selection of the backfill gas is important in maintaining peak rod temperatures below allowable values for rods with relatively high decay heat generation rates.

  8. Impacts of a high-burnup spent fuel on a geological disposal system design

    SciTech Connect

    Cho, D.K.; Lee, Y.; Lee, J.Y.; Choi, H.J.; Choi, J.W.

    2007-07-01

    The influence of a burnup increase of a spent nuclear fuel on a deep geological disposal system was evaluated in this study. First, the impact of a burnup increase on each aspect related to thermal and nuclear safety concerns was quantified. And then, the tunnel length, excavation volume, and the raw materials for a cast insert, copper, bentonite, and backfill needed to constitute a disposal system were comprehensively analyzed based on the spent fuel inventory to generate 1 Terawatt-year (TWa), to establish the overall effects and consequences on a geological disposal. As a result, impact of a burnup increase on the criticality safety and radiation shielding was shown to be negligible. The disposal area, however, is considerably affected because of a higher thermal load. And, it is reasonable to use a canister such as the Korean Reference Disposal Canister (KDC-1) containing 4 spent fuels up to 50 GWD/MtU, and to use a canister containing 3 spent fuels beyond 50 GWD/MtU. Although a considerable increased, 33 % in the tunnel length and 30 % in the excavation volume, was observed as the burnup increases from 50 to 60 GWD/MtU, because a decrease in the canister needs can offset an increase in the excavation volume, it can be concluded that a burnup increase of a spent fuel is not a critical concern for a geological disposal of a spent fuel. (authors)

  9. INTERIM STORAGE AND LONG TERM DISPOSAL OF RESEARCH REACTOR SPENT FUEL

    SciTech Connect

    Vinson, D

    2006-08-22

    Aluminum clad research reactor spent nuclear fuel (SNF) is currently being consolidated in wet storage basins (pools). Approximately 20 metric tons (heavy metal) of aluminum-based spent nuclear fuel (Al-SNF) is being consolidated for treatment, packaging, interim storage, and preparation for ultimate disposal in a geologic repository. The storage and disposal of Al-SNF are subject to requirements that provide for safety and acceptable radionuclide release. The options studied for interim storage of SNF include wet storage and dry storage. Two options have also been studied to develop the technical basis for the qualification and repository disposal of aluminum spent fuel. The two options studied include Direct Disposal and Melt-Dilute treatment. The implementation of these options present relative benefits and challenges. Both the Direct Disposal and the Melt-Dilute treatment options have been developed and their technical viability assessed. Adaptation of the melt-dilute technology for the treatment of spent fuel offers the benefits of converting the spent fuel into a proliferation resistant form and/or significantly reducing the volume of the spent fuel. A Mobile Melt-Dilute system concept has emerged to realize these benefits and a prototype system developed. The application of the melt-dilute technology for the treatment of legacy nuclear materials has been evaluated and also offers the promise for the safe disposal of these materials.

  10. Site Specific Analyses of a Spent Nuclear Fuel Transportation Accident

    SciTech Connect

    Biwer, B. M.; Chen, S. Y.

    2003-02-24

    The number of spent nuclear fuel (SNF) shipments is expected to increase significantly during the time period that the United States' inventory of SNF is sent to a final disposal site. Prior work estimated that the highest accident risks of a SNF shipping campaign to the proposed geologic repository at Yucca Mountain were in the corridor states, such as Illinois. The largest potential human health impacts would be expected to occur in areas with high population densities such as urban settings. Thus, our current study examined the human health impacts from the most plausible severe SNF transportation accidents in the Chicago metropolitan area. The RISKIND 2.0 program was used to model site-specific data for an area where the largest impacts might occur. The results have shown that the radiological human health consequences of a severe SNF rail transportation accident on average might be similar to one year of exposure to natural background radiation for those persons living a nd working in the most affected areas downwind of the actual accident location. For maximally exposed individuals, an exposure similar to about two years of exposure to natural background radiation was estimated. In addition to the accident probabilities being very low (approximately 1 chance in 10,000 or less during the entire shipping campaign), the actual human health impacts are expected to be lower if any of the accidents considered did occur, because the results are dependent on the specific location and weather conditions, such as wind speed and direction, that were selected to maximize the results. Also, comparison of the results of longer duration accident scenarios against U.S. Environmental Protection Agency guidelines was made to demonstrate the usefulness of this site-specific analysis for emergency planning purposes.

  11. INEL integrated spent nuclear fuel consolidation task team report

    SciTech Connect

    Henry, R.N.; Clark, J.H.; Chipman, N.A.

    1994-09-12

    This document describes a draft plan and schedule to consolidate spent nuclear fuel (SNF) and special nuclear material (SNW) from aging storage facilities throughout the Idaho National Engineering Laboratory (INEL) to the Idaho Chemical Processing Plant (ICPP) in a safe, cost-effective, and expedient manner. A fully integrated and resource-loaded schedule was developed to achieve consolidation as soon as possible. All of the INEL SNF and SNM management task, projects, and related activities from fiscal year 1994 to the end of the consolidation period are logic-tied and integrated with each other. The schedule and plan are presented to initiate discussion of their implementation, which is expected to generate alternate concepts that can be evaluated using the methodology described in this report. Three perturbations to consolidating SNF as soon as possible are also explored. If the schedule is executed as proposed, the new and on-going consolidation activities will require about 6 years to complete and about $25.3M of additional funding. Reduced annual operating costs are expected to recover the additional investment in about 6.4 years. The total consolidation program as proposed will cost about $66.8M and require about 6 years to recover via reduced operating costs from retired SNF/SNM storage facilities. Detailed schedules and cost estimates for the Test Reactor Area Materials Test Reactor canal transfers are included as an example of the level of detail that is typical of the entire schedule (see Appendix D). The remaining work packages for each of the INEL SNF consolidation transfers are summarized in this document. Detailed cost and resource information is available upon request for any of the SNF consolidation transfers.

  12. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    NASA Astrophysics Data System (ADS)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

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

    SciTech Connect

    Hazelton, R.F.

    1987-09-01

    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. In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

    SciTech Connect

    Gauld, Ian C.; Hu, Jianwei; De Baere, P.; Vaccaro, S.; Schwalbach, P.; Liljenfeldt, Henrik; Tobin, Stephen

    2015-01-01

    Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel

  15. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    SciTech Connect

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the world’s first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  16. LWR fuel assembly designs for the transmutation of LWR Spent Fuel TRU with FCM and UO{sub 2}-ThO{sub 2} Fuels

    SciTech Connect

    Bae, G.; Hong, S. G.

    2013-07-01

    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)

  17. Status report on the Spent-Fuel Test-Climax, Nevada Test Site: a test of dry storage of spent fuel in a deep granite location

    SciTech Connect

    Ramspott, L.D.; Ballou, L.B.; Patrick, W.C.

    1982-03-01

    The Spent Fuel Test-Climax (SFT-C) is located at a depth of 420 m in the Climax granite at the Nevada Test Site. The test array contains 11 canistered PWR fuel assemblies, plus associated electrical simulators and electrical heaters. There are nearly 900 channels of thermal, radiation, stress, displacement, and test control instrumentation. This paper is a general status report on the test, which started in May 1980.

  18. Development of the GA-4 and GA-9 legal weight spent fuel casks

    SciTech Connect

    Grenier, R.M.; Meyer, R.J.; Mings, W.J.

    1992-09-01

    GA is nearing the completion of the final design of two legal weight truck spent fuel shipping casks, the GA-4 Cask for PWR fuel and the GA-9 Cask for BWR fuel. GA is developing the casks under contract to the US Department of Energy (DOE) Field Office, Idaho, as part of the Office of Civilian Radioactive Waste Management (OCRWM) Cask Systems Development Program (CSDP). The casks will transport intact spent fuel assemblies fro commercial nuclear reactors sites to a monitored retrievable storage facility or a permanent repository. The DOE initiated the Cask Systems Development Program in response to the Nuclear Waste Policy Act of 1982 which made DOE responsible for managing the program for permanent disposal of spent nuclear fuel and high-level waste. This paper describes developmental and design verification testing programs, and the present status of the GA-4 and GA-9 Cask designs.

  19. Analysis of spent fuel assay with a lead slowing down spectrometer

    SciTech Connect

    Gavron, Victor I; Smith, L Eric; Ressler, Jennifer J

    2008-01-01

    Assay of fissile materials in spent fuel that are produced or depleted during the operation of a reactor, is of paramount importance to nuclear materials accounting, verification of the reactor operation history, as well as for criticality considerations for storage. In order to prevent future proliferation following the spread of nuclear energy, we must develop accurate methods to assay large quantities of nuclear fuels. We analyze the potential of using a Lead Slowing Down Spectrometer for assaying spent fuel. We conclude that it is possible to design a system that will provide around 1% statistical precision in the determination of the {sup 239}Pu, {sup 241}Pu and {sup 235}U concentrations in a PWR spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and a system of {sup 238}U threshold fission detectors. Pending further analysis of systematic errors, it is possible that missing pins can be detected, as can asymmetry in the fuel bundle.

  20. Analysis of spent fuel assay with a lead slowing down spectrometer

    SciTech Connect

    Gavron, Victor I; Smith, L. Eric; Ressler, Jennifer J

    2010-10-29

    Assay of fissile materials in spent fuel that are produced or depleted during the operation of a reactor, is of paramount importance to nuclear materials accounting, verification of the reactor operation history, as well as for criticality considerations for storage. In order to prevent future proliferation following the spread of nuclear energy, we must develop accurate methods to assay large quantities of nuclear fuels. We analyze the potential of using a Lead Slowing Down Spectrometer for assaying spent fuel. We conclude that it is possible to design a system that will provide around 1% statistical precision in the determination of the {sup 239}Pu, {sup 241}Pu and {sup 235}U concentrations in a PWR spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and a system of {sup 238}U threshold fission detectors. Pending further analysis of systematic errors, it is possible that missing pins can be detected, as can asymmetry in the fuel bundle.

  1. Analysis of near-term spent fuel transportation hardware requirements and transportation costs

    SciTech Connect

    Daling, P.M.; Engel, R.L.

    1983-01-01

    A computer model was developed to quantify the transportation hardware requirements and transportation costs associated with shipping spent fuel in the commercial nucler fuel cycle in the near future. Results from this study indicate that alternative spent fuel shipping systems (consolidated or disassembled fuel elements and new casks designed for older fuel) will significantly reduce the transportation hardware requirements and costs for shipping spent fuel in the commercial nuclear fuel cycle, if there is no significant change in their operating/handling characteristics. It was also found that a more modest cost reduction results from increasing the fraction of spent fuel shipped by truck from 25% to 50%. Larger transportation cost reductions could be realized with further increases in the truck shipping fraction. Using the given set of assumptions, it was found that the existing spent fuel cask fleet size is generally adequate to perform the needed transportation services until a fuel reprocessing plant (FRP) begins to receive fuel (assumed in 1987). Once the FRP opens, up to 7 additional truck systems and 16 additional rail systems are required at the reference truck shipping fraction of 25%. For the 50% truck shipping fraction, 17 additional truck systems and 9 additional rail systems are required. If consolidated fuel only is shipped (25% by truck), 5 additional rail casks are required and the current truck cask fleet is more than adequate until at least 1995. Changes in assumptions could affect the results. Transportation costs for a federal interim storage program could total about $25M if the FRP begins receiving fuel in 1987 or about $95M if the FRP is delayed until 1989. This is due to an increased utilization of federal interim storage facility from 350 MTU for the reference scenario to about 750 MTU if reprocessing is delayed by two years.

  2. COBRA-SFS CYCLE 3. Thermal Hydraulic Analysis of Spent Fuel Casks

    SciTech Connect

    Rector, D.R.; Cuta, J.M.; Enderlin, C.W.

    1995-09-01

    COBRA-SFS (Spent Fuel Storage) is a code for thermal-hydraulic analysis of multi-assembly spent fuel storage and transportation systems. It uses a lumped parameter finite difference approach to predict flow and temperature distributions in spent fuel storage systems and fuel assemblies, under forced and natural convection heat transfer conditions. Derived from the COBRA family of codes, which have been extensively evaluated against in-pile and out-of-pile data, COBRA-SFS retains all the important features of the COBRA codes for single phase fluid analysis, and extends the range application to include problems with two-dimensional radiative and three-dimensional conductive heat transfer. COBRA-SFS has been used to analyze various single- and multi-assembly spent fuel storage systems containing unconsolidated and consolidated fuel rods, with a variety of fill media, including air, helium and vacuum. Cycle 0 of COBRA-SFS was released in 1986. Subsequent applications of the code led to development of additional capabilities, which resulted in the release of Cycle 1 in February 1989. Since then, the code has undergone an independent technical review as part of a submittal to the Nuclear Regulatory Commission for a generic license to apply the code to spent fuel storage system analysis. Modifications and improvements to the code have been combined to form Cycle 2. Cycle 3., the newest version of COBRA-SFS, has been validated and verified for transient applications, such as a storage cask thermal response to a pool fire.

  3. Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

    SciTech Connect

    Hu, Jianwei; Gauld, Ian C.

    2014-12-01

    The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried out to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.

  4. Impact of Nuclear Data Uncertainties on Calculated Spent Fuel Nuclide Inventories and Advanced NDA Instrument Response

    DOE PAGESBeta

    Hu, Jianwei; Gauld, Ian C.

    2014-12-01

    The U.S. Department of Energy’s Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project is nearing the final phase of developing several advanced nondestructive assay (NDA) instruments designed to measure spent nuclear fuel assemblies for the purpose of improving nuclear safeguards. Current efforts are focusing on calibrating several of these instruments with spent fuel assemblies at two international spent fuel facilities. Modelling and simulation is expected to play an important role in predicting nuclide compositions, neutron and gamma source terms, and instrument responses in order to inform the instrument calibration procedures. As part of NGSI-SF project, this work was carried outmore » to assess the impacts of uncertainties in the nuclear data used in the calculations of spent fuel content, radiation emissions and instrument responses. Nuclear data is an essential part of nuclear fuel burnup and decay codes and nuclear transport codes. Such codes are routinely used for analysis of spent fuel and NDA safeguards instruments. Hence, the uncertainties existing in the nuclear data used in these codes affect the accuracies of such analysis. In addition, nuclear data uncertainties represent the limiting (smallest) uncertainties that can be expected from nuclear code predictions, and therefore define the highest attainable accuracy of the NDA instrument. This work studies the impacts of nuclear data uncertainties on calculated spent fuel nuclide inventories and the associated NDA instrument response. Recently developed methods within the SCALE code system are applied in this study. The Californium Interrogation with Prompt Neutron instrument was selected to illustrate the impact of these uncertainties on NDA instrument response.« less

  5. Safeguards-by-Design: Guidance for Independent Spent Fuel Dry Storage Installations (ISFSI)

    SciTech Connect

    Trond Bjornard; Philip C. Durst

    2012-05-01

    This document summarizes the requirements and best practices for implementing international nuclear safeguards at independent spent fuel storage installations (ISFSIs), also known as Away-from- Reactor (AFR) storage facilities. These installations may provide wet or dry storage of spent fuel, although the safeguards guidance herein focuses on dry storage facilities. In principle, the safeguards guidance applies to both wet and dry storage. The reason for focusing on dry independent spent fuel storage installations is that this is one of the fastest growing nuclear installations worldwide. Independent spent fuel storage installations are typically outside of the safeguards nuclear material balance area (MBA) of the reactor. They may be located on the reactor site, but are generally considered by the International Atomic Energy Agency (IAEA) and the State Regulator/SSAC to be a separate facility. The need for this guidance is becoming increasingly urgent as more and more nuclear power plants move their spent fuel from resident spent fuel ponds to independent spent fuel storage installations. The safeguards requirements and best practices described herein are also relevant to the design and construction of regional independent spent fuel storage installations that nuclear power plant operators are starting to consider in the absence of a national long-term geological spent fuel repository. The following document has been prepared in support of two of the three foundational pillars for implementing Safeguards-by-Design (SBD). These are: i) defining the relevant safeguards requirements, and ii) defining the best practices for meeting the requirements. This document was prepared with the design of the latest independent dry spent fuel storage installations in mind and was prepared specifically as an aid for designers of commercial nuclear facilities to help them understand the relevant international requirements that follow from a country’s safeguards agreement with

  6. Compton suppressed LaBr3 detection system for use in nondestructive spent fuel assay

    NASA Astrophysics Data System (ADS)

    Bender, S.; Heidrich, B.; Ünlü, K.

    2015-06-01

    Current methods for safeguarding and accounting for spent nuclear fuel in reprocessing facilities are extremely resource and time intensive. The incorporation of autonomous passive gamma-ray detectors into the procedure could make the process significantly less burdensome. In measured gamma-ray spectra from spent nuclear fuel, the Compton continuum from dominant fission product photopeaks obscure the lower energy lines from other isotopes. The application of Compton suppression to gamma-ray measurements of spent fuel may reduce this effect and allow other less intense, lower energy peaks to be detected, potentially improving the accuracy of multivariate analysis algorithms. Compton suppressed spectroscopic measurements of spent nuclear fuel using HPGe, LaBr3, and NaI(Tl) primary detectors were performed. Irradiated fuel was measured in two configurations: as intact fuel elements viewed through a collimator and as feed solutions in a laboratory to simulate the measurement of a dissolved process stream. These two configurations allowed the direct assessment and quantification of the differences in measured gamma-ray spectra from the application of Compton suppression. In the first configuration, several irradiated fuel elements of varying cooling times from the Penn State Breazeale Reactor spent fuel inventory were measured using the three collimated Compton suppression systems. In the second geometry, Compton suppressed measurements of two samples of Approved Test Material commercial fuel elements were recorded inside the guard detector annulus to simulate the siphoning of small quantities from the main process stream for long dwell measurement periods. Compton suppression was found to improve measured gamma-ray spectra of spent fuel for multivariate analysis by notably lowering the Compton continuum from dominant photopeaks such as 137Cs and 140La, due to scattered interactions in the detector, which allowed more spectral features to be resolved. There was a

  7. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    SciTech Connect

    Wang, Jy-An John; Jiang, Hao

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

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

    SciTech Connect

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

    1992-04-01

    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.

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

    SciTech Connect

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

    1992-04-01

    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.

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

    DOEpatents

    Reichner, Philip; Dollard, Walter J.

    1991-01-01

    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).

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

    SciTech Connect

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

    2013-07-01

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

  12. Analysis of subcritical experiments using fresh and spent research reactor fuel assemblies

    NASA Astrophysics Data System (ADS)

    Zino, John Frederick

    1999-11-01

    This research investigated the concepts associated with crediting the burnup of spent nuclear fuel assemblies for the purposes of criticality safety. To accomplish this, a collaborative experimental research program was undertaken between Westinghouse, the University of Missouri Research Reactor (MURR) facility and Oak Ridge National Laboratory (ORNL). The purpose of the program was to characterize the subcritical behavior of a small array of fresh and spent MURR fuel assemblies using the 252Cf Source-driven noise technique. An aluminum test rig was built which was capable of holding up to four, highly enriched (93.15 wt.% 235U) MURR fuel assemblies in a 2 x 2 array. The rig was outfitted with one source and four detector drywells which allowed researchers to perform active neutron noise measurements on the array of fuel assemblies. The 1 atmosphere gas 3He neutron detectors used to perform the measurements were quenched with CF4 gas to allow improved discrimination of the neutron signals in the very high gamma-ray fields associated with spent fuel (˜8000 R/hr). In addition, the detector drywells were outfitted with 1″ lead collars to provide additional gamma-ray shielding from the spent fuel. Reactivity changes were induced in the subcritical lattice by replacing individual fresh assemblies (in a 4-assembly array) with spent assemblies of known, maximum burnup (143 Mw-D). The absolute and relative measured reactivity changes were then compared to those predicted by three-dimensional Monte Carlo calculations. The purpose of these comparisons was to investigate the accuracy of modern transport theory depletion calculations to accurately simulate the reactivity effects of burnup in spent nuclear fuel. A total of seven subcritical measurements were performed at the MURR reactor facility on July 20th and 27th, 1998. These measurements generated several estimates of prompt neutron decay constants (alpha) and ratios of spectral densities through frequency correlations

  13. Spent nuclear fuel project high-level information management plan

    SciTech Connect

    Main, G.C.

    1996-09-13

    This document presents the results of the Spent Nuclear Fuel Project (SNFP) Information Management Planning Project (IMPP), a short-term project that identified information management (IM) issues and opportunities within the SNFP and outlined a high-level plan to address them. This high-level plan for the SNMFP IM focuses on specific examples from within the SNFP. The plan`s recommendations can be characterized in several ways. Some recommendations address specific challenges that the SNFP faces. Others form the basis for making smooth transitions in several important IM areas. Still others identify areas where further study and planning are indicated. The team`s knowledge of developments in the IM industry and at the Hanford Site were crucial in deciding where to recommend that the SNFP act and where they should wait for Site plans to be made. Because of the fast pace of the SNFP and demands on SNFP staff, input and interaction were primarily between the IMPP team and members of the SNFP Information Management Steering Committee (IMSC). Key input to the IMPP came from a workshop where IMSC members and their delegates developed a set of draft IM principles. These principles, described in Section 2, became the foundation for the recommendations found in the transition plan outlined in Section 5. Availability of SNFP staff was limited, so project documents were used as a basis for much of the work. The team, realizing that the status of the project and the environment are continually changing, tried to keep abreast of major developments since those documents were generated. To the extent possible, the information contained in this document is current as of the end of fiscal year (FY) 1995. Programs and organizations on the Hanford Site as a whole are trying to maximize their return on IM investments. They are coordinating IM activities and trying to leverage existing capabilities. However, the SNFP cannot just rely on Sitewide activities to meet its IM requirements

  14. Concept of advanced spent fuel reprocessing based on ion exchange

    SciTech Connect

    Suzuki, Tatsuya; Takahashi, Kazuyuki; Nogami, Masanobu; Nomura, Masao; Fujii, Yasuhiko; Ozawa, Masaki |; Koyama, Shinichi; Mimura, Hitosi; Fujita, Reiko

    2007-07-01

    . Furthermore, the ion exchange is appropriate for multi-element mutual separation rather than single element extraction. In the future, ion exchange reprocessing would be expected to be the comprehensive separation process for spent fuels to recover precious and usable elements and to reduce the amount of wastes. (authors)

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

    SciTech Connect

    Brucher, Wenzel; Koch, Wolfgang; Pretzsch, Gunter Guido; Loiseau, Olivier; Mo, Tin; Billone, Michael C.; Autrusson, Bruno A.; Young, F. I.; Coats, Richard Lee; Burtseva, Tatiana; Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver; Thompson, Nancy Slater; Hibbs, Russell S.; Gregson, Michael Warren; Lange, Florentin; Molecke, Martin Alan; Tsai, Han-Chung

    2005-07-01

    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

  16. Severe accidents in spent fuel pools in support of generic safety, Issue 82

    SciTech Connect

    Sailor, V.L.; Perkins, K.R.; Weeks, J.R.; Connell, H.R.

    1987-07-01

    This investigation provides an assessment of the likelihood and consequences of a severe accident in a spent fuel storage pool - the complete draining of the pool. Potential mechanisms and conditions for failure of the spent fuel, and the subsequent release of the fission products, are identified. Two older PWR and BWR spent fuel storage pool designs are considered based on a preliminary screening study which tried to identify vulnerabilities. Internal and external events and accidents are assessed. Conditions which could lead to failure of the spent fuel Zircaloy cladding as a result of cladding rupture or as a result of a self-sustaining oxidation reaction are presented. Propagation of a cladding fire to older stored fuel assemblies is evaluated. Spent fuel pool fission product inventory is estimated and the releases and consequences for the various cladding scenarios are provided. Possible preventive or mitigative measures are qualitatively evaluated. The uncertainties in the risk estimate are large, and areas where additional evaluations are needed to reduce uncertainty are identified.

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

    SciTech Connect

    N /A

    2000-04-14

    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.

  18. Spent fuel waste form characteristics: Grain and fragment size statistical dependence for dissolution response

    SciTech Connect

    Stout, R.B.; Leider, H.; Weed, H.; Nguyen, S.; McKenzie, W.; Prussin, S.; Wilson, C.N.; Gray, W.J.

    1991-04-01

    The Yucca Mountain Project of the US Department of Energy is investigating the suitability of the unsaturated zone at Yucca Mountain, NV, for a high-level nuclear waste repository. All of the nuclear waste will be enclosed in a container package. Most of the nuclear waste will be in the form of fractured UO{sub 2} spent fuel pellets in Zircaloy-clad rods from electric power reactors. If failure of both the container and its enclosed clad rods occurs, then the fragments of the fractured UO{sub 2} spent fuel will be exposed to their surroundings. Even though the surroundings are an unsaturated zone, a possibility of water transport exists, and consequently, UO{sub 2} spent fuel dissolution may occur. A repository requirement imposes a limit on the nuclide release per year during a 10,000 year period; thus the short term dissolution response from fragmented fuel pellet surfaces in any given year must be understood. This requirement necessitates that both experimental and analytical activities be directed toward predicting the relatively short term dissolution response of UO{sub 2} spent fuel. The short term dissolution response involves gap nuclides, grain boundary nuclides, and grain volume nuclides. Analytical expressions are developed that describe the combined geometrical influences of grain boundary nuclides and grain volume nuclides on the dissolution rate of spent fuel. 7 refs., 1 fig.

  19. 10 CFR 171.15 - Annual fees: Reactor licenses and independent spent fuel storage licenses.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Annual fees: Reactor licenses and independent spent fuel storage licenses. 171.15 Section 171.15 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) ANNUAL FEES FOR REACTOR LICENSES AND FUEL CYCLE LICENSES AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF COMPLIANCE, REGISTRATIONS, AND...

  20. 75 FR 42339 - List of Approved Spent Fuel Storage Casks: NAC-MPC System, Revision 6

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-21

    ... closure ring for redundant closure into the Transportable Storage Canister (TSC) design; modification of the TSC and basket design to accommodate up to 68 La Crosse Boiling Water Reactor spent fuel... pattern)) that may contain undamaged Exxon fuel assemblies and damaged Exxon and Allis Chalmers...

  1. OECD NEA Benchmark Database of Spent Nuclear Fuel Isotopic Compositions for World Reactor Designs

    SciTech Connect

    Gauld, Ian C; Sly, Nicholas C; Michel-Sendis, Franco

    2014-01-01

    Experimental data on the isotopic concentrations in irradiated nuclear fuel represent one of the primary methods for validating computational methods and nuclear data used for reactor and spent fuel depletion simulations that support nuclear fuel cycle safety and safeguards programs. Measurement data have previously not been available to users in a centralized or searchable format, and the majority of accessible information has been, for the most part, limited to light-water-reactor designs. This paper describes a recent initiative to compile spent fuel benchmark data for additional reactor designs used throughout the world that can be used to validate computer model simulations that support nuclear energy and nuclear safeguards missions. Experimental benchmark data have been expanded to include VVER-440, VVER-1000, RBMK, graphite moderated MAGNOX, gas cooled AGR, and several heavy-water moderated CANDU reactor designs. Additional experimental data for pressurized light water and boiling water reactor fuels has also been compiled for modern assembly designs and more extensive isotopic measurements. These data are being compiled and uploaded to a recently revised structured and searchable database, SFCOMPO, to provide the nuclear analysis community with a centrally-accessible resource of spent fuel compositions that can be used to benchmark computer codes, models, and nuclear data. The current version of SFCOMPO contains data for eight reactor designs, 20 fuel assembly designs, more than 550 spent fuel samples, and measured isotopic data for about 80 nuclides.

  2. Extended Storage for Research and Test Reactor Spent Fuel for 2006 and Beyond

    SciTech Connect

    Hurt, William Lon; Moore, K.M.; Shaber, Eric Lee; Mizia, Ronald Eugene

    1999-10-01

    This paper will examine issues associated with extended storage of a variety of spent nuclear fuels. Recent experiences at the Idaho National Engineering and Environmental Laboratory and Hanford sites will be described. Particular attention will be given to storage of damaged or degraded fuel. The first section will address a survey of corrosion experience regarding wet storage of spent nuclear fuel. The second section will examine issues associated with movement from wet to dry storage. This paper also examines technology development needs to support storage and ultimate disposition.

  3. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    SciTech Connect

    DeHart, M.D.

    1999-08-01

    Utilization of burnup credit in criticality safety analysis for long-term disposal of spent nuclear fuel allows improved design efficiency and reduced cost due to the large mass of fissile material that will be present in the repository. Burnup-credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents (in terms of criticality potential), followed by criticality calculations to assess the value of the effective neutron multiplication factor (k(sub)eff) for the a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models.

  4. Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies

    SciTech Connect

    Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

    1980-05-01

    Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study.

  5. A COMPARISON OF CHALLENGES ASSOCIATED WITH SLUDGE REMOVAL & TREATMENT & DISPOSAL AT SEVERAL SPENT FUEL STORAGE LOCATIONS

    SciTech Connect

    PERES, M.W.

    2007-01-09

    Challenges associated with the materials that remain in spent fuel storage pools are emerging as countries deal with issues related to storing and cleaning up nuclear fuel left over from weapons production. The K Basins at the Department of Energy's site at Hanford in southeastern Washington State are an example. Years of corrosion products and piles of discarded debris are intermingled in the bottom of these two pools that stored more 2,100 metric tons (2,300 tons) of spent fuel. Difficult, costly projects are underway to remove radioactive material from the K Basins. Similar challenges exist at other locations around the globe. This paper compares the challenges of handling and treating radioactive sludge at several locations storing spent nuclear fuel.

  6. Achieving increased spent fuel storage capacity at the High Flux Isotope Reactor (HFIR)

    SciTech Connect

    Cook, D.H.; Chang, S.J.; Dabs, R.D.; Freels, J.D.; Morgan, K.A.; Rothrock, R.B.; Griess, J.C.

    1994-12-31

    The HFIR facility was originally designed to store approximately 25 spent cores, sufficient to allow for operational contingencies and for cooling prior to off-site shipment for reprocessing. The original capacity has now been increased to 60 positions, of which 53 are currently filled (September 1994). Additional spent cores are produced at a rate of about 10 or 11 per year. Continued HFIR operation, therefore, depends on a significant near-term expansion of the pool storage capacity, as well as on a future capability of reprocessing or other storage alternatives once the practical capacity of the pool is reached. To store the much larger inventory of spent fuel that may remain on-site under various future scenarios, the pool capacity is being increased in a phased manner through installation of a new multi-tier spent fuel rack design for higher density storage. A total of 143 positions was used for this paper as the maximum practical pool capacity without impacting operations; however, greater ultimate capacities were addressed in the supporting analyses and approval documents. This paper addresses issues related to the pool storage expansion including (1) seismic effects on the three-tier storage arrays, (2) thermal performance of the new arrays, (3) spent fuel cladding corrosion concerns related to the longer period of pool storage, and (4) impacts of increased spent fuel inventory on the pool water quality, water treatment systems, and LLLW volume.

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

    SciTech Connect

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

    2009-01-01

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

  8. HEDL contribution to Office of Nuclear Waste Isolation. July through September quarterly report: spent fuel characterization equipment

    SciTech Connect

    Cash, R.J.

    1980-10-15

    Progress reports are presented for the following areas of study: spent fuel characterization equipment; spent fuel characterization; and spent fuel/package performance. Some of the highlights are: calorimetry was performed on spent fuel assemblies D15 and D22 which were destined for packaging and emplaced into the Climax-Spent Fuel Test (C-SFT); review and analysis of the destructive examination data on five fuel rods pulled from the C-SFT assemblies is continuing; under the disposal condition loads assumed in the structural analyses. The spent fuel cladding has very high mechanical integrity; a program plan to develop a spent fuel data base and predict the in-respository performance of spent fuel is being prepared; documentation was completed describing the test capsule hardware, specimens, safety analysis, and as-built assembly for the Climax-Materials Interaction Test; gamma dose calculations for 30 spent fuel canister filler materials were completed and documented; in support of the stabilizer material screening effort, all preliminary interaction, cost, and availability evaluations were completed; detailed planning for temperature limit, fill process, and prebreach disposal condition compatibility testing was completed.

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

    SciTech Connect

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

    1994-10-01

    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.

  10. Spent fuel dissolution rates as a function of burnup and water chemistry

    SciTech Connect

    Gray, W.J.

    1998-06-01

    To help provide a source term for performance-assessment calculations, dissolution studies on light-water-reactor (LWR) spent fuel have been conducted over the past few years at Pacific Northwest National Laboratory in support of the Yucca Mountain Site Characterization Project. This report describes that work for fiscal years 1996 through mid-1998 and includes summaries of some results from previous years for completeness. The following conclusions were based on the results of various flowthrough dissolution rate tests and on tests designed to measure the inventories of {sup 129}I located within the fuel/cladding gap region of different spent fuels: (1) Spent fuels with burnups in the range 30 to 50 MWd/kgM all dissolved at about the same rate over the conditions tested. To help determine whether the lack of burnup dependence extends to higher and lower values, tests are in progress or planned for spent fuels with burnups of 13 and {approximately} 65 MWd/kgM. (2) Oxidation of spent fuel up to the U{sub 4}O{sub 9+x} stage does not have a large effect on intrinsic dissolution rates. However, this degree of oxidation could increase the dissolution rates of relatively intact fuel by opening the grain boundaries, thereby increasing the effective surface area that is available for contact by water. From a disposal viewpoint, this is a potentially more important consideration than the effect on intrinsic rates. (3) The gap inventories of {sup 129}I were found to be smaller than the fission gas release (FGR) for the same fuel rod with the exception of the rod with the highest FGR. Several additional fuels would have to be tested to determine whether a generalized relationship exists between FGR and {sup 129}I gap inventory for US LWR fuels.

  11. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    SciTech Connect

    DeHart, M.D.

    1996-05-01

    Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

  12. Assessment of spent-fuel waste-form/stabilizer alternatives for geologic disposal

    SciTech Connect

    Einziger, R.E.; Himes, D.A.

    1982-06-01

    The Office of Nuclear Waste Isolation (ONWI) is studying the possibility of burying canisterized unreprocessed spent fuel in a deep geologic repository. One aspect of this study is an assessment of the possible spent fuel waste forms. The fuel performance portion of the Waste Form Assessment was to evaluate five candidate spent fuel waste forms for postemplacement performance with emphasis on their ability to retard the release of radionuclides to the repository geology. Spent fuel waste forms under general consideration were: (1) unaltered fuel assembly; (2) fuel assembly with end fittings removed to shorten the length; (3 rods vented to remove gases and resealed; (4) disassembled fuel bundles to close-pack the rods; and (5) rods chopped and fragments immobilized in a matrix material. Thirteen spent fuel waste forms, classified by generic stabilizer type, were analyzed for relative in-repository performance based on: (1) waste form/stabilizer support against lithostatic pressure; (2) long-term stability for radionuclide retention; (3) minimization of cladding degradation; (4) prevention of canister/repository breach due to pressurization; (5) stabilizer heat transfer; (6) the stabilizer as an independent barrier to radionuclide migration; and (7) prevention of criticality. The waste form candidates were ranked as follows: (1) the best waste form/stabilizer combination is the intact assembly, with or without end bells, vented (and resealed) or unvented, with a solid stabilizer; (2) a suitable alternative is the combination of bundled close-packed rods with a solid stabilizer around the outside of the bundle to resist lithostatic pressure; and (3) the other possible waste forms are of lower ranking with the worst waste form/stabilizer combination being the intact assembly with a gas stabilizer or the chopped fuel.

  13. Recent advances in hardware and software are to improve spent fuel measurements

    SciTech Connect

    Staples, P.; Beddingfield, D. H.; Lestone, J. P.; Pelowitz, D. G.; Bytchkov, M.; Starovich, Z.; Harizanov, I.; Luna-Vellejo, J.; Lavender, C.

    2001-01-01

    Vast quantities of spent fuel are available for safeguard measurements, primarily in Commonwealth of Independent States (CIS) of the former Soviet Union. This spent fuel, much of which consists of long-cooling-time material, is going to become less unique in the world safeguards arena as reprocessing projects or permanent repositories continue to be delayed or postponed. The long cooling time of many of the spent fuel assemblies being prepared for intermediate term storage in the CIS countries promotes the possibility of increased accuracy in spent fuel assays. This improvement is made possible through the process of decay of the Curium isotopes and of fission products. An important point to consider for the future that could advance safeguards measurements for reverification and inspection would be to determine what safeguards requirements should be imposed upon this 'new' class of spent fuel, Improvements in measurement capability will obviously affect the safeguards requirements. What most significantly enables this progress in spent fuel measurements is the improvement in computer processing power and software enhancements leading to user-friendly Graphical User Interfaces (GUT's). The software used for these projects significantly reduces the IAEA inspector's time expenditure for both learning and operating computer and data acquisition systems, At the same time, by standardizing the spent fuel measurements, it is possible to increase reproducibility and reliability of the measurement data. Hardware systems will be described which take advantage of the increased computer control available to enable more complex measurement scenarios. A specific example of this is the active regulation of a spent fuel neutron coincident counter's {sup 3}He tubes high voltage, and subsequent scaling of measurement results to maintain a calibration for direct assay of the plutonium content of Fast Breeder Reactor spent fuel. The plutonium content has been successfully determined

  14. Spent fuel storage and management in the United Kingdom

    SciTech Connect

    Sills, R.J.

    1989-04-01

    During the past 33 years, fuel of various types have been stored, transported and reprocessed in the United Kingdom. This paper provides an overview of those programs starting from the Magnox stations, through the AGR program and the move to LWR fuel. Throughout this time BNFL has provided services for fuel storage, reprocessing, transportation and the enrichment and fabrication of new fuel. The development of new plants and processes to handle the changing fuel types and the associated waste management schemes will be addressed. A description of future plans for fuel storage and reprocessing is included.

  15. Chemical Forms and Distribution of Platinum Group Metals and Technetium During Spent Fuel Reprocessing

    SciTech Connect

    Pokhitonov, Y.

    2007-07-01

    Amongst the fission products present in spent nuclear fuel of Nuclear Power Plants there are considerable quantities of platinum group metals (PGMs): ruthenium, rhodium and palladium. At the same time there are considerable amounts of technetium in the spent fuel, the problem of its removal at radiochemical plants being in operation encountering serious difficulties. Increased interest in this radionuclides is due not only to its rather large yield, but to higher mobility in the environment as well. However, the peculiarities of technetium chemistry in nitric acid solutions create certain problems when trying to separate it as a single product in the course of NPP's spent fuel reprocessing. The object of this work was to conduct a comprehensive analysis of platinum group metals and technetium behavior at various stages of spent fuel reprocessing and to seek the decisions which could make it possible to separate its as a single product. The paper will report data on platinum metals (PGM) and technetium distribution in spent fuel reprocessing products. The description of various techniques for palladium recovery from differing in composition radioactive solutions arising from reprocessing is given. (authors)

  16. Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

    NASA Astrophysics Data System (ADS)

    Wiss, Thierry; Hiernaut, Jean-Pol; Roudil, Danièle; Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J. M.; Matzke, Hans-Joachim; Weber, William J.

    2014-08-01

    Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted.

  17. Surrogate/spent fuel sabotage : aerosol ratio test program and Phase 2 test results.

    SciTech Connect

    Borek, Theodore Thaddeus III; Thompson, N. Slater; Sorenson, Ken Bryce; Hibbs, R.S.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno; Young, F. I.; Koch, Wolfgang; Brochard, Didier; Pretzsch, Gunter Guido; Lange, Florentin

    2004-05-01

    A multinational test program is in progress to quantify the aerosol particulates produced when a high energy density device, HEDD, impacts surrogate material and actual spent fuel test rodlets. This program provides needed 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 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 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 crucial for predicting radiological impacts. This document includes a thorough description of the test program, including the current, detailed test plan, concept and design, plus a description of all test components, and requirements for future components and related nuclear facility needs. It also serves as a program status report as of the end of FY 2003. All available test results, observations, and analyses - primarily for surrogate material Phase 2 tests using cerium oxide sintered ceramic pellets are included. 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 Nuclear Regulatory Commission.

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

    SciTech Connect

    Not Available

    1994-04-01

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

    SciTech Connect

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

    1990-04-01

    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.

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

    SciTech Connect

    Patrick, W.C.

    1986-03-30

    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. Evaluation of measured LWR spent fuel composition data for use in code validation

    SciTech Connect

    Hermann, O.W.; DeHart, M.D.; Murphy, B.D.

    1998-02-01

    Burnup credit (BUC) is a concept applied in the criticality safety analysis of spent nuclear fuel in which credit or partial credit is taken for the reduced reactivity worth of the fuel due to both fissile depletion and the buildup of actinides and fission products that act as net neutron absorbers. Typically, a two-step process is applied in BUC analysis: first, depletion calculations are performed to estimate the isotopic content of spent fuel based on its burnup history; second, three-dimensional (3-D) criticality calculations are performed based on specific spent fuel packaging configurations. In seeking licensing approval of any BUC approach (e.g., disposal, transportation, or storage) both of these two computational procedures must be validated. This report was prepared in support of the validation process for depletion methods applied in the analysis of spent fuel from commercial light-water-reactor (LWR) designs. Such validation requires the comparison of computed isotopic compositions with those measured via radiochemical assay to assess the ability of a computer code to predict the contents of spent fuel samples. The purpose of this report is to address the availability and appropriateness of measured data for use in the validation of isotopic depletion methods. Although validation efforts to date at ORNL have been based on calculations using the SAS2H depletion sequence of the SCALE code system, this report has been prepared as an overview of potential sources of validation data independent of the code system used. However, data that are identified as in use in this report refer to earlier validation work performed using SAS2H in support of BUC. This report is the result of a study of available assay data, using the experience gained in spent fuel isotopic validation and with a consideration of the validation issues described earlier. This report recommends the suitability of each set of data for validation work similar in scope to the earlier work.

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

    SciTech Connect

    Chen, J.; Wen, M.

    2013-07-01

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

  3. Partial Defect Verification of the Pressurized Water Reactor Spent Fuel Assemblies

    SciTech Connect

    Ham, Y S; Sitaraman, S

    2010-02-05

    The International Atomic Energy Agency (IAEA) has the responsibility to carry out independent inspections of all nuclear material and facilities subject to safeguards agreements in order to verify compliance with non-proliferation commitments. New technologies have been continuously explored by the IAEA and Member States to improve the verification measures to account for declared inventory of nuclear material and detect clandestine diversion and production of nuclear materials. Even with these efforts, a technical safeguards challenge has remained for decades for the case of developing a method in identifying possible diversion of nuclear fuel pins from the Light Water Reactor (LWR) spent fuel assemblies. We had embarked on this challenging task and successfully developed a novel methodology in detecting partial removal of fuel from pressurized water reactor spent fuel assemblies. The methodology uses multiple tiny neutron and gamma detectors in the form of a cluster and a high precision driving system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly without any movement of the fuel. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. The combined information of gamma and neutron signature is used to produce base signatures and they are principally dependent on the geometry of the detector locations, and exhibit little sensitivity to initial enrichment, burn-up or cooling time. A small variation in the fuel bundle such as a few missing pins changes the shape of the signature to enable detection. This resulted in a breakthrough method which can be used to detect pin diversion without relying on the nuclear power plant operator's declared operation data. Presented are the results of various Monte Carlo simulation studies and experiments from actual commercial PWR spent fuel assemblies.

  4. SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

    SciTech Connect

    BSC

    2004-12-01

    Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier

  5. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    SciTech Connect

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-02-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses.

  6. Development of a techno-economic model to optimization DOE spent nuclear fuel disposition

    SciTech Connect

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

    1997-11-01

    The purpose of the National Spent Nuclear Fuel (NSNF) Program conducted by Lockheed Martin Idaho Technology Co. (LMITCO) at the Idaho National Engineering and Environmental Laboratory (INEEL) is to evaluate what to do with the spent nuclear fuel (SNF) in the Department of Energy (DOE) complex. Final disposition of the SNF may require that the fuel be treated to minimize material concerns. The treatments may range from electrometallurgical treatment and chemical dissolution to engineering controls. Treatment options and treatment locations will depend on the fuel type and the current locations of the fuel. One of the first steps associated with selecting one or more sites for treating the SNF in the DOE complex is to determine the cost of each option. An economic analysis will assist in determining which fuel treatment alternative attains the optimum disposition of SNF at the lowest possible cost to the government and the public. For this study, a set of questions was developed for the electrometallurgical treatment process for fuels at several locations. The set of questions addresses all issues associated with the design, construction, and operation of a production facility. A matrix table was developed to determine questions applicable to various fuel treatment options. A work breakdown structure (WBS) was developed to identify a treatment process and costs from initial design to shipment of treatment products to final disposition. Costs will be applied to determine the life-cycle cost of each option. This technique can also be applied to other treatment techniques for treating spent nuclear fuel.

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

    SciTech Connect

    Parks, P.B.

    1993-10-01

    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.

  8. Refinements to temperature calculations of spent fuel assemblies when in a stagnant gas environment

    SciTech Connect

    Rhodes, C.A.; Haire, M.J.

    1984-01-01

    Undesirably high temperatures are possible in irradiated fuel assemblies because of the radioactive decay of fission products formed while in the reactor. The COXPRO computer code has been used for some time to calculate temperatures in spent fuel when the fuel is suspended in a stagnant gas environment. This code assumed radiation to be the only mode of heat dissipation within the fuel pin bundle. Refinements have been made to include conduction as well as radiation heat transfer within this code. Comparison of calculated and measured temperatures in four separate and independent tests indicate that maximum fuel assembly temperatures can be predicted to within about 6%. 2 references, 5 figures.

  9. Electrodialysis with bipolar membrane for regeneration of a spent activated carbon.

    PubMed

    Drouiche, N; Grib, H; Abdi, N; Lounici, H; Pauss, A; Mameri, N

    2009-10-15

    The main purpose of the present work was to develop a treatment method to regenerate granular adsorbent beds saturated with H(2)S by utilizing three electrodialysis compartments equipped with a cation or an anion exchange membrane or a bipolar membrane. Three electrodialysis compartments were utilized under various experimental parameters to determine the optimum conditions for the recovery of column particles saturated by H(2)S. The desulphurization operation is achieved with the extent of extraction close to 90% and an electric current density of about 30%. Use of the bipolar membrane makes it possible to regenerate the saturated adsorbent granules without adding chemical products. Since the only reagent was electricity, the projected economics are very attractive. PMID:19473766

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

    SciTech Connect

    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.; Brucher, W (Gesellschaft fur anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2006-10-01

    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

  11. Application of spent fuel characterization and leaching studies for validating alteration models

    SciTech Connect

    Quinones, Javier; Iglesias, Eduardo; Rodriguez, Nieves; Cobos Sabate, Joaquin; Martinez-Esparza, Aurora

    2007-07-01

    From the Spanish point of view, one of the key issues related to the HLW performance assessment is knowing and predicting, or modelling, the behaviour of spent fuel under geological repository conditions. Taking into account this objective, several experiments have been performed in order to split and determine the influence of different variables on the final stability of the spent fuel matrix in the geological repository. This paper presents some of the leaching results obtained with spent fuel and chemical analogues (UO{sub 2}, alpha doped-UO{sub 2}, SIMFUEL,) their application to extrapolate the corrosion behaviour for a long period of time and compare with corresponding data obtained using models. This procedure allows pointing out some of the uncertainties whose minimization is necessary to improve the models useful for performance assessment studies. (authors)

  12. Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

    SciTech Connect

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Svard, Staffan Jacobsson; Jansson, Peter; Swinhoe, Martyn T.; Tobin, Stephen J.

    2015-03-04

    Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.

  13. Historical overview of domestic spent nuclear fuel shipments in the United States

    SciTech Connect

    Pope, R.B.; Wankerl, M.W. ); Hamberger, C.R.; Schmid, S.P. )

    1992-01-01

    The information in this paper summarizes historical data on spent nuclear fuel shipments in the United States (US) from the period from 1964 to 1991. Information on shipments has been developed to establish a basis for developing a transportation system in the US for initiating shipments of spent nuclear fuel beginning in 1998. The paper shows that approximately 2700 power reactor spent nuclear fuel rail and truck casks have been shipped within the US during the past 28 years. In total, approximately 2000 metric tonnes of uranium (MTU) have been shipped to date, which compares with projected shipping rates of from 3000 to greater than 6000 MM per year when the US Civilian Radioactive Waste Management System is in full operation.

  14. Fission product partitioning in aerosol release from simulated spent nuclear fuel

    NASA Astrophysics Data System (ADS)

    Di Lemma, F. G.; Colle, J. Y.; Rasmussen, G.; Konings, R. J. M.

    2015-10-01

    Aerosols created by the vaporization of simulated spent nuclear fuel (simfuel) were produced by laser heating techniques and characterised by a wide range of post-analyses. In particular attention has been focused on determining the fission product behaviour in the aerosols, in order to improve the evaluation of the source term and consequently the risk associated with release from spent fuel sabotage or accidents. Different simulated spent fuels were tested with burn-up up to 8 at. %. The results from the aerosol characterisation were compared with studies of the vaporization process by Knudsen Effusion Mass Spectrometry and thermochemical equilibrium calculations. These studies permit an understanding of the aerosol gaseous precursors and the gaseous reactions taking place during the aerosol formation process.

  15. Comparative economics for DUCRETE spent fuel storage cask handling, transportation, and capital requirements

    SciTech Connect

    Powell, F.P.

    1995-04-01

    This report summarizes economic differences between a DUCRETE spent nuclear fuel storage cask and a conventional concrete storage cask in the areas of handling, transportation, and capital requirements. The DUCRETE cask is under evaluation as a new technology that could substantially reduce the overall costs of spent fuel and depleted U disposal. DUCRETE incorporates depleted U in a Portland cement mixture and functions as the cask`s primary radiation barrier. The cask system design includes insertion of the US DOE Multi-Purpose Canister inside the DUCRETE cask. The economic comparison is from the time a cask is loaded in a spent fuel pool until it is placed in the repository and includes the utility and overall US system perspectives.

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

    SciTech Connect

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

    2009-07-01

    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.

  17. An evaluation of near-field host rock temperatures for a spent fuel repository

    SciTech Connect

    Altenhofen, M.K.; Lowery, P.S.

    1988-11-01

    A repository heat transfer analysis has been performed by the Pacific Northwest Laboratory (PNL) for the US Department of Energy's Performance Assessment Scientific Support Program. The objective of this study was to evaluate the near-field thermal environmental conditions for a spent fuel repository system. A spent fuel logistics analysis was performed using a waste management system simulation model, WASTES-II, to evaluate the thermal characteristics of spent fuel received at the repository. A repository-scale thermal analysis was performed using a finite difference heat transfer code, TEMPEST, to evaluate the near-field host rock temperature. The calculated temporal and spatial distributions of near-field host rock temperatures provide input to the repository source term model in evaluations of engineered barrier system performance. 9 refs., 10 figs., 2 tabs.

  18. Assessment of nitrogen as an atmosphere for dry storage of spent LWR fuel

    SciTech Connect

    Gilbert, E.R.; Knox, C.A.; White, G.D.

    1985-09-01

    Interim dry storage of spent light-water reactor (LWR) fuel is being developed as a licensed technology in the United States. Because it is anticipated that license agreements will specify dry storage atmospheres, the behavior of spent LWR fuel in a nitrogen atmosphere during dry storage was investigated. In particular, the thermodynamics of reaction of nitrogen compounds (expected to form in the cover gas during dry storage) and residual impurities (such as moisture and oxygen) with Zircaloy cladding and with spent fuel at sites of cladding breaches were examined. The kinetics of reaction were not considered it was assumed that the 20 to 40 years of interim dry storage would be sufficient for reactions to proceed to completion. The primary thermodynamics reactants were found to be NO/sub 2/, N/sub 2/O, H/sub 2/O/sub 2/, and O/sub 2/. The evaluation revealed that the limited inventories of these reactants produced by the source terms in hermetically sealed dry storage systems would be too low to cause significant spent fuel degradation. Furthermore, the oxidation of spent fuel to degrading O/U ratios is unlikely because the oxidation potential in moist nitrogen limits O/U ratios to values less than UO/sub 2.006/ (the equilibrium stoichiometric form in equilibrium with moist nitrogen). Tests were performed with bare spent UO/sub 2/ fuel and nonirradiated UO/sub 2/ pellets (with no Zircaloy cladding) in a nitrogen atmosphere containing moisture concentrations greater than encountered under dry storage conditions. These tests were performed for at least 1100 h at temperatures as high as 380/sup 0/C, where oxidation reactions proceed in a matter of minutes. No visible degradation was detected, and weight changes were negligible.

  19. Chemical Speciation of Neptunium in Spent Fuel. 1st Progress Report

    SciTech Connect

    Czerwinski, Ken; Sherman, Christi; Reed, Don

    2000-03-02

    This project will examine the chemical speciation of neptunium in spent nuclear fuel. The R&D fields covered by the project include waste host materials and actinide chemistry. Examination of neptunium is chosen since it was identified as a radionuclide of concern by the NERI workshop. Additionally, information on the chemical form of neptunium in spent fuel is lacking. The identification of the neptunium species in spent fuel would allow a greater scientific based understanding of its long-term fate and behavior in waste forms. Research to establish the application and development of X-ray synchrotrons radiation (XSR) techniques to determine the structure of aqueous, adsorbed, and solid actinide species of importance to nuclear considerations is being conducted at Argonne. These studies extend current efforts within the Chemical Technology Division at Argonne National Laboratory to investigate actinide speciation with more conventional spectroscopic and solids characterization (e.g. SEM, TEM, and XRD) methods. Our project will utilize all these techniques for determining neptunium speciation in spent fuel. We intend to determine the chemical species and oxidation state of neptunium in spent fuel and alteration phases. Different types of spent fuel will be examined. Once characterized, the chemical behavior of the identified neptunium species will be evaluated if it is not present in the literature. Special attention will be given to the behavior of the neptunium species under typical repository near-field conditions (elevated temperature, high pH, varying Eh). This will permit a timely inclusion of project results into near-field geochemical models. Additionally, project results and methodologies have applications to neptunium in the environment, or treatment of neptunium containing waste.

  20. Determining Spent Nuclear Fuel's Plutonium Content, Initial Enrichment, Burnup, and Cooling Time

    SciTech Connect

    Cheatham, Jesse R; Francis, Matthew W

    2011-01-01

    The Next Generation of Safeguards Initiative is examining nondestructive assay techniques to determine the total plutonium content in spent nuclear fuel. The goal of this research was to develop new techniques that can independently verify the plutonium content in a spent fuel assembly without relying on an operator's declarations. Fundamentally this analysis sought to answer the following questions: (1) do spent fuel assemblies contain unique, identifiable isotopic characteristics as a function of their burnup, cooling time, and initial enrichment; (2) how much variation can be seen in spent fuel isotopics from similar and dissimilar reactor power operations; and (3) what isotopes (if any) could be used to determine burnup, cooling time, and initial enrichment? To answer these questions, 96,000 ORIGEN cases were run that simulated typical two-cycle operations with burnups ranging from 21,900 to 72,000 MWd/MTU, cooling times from 5 to 25 years, and initial enrichments between 3.5 and 5.0 weight percent. A relative error coefficient was determined to show how numerically close a reference solution has to be to another solution for the two results to be indistinguishable. By looking at the indistinguishable solutions, it can be shown how a precise measurement of spent fuel isotopics can be inconclusive when used in the absence of an operator's declarations. Using this Method of Indistinguishable Solutions (MIS), we evaluated a prominent method of nondestructive analysis - gamma spectroscopy. From this analysis, a new approach is proposed that demonstrates great independent forensic examination potential for spent nuclear fuel by examining both the neutron emissions of Cm-244 and the gamma emissions of Cs-134 and Eu-154.