WorldWideScience

Sample records for pwr spent nuclear

  1. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  2. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  3. Criticality safety and sensitivity analyses of PWR spent nuclear fuel repository facilities

    NARCIS (Netherlands)

    Maucec, M; Glumac, B

    2005-01-01

    Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based stor

  4. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  5. Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1997-04-01

    A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package

  6. Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    1998-09-01

    The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the

  7. Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

    1999-02-17

    Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

  8. Experiment on the improvement of sinterability for dry recycling nuclear fuel pellets by using simulated spent PWR fuel of high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Park, G. I.; Lee, Jae W.; Cho, K. H.; Lee, D. Y.; Lee, Y. S.; Lee, J. W.; Yang, M. S.; Shin, W. C

    2004-09-01

    To study the fabrication characteristics of dry recycling nuclear fuel using spent PWR fuel with high burnup of 60,000 MWd/tU, the fission products of spent PWR fuel was analyzed by ORIGEN-2 code. Simulated spent PWR fuel pellets were fabricated by using UO{sub 2} powder added by the simulated fission products. The simulated dry-recycling-fuel pellets were fabricated by dry recycling fuel fabrication flow including 3 cycle treated OREOX(Oxidation and REduction of OXide fuel) process. A small amount of dopant such as TiO{sub 2}, Nb{sub 2}O{sub 5}, Li{sub 2}O are added to increase sinterability of the OREOX treated powder. As the results of experiments, the densities of sintered pellets without dopant ranged from 10.04 to 10.34 g/cm{sup 3}(94.3 to 97.1% of T.D.), the grain size of the pellets ranged from 3 to 4 {mu}m. The sintered density of the pellets with TiO{sub 2} or Nb{sub 2}O{sub 5} ranged from 10.46 to 10.32 g/cm{sup 3}(98.2 to 96.9 % of T.D.) The grain size of the pellets with TiO{sub 2}, Nb{sub 2}O{sub 5} or Li{sub 2}O ranged from 7.3 to 12.2 {mu}m.

  9. PWR-UO{sub 2} nuclear fuel criticality study: control rod effects on infinite neutron multiplication factor and spent fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Sousa, R.V.; Pereira, C., E-mail: claubia@nuclear.ufmg.br; Silva, C.A.M.; Costa, A.L.; Veloso, M.A.F.; Oliveira, A.H. de

    2013-10-15

    Highlights: • A three-dimensional model of a PWR fuel were simulated. • Results using TRITON/T6-DEPL module in SCALE 6.0 and two libraries (238 and 44 groups) were compared. • Variations in the infinite neutron multiplication factor and the nuclides concentrations, both under control rod insertion effects were analysed. • Results show very good agreement with those published by OECD. -- Abstract: Deterministic and stochastic nuclear codes are software packages used to perform reactor physics calculations, especially in PWRs, the most common type of nuclear reactor currently in operation. The NEA Expert Group on Burn-up Credit Criticality Safety has published a Benchmark with results obtained from simulations of PWR-UO{sub 2} nuclear fuel. The same simulations were performed at DEN/UFMG with SCALE 6.0, a modular nuclear system code developed by Oak Ridge National Laboratory using two different neutron energy libraries (238 and 44 groups). The results obtained using a three-dimensional model with the T6-DEPL sequence of the TRITON module in SCALE 6.0 for spent fuel inventory and infinite neutron multiplication factor calculations show very good agreement with those published by the OECD. The main goal of this work is to validate the methodology at DEN/UFMG for future use in simulations related to Angra I, II and III Nuclear Power Plants.

  10. PWR and BWR spent fuel assembly gamma spectra measurements

    Science.gov (United States)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    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) detect 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. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  11. Calculation of the radionuclides in PWR spent fuel samples for SFR experiment planning.

    Energy Technology Data Exchange (ETDEWEB)

    Naegeli, Robert Earl

    2004-06-01

    This report documents the calculation of radionuclide content in the pressurized water reactor (PWR) spent fuel samples planned for use in the Spent Fuel Ratio (SPR) Experiments at Sandia National Laboratories, Albuquerque, New Mexico (SNL) to aid in experiment planning. The calculation methods using the ORIGEN2 and ORIGEN-ARP computer codes and the input modeling of the planned PWR spent fuel from the H. B. Robinson and the Surry nuclear power plants are discussed. The safety hazards for the calculated nuclide inventories in the spent fuel samples are characterized by the potential airborne dose and by the portion of the nuclear facility hazard category 2 and 3 thresholds that the experiment samples would present. In addition, the gamma ray photon energy source for the nuclide inventories is tabulated to facilitate subsequent calculation of the direct and shielded dose rates expected from the samples. The relative hazards of the high burnup 72 gigawatt-day per metric ton of uranium (GWd/MTU) spent fuel from H. B. Robinson and the medium burnup 36 GWd/MTU spent fuel from Surry are compared against a parametric calculation of various fuel burnups to assess the potential for higher hazard PWR fuel samples.

  12. Disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    This report addresses the topic of the mined geologic disposal of spent nuclear fuel from Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Although some fuel processing options are identified, most of the information in this report relates to the isolation of spent fuel in the form it is removed from the reactor. The characteristics of the waste management system and research which relate to spent fuel isolation are discussed. The differences between spent fuel and processed HLW which impact the waste isolation system are defined and evaluated for the nature and extent of that impact. What is known and what needs to be determined about spent fuel as a waste form to design a viable waste isolation system is presented. Other waste forms and programs such as geologic exploration, site characterization and licensing which are generic to all waste forms are also discussed. R and D is being carried out to establish the technical information to develop the methods used for disposal of spent fuel. All evidence to date indicates that there is no reason, based on safety considerations, that spent fuel should not be disposed of as a waste.

  13. Study of a brazilian cask and its installation for PWR spent nuclear fuel dry storage; Estudo de um casco nacional e sua instalacao para armazenagem seca de combustivel nuclear queimado gerado em reatores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, Luiz Sergio

    2009-07-01

    Spent nuclear fuel (SNF) is removed from the nuclear reactor after the depletion on efficiency in generating energy. After the withdrawal from the reactor core, the SNF is temporarily stored in pools at the same site of the reactor. At this time, the generated heat and the short and medium lived radioactive elements decay to levels that allow removing SNF from the pool and sending it to temporary dry storage. In that phase, the fuel needs to be safely and efficiently stored, and then, it can be retrieved in a future, or can be disposed as radioactive waste. The amount of spent fuel increases annually and, in the next years, will still increase more, because of the construction of new nuclear plants. Today, the number of new facilities back up to levels of the 1970's, since it is greater than the amount of decommissioning in old installations. As no final decision on the back-end of the nuclear fuel cycle is foreseen in the near future in Brazil, either to recover the SNF or to consider it as radioactive waste, this material has to be isolated in some type of storage model existing around the world. In the present study it is shown that dry SNF storage is the best option. A national cask model for SNF as well these casks storage installation are proposed. It is a multidisciplinary study in which the engineering conceptual task was developed and may be applied to national SNF removed from the Brazilian power reactors, to be safely stored for a long time until the Brazilian authorities will decide about the site for final disposal. (author)

  14. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Science.gov (United States)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  15. Spent fuel data base: commercial light water reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hauf, M.J.; Kniazewycz, B.G.

    1979-12-01

    As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel.

  16. The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

    Science.gov (United States)

    Park, S D; Kim, J S; Han, S H; Ha, Y K; Song, K S; Jee, K Y

    2009-09-01

    In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

  17. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  18. An extension of the validation of SCALE (SAS2H) isotopic predictions for PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Hermann, O.W.

    1996-09-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms. Unlike fresh fuel assumptions typically used in criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict spent fuel composition; these isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the depletion codes and data, experiment is compared with predictions; such comparisons have been done in earlier ORNL work. This report describes additional independent measurements and corresponding calculations as a supplement. The current work includes measured isotopic data from 19 spent fuel samples from the Italian Trino Vercelles PWR and the US Turkey Point-3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results on combination of measured-to-calculated ratios are presented. The results described herein represent an extension to a new reactor design and spent fuel samples with enrichment as high as 3.9 wt% {sup 235}U. Consistency with the earlier work for each of two different cross-section libraries suggests that the estimated biases for each of the isotopes in the earlier work are reasonably good estimates.

  19. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2014-10-15

    Highlights: • A new design for the PWR assemblies for direct use of spent fuel was proposed. • The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors. • The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. • MCNPX is used for the calculations that showed that the burnup can be increased by about 25%. • Acceptable linear heat generation rate in hot rods and improved Pu proliferation resistance. - Abstract: In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO{sub 2} fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

  20. MELCOR Modeling of Air-Cooled PWR Spent Fuel Assemblies in Water empty Fuel Pools

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L. E.; Lopez, C.

    2013-07-01

    The OECD Spent Fuel Project (SFP) investigated fuel degradation in case of a complete Loss-Of- Coolant-Accident in a PWR spent fuel pool. Analyses of the SFP PWR ignition tests have been conducted with the 1.86.YT.3084.SFP MELCOR version developed by SNL. The main emphasis has been placed on assessing the MELCOR predictive capability to get reasonable estimates of time-to-ignition and fire front propagation under two configurations: hot neighbor (i.e., adiabatic scenario) and cold neighbor (i.e., heat transfer to adjacent fuel assemblies). A detailed description of hypotheses and approximations adopted in the MELCOR model are provided in the paper. MELCOR results accuracy was notably different between both scenarios. The reasons are highlighted in the paper and based on the results understanding a set of remarks concerning scenarios modeling is given.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cardoni, Jeffrey

    2010-11-01

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

  2. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  3. An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions for PWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Isotopic characterization of spent fuel via depletion and decay calculations is necessary for determination of source terms for subsequent system analyses involving heat transfer, radiation shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in subsequent criticality calculations to assess the reduced worth of spent fuel. To validate the codes and data used in depletion approaches, experimental measurements are compared with numerical predictions for relevant spent fuel samples. Such comparisons have been performed in earlier work at the Oak Ridge National Laboratory (ORNL). This report describes additional independent measurements and corresponding calculations, which supplement the results of the earlier work. The current work includes measured isotopic data from 19 spent fuel samples obtained from the Italian Trino Vercelles pressurized-water reactor (PWR) and the U.S. Turkey Point Unit 3 PWR. In addition, an approach to determine biases and uncertainties between calculated and measured isotopic concentrations is discussed, together with a method to statistically combine these terms to obtain a conservative estimate of spent fuel isotopic concentrations. Results are presented based on the combination of measured-to-calculated ratios for earlier work and the current analyses. The results described herein represent an extension to a new reactor design not included in the earlier work, and spent fuel samples with enrichment as high as 3.9 wt % {sup 235}U. Results for the current work are found to be, for the most part, consistent with the findings of the earlier work. This consistency was observed for results obtained from each of two different cross-section libraries and suggests that the estimated biases determined for

  4. Application of SCALE4.4 system for burnup credit criticality analysis of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seung gy; Bae, Kang mok; Shin, YoungJoon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-07-01

    An investigation on the application of burnup credit for a PWR spent fuel storage pool has been carried out with the use of the SCALE 4.4 computer code system consisting of SAS2H and CSAS6 modules in association with 44-group SCALE cross-section library. Prior to the application of the computer code system, a series of bench markings have been performed in comparison with available data. A benchmarking of the SAS2h module has been done for experimental concentration data of 54 PWR spent fuel and then correction factors with a 95% probability at a 95% confidence level have been determined on the basis of the calculated and measured concentrations of 38 nuclides. After that, the bias which might have resulted from the use of the CSAS6 module has been calculated for 46 criticality experimental data of UO{sub 2} fuel and MOX fuel assemblies. The calculation bias with one-sided tolerance limit factor (2.086) corresponding to a 95% probability at a 95% confidence level has consequently been obtained to be 0.00834. Burnup credit criticality analysis has been done for the PWR spent fuel storage pool by means of the benchmarked or validated code system. It is revealed that the minimum burnup for safe storage is 7560 MWd/tU in 5 wt% enriched fuel if both actinides and fission products in spent fuel are taken into account. However, the minimum value required seems to be 9,565 MWd/tU in the same enriched fuel provided that only the actinides are taken into consideration. (author)

  5. DUPIC fuel fabrication using spent PWR fuels at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

    2000-12-01

    This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter III, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

  6. DUPIC fuel fabrication using spent PWR fuel at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Yang, Myung Seung; Ko, Won Il and others

    2001-09-01

    This document contains DUPIC fuel cycle R and D activities to be carried out for 5 years beyond the scope described in the report KAERI/AR-510/98, which was attached to Joint Determination for Post-Irradiation Examination of irradiated nuclear fuel, by MOST and US Embassy in Korea, signed on April 8, 1999. This document is purposely prepared as early as possible to have ample time to review that the over-all DUPIC activities are within the scope and contents in compliance to Article 8(C) of ROK-U.S. cooperation agreement, and also maintain the current normal DUPIC project without interruption. Manufacturing Program of DUPIC Fuel in DFDF and Post Irradiation Examination of DUPIC Fuel are described in Chapter I and Chapter II, respectively. In Chapter 3/4y, safeguarding procedures in DFDF and on-going R and D on DUPIC safeguards such as development of nuclear material accounting system and development of containment/surveillance system are described in details.

  7. Analysis of experimental measurements of PWR fresh and spent fuel assemblies using Self-Interrogation Neutron Resonance Densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov; Menlove, Howard O., E-mail: hmenlove@lanl.gov

    2015-05-01

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is a new NDA technique that was developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for LWR fuel assemblies. The SINRD detector consists of four fission chambers (FCs) wrapped with different absorber filters to isolate different parts of the neutron energy spectrum and one ion chamber (IC) to measure the gross gamma rate. As a result, two different techniques can be utilized using the same SINRD detector unit and hardware. These techniques are the Passive Neutron Multiplication Counter (PNMC) method and the SINRD method. The focus of the work described in this paper is the analysis of experimental measurements of fresh and spent PWR fuel assemblies that were performed at LANL and the Korea Atomic Energy Research Institute (KAERI), respectively, using the SINRD detector. The purpose of these experiments was to assess the following capabilities of the SINRD detector: 1) reproducibility of measurements to quantify systematic errors, 2) sensitivity to water gap between detector and fuel assembly, 3) sensitivity and penetrability to the removal of fuel rods from the assembly, and 4) use of PNMC/SINRD ratios to quantify neutron multiplication and/or fissile content. The results from these simulations and measurements provide valuable experimental data that directly supports safeguards research and development (R&D) efforts on the viability of passive neutron NDA techniques and detector designs for partial defect verification of spent fuel assemblies. - Highlights: • Experimental measurements of PWR fresh and spent FAs were performed with SINRD. • Good agreement of MCNPX and measured results confirmed accuracy of SINRD model. • For fresh fuel, SINRD and PNMC ratios were not sensitive to water gaps of ≤5-mm. • Practical use of SINRD would be in Fork detector to reduce systematic uncertainties.

  8. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  9. Assessment of PWR plutonium burners for nuclear energy centers

    Energy Technology Data Exchange (ETDEWEB)

    Frankel, A J; Shapiro, N L

    1976-06-01

    The purpose of the study was to explore the performance and safety characteristics of PWR plutonium burners, to identify modifications to current PWR designs to enhance plutonium utilization, to study the problems of deploying plutonium burners at Nuclear Energy Centers, and to assess current industrial capability of the design and licensing of such reactors. A plutonium burner is defined to be a reactor which utilizes plutonium as the sole fissile addition to the natural or depleted uranium which comprises the greater part of the fuel mass. The results of the study and the design analyses performed during the development of C-E's System 80 plant indicate that the use of suitably designed plutonium burners at Nuclear Energy Centers is technically feasible.

  10. Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.C.; Bang, K.S.; Seo, K.S.; Kim, H.D. [Korea Atomic Energy Research Inst., Daejeon (Korea); Choi, B.I.; Lee, H.Y.; Song, M.J. [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea)

    2004-07-01

    The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal and off-normal conditions. The environmental temperature is assumed to be 27 {open_square} under the normal condition. The off-normal condition has an environmental temperature of 40 {open_square}. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of ventilation system have been carried out for the determination of the optimum duct size and shape. The finite volume computational fluid dynamics code FLUENT was used for the thermal analysis. In the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal condition and off-normal conditions.

  11. Life management plants at nuclear power plants PWR; Planes de gestion de vida en centrales nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Esteban, G.

    2014-10-01

    Since in 2009 the CSN published the Safety Instruction IS-22 (1) which established the regulatory framework the Spanish nuclear power plants must meet in regard to Life Management, most of Spanish nuclear plants began a process of convergence of their Life Management Plants to practice 10 CFR 54 (2), which is the current standard of Spanish nuclear industry for Ageing Management, either during the design lifetime of the plant, as well as for Long-Term Operation. This article describe how Life Management Plans are being implemented in Spanish PWR NPP. (Author)

  12. Spent fuel dry storage technology development: thermal evaluation of isolated drywells containing spent fuel (1 kW PWR spent fuel assembly)

    Energy Technology Data Exchange (ETDEWEB)

    Unterzuber, R; Wright, J B

    1980-09-01

    A spent fuel Isolated Drywell Test was conducted at the Engine-Maintenance, Assembly and Disassembly (E-MAD) facility on the Nevada Test Site. Two PWR spent fuel assemblies having a decay heat level of approximately 1.1 kW were encapsulated inside the E-MAD Hot Bay and placed in instrumented near-surface drywell storage cells. Temperatures from the two isolated drywells and the adjacent soil have been recorded throughout the 19 month Isolated Drywell Test. Canister and drywell liner temperatures reached their peak values (254{sup 0}F and 203{sup 0}F, respectively) during August 1979. Thereafter, all temperatures decreased and showed a cycling pattern which responded to seasonal atmospheric temperature changes. A computer model was utilized to predict the thermal response of the drywell. Computer predictions of the drywell temperatures and the temperatures of the surrounding soil are presented and show good agreement with the test data.

  13. Spent Nuclear Fuel (SNF) Project Execution Plan

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Burn-up credit in criticality safety of PWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoud, Rowayda F., E-mail: Rowayda_mahmoud@yahoo.com [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Shaat, Mohamed K. [Nuclear Engineering, Reactors Department, Nuclear Research Center, Atomic Energy Authority (Egypt); Nagy, M.E.; Agamy, S.A. [Professor of Nuclear Engineering, Nuclear and Radiation Department, Alexandria University (Egypt); Abdelrahman, Adel A. [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority (Egypt)

    2014-12-15

    Highlights: • Designing spent fuel wet storage using WIMS-5D and MCNP-5 code. • Studying fresh and burned fuel with/out absorber like “B{sub 4}C and Ag–In–Cd” in racks. • Sub-criticality was confirmed for fresh and burned fuel under specific cases. • Studies for BU credit recommend increasing fuel burn-up to 60.0 GWD/MTU. • Those studies require new core structure materials, fuel composition and cladding. - Abstract: The criticality safety calculations were performed for a proposed design of a wet spent fuel storage pool. This pool will be used for the storage of spent fuel discharged from a typical pressurized water reactor (PWR). The mathematical model based on the international validated codes, WIMS-5 and MCNP-5 were used for calculating the effective multiplication factor, k{sub eff}, for the spent fuel stored in the pool. The data library for the multi-group neutron microscopic cross-sections was used for the cell calculations. The k{sub eff} was calculated for several changes in water density, water level, assembly pitch and burn-up with different initial fuel enrichment and new types and amounts of fixed absorbers. Also, k{sub eff} was calculated for the conservative fresh fuel case. The results of the calculations confirmed that the effective multiplication factor for the spent fuel storage is sub-critical for all normal and abnormal states. The future strategy for the burn-up credit recommends increasing the fuel burn-up to a value >60.0 GWD/MTU, which requires new fuel composition and new fuel cladding material with the assessment of the effects of negative reactivity build up.

  15. Criticality safety aspects of spent fuel arrays from emerging nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Nicolaou, G. [University of Thrace, Department of Electrical and Computer Engineering, Laboratory of Nuclear Technology, Kimmerria Campus, 67100 Xanthi (Greece)

    2010-07-01

    Emerging nuclear fuel cycles: fuels with Pu or minor actinides (MA) for their self-generated recycling or transmutation in PWR or FR {yields} reduction of radiotoxicity of HLW. The aim of work is to assess criticality (k{sub {infinity}}) of arrays of spent nuclear fuels from these emerging fuel cycles. Procedures: Calculations of - k{sub {infinity}}, using MCNP5 based on fresh and spent fuel compositions (infinite arrays), - spent fuel compositions using ORIGEN. Fuels considered: - commercial PWR-UO{sub 2} (R1) and -MOX (R2), [45 GWd/t] and fast reactor [100 GWd/t] (R3), - PWR self-generated Pu recycling (S1) and MA recycling (S2), FR self-generated MA recycling (S3), FR with 2% {sup 237}Np for transmutation purposes (T). Results: k{sub {infinity}} based on fresh and spent fuel compositions is shown. Fuels are clustered in two distinct families: - fast reactor fuels, - thermal reactor fuels; k{sub {infinity}} decreases when calculated on the basis of actinide and fission product inventory. In conclusions: - Emerging fuels considered resemble their corresponding commercial fuels; - k{sub {infinity}} decreases in all cases when calculated on the basis of spent fuel compositions (reactivity worth {approx}-20%{Delta}k/k), hence improving the effectiveness of packaging. (author)

  16. Design of the Ancillary Equipment for Spent Fuel Transport of PWR Nuclear Power Plant%压水堆核电站乏燃料运输辅助设备设计

    Institute of Scientific and Technical Information of China (English)

    瓮松峰

    2012-01-01

    In order to load and unload spent fuel assemblies safely in NPP, the ancillary equipment for spent fuel transport used to perform fluid transfer activities such as water feeding, air charging, and cask drying and heat rejection. This paper introduces the requirements, the scheme and the principle of the ancillary equipment, which adopts centralized-control and modular design, and have the performances such as higher work efficiency, higher safety, less radioactive substance accumulation and friendly man-machine conversation.%乏燃料运输辅助设备是用于对乏燃料运输容器进行充水排气、充气排水、充气风干的设备,并可对已装载乏燃料组件的容器进行充水冷却,实现乏燃料组件的安全装卸.本文介绍乏燃料运输辅助设备的功能要求、设计方案和工作原理.设备采用模块化设计和集中控制,具有工作效率高、结构简单、安全性高、不易积聚放射性物质和人-机交互友好的特点.

  17. Theoretical estimation of the impact velocity during the PWR spent drop in water condition

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Oh Joon; Park, Nam Gyu; Lee, Seong Ki; Kim, Jae Ik [KEPCO NF, Daejeon (Korea, Republic of)

    2016-06-15

    The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the 3×3 short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.

  18. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  19. Validation of the scale system for PWR spent fuel isotopic composition analyses

    Energy Technology Data Exchange (ETDEWEB)

    Hermann, O.W.; Bowman, S.M.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Laboratories, Las Vegas, NV (United States)

    1995-03-01

    The validity of the computation of pressurized-water-reactor (PWR) spent fuel isotopic composition by the SCALE system depletion analysis was assessed using data presented in the report. Radiochemical measurements and SCALE/SAS2H computations of depleted fuel isotopics were compared with 19 benchmark-problem samples from Calvert Cliffs Unit 1, H. B. Robinson Unit 2, and Obrigheim PWRs. Even though not exhaustive in scope, the validation included comparison of predicted and measured concentrations for 14 actinides and 37 fission and activation products. The basic method by which the SAS2H control module applies the neutron transport treatment and point-depletion methods of SCALE functional modules (XSDRNPM-S, NITAWL-II, BONAMI, and ORIGEN-S) is described in the report. Also, the reactor fuel design data, the operating histories, and the isotopic measurements for all cases are included in detail. The underlying radiochemical assays were conducted by the Materials Characterization. Center at Pacific Northwest Laboratory as part of the Approved Testing Material program and by four different laboratories in Europe on samples processed at the Karlsruhe Reprocessing Plant.

  20. PWR-FBR with closed fuel cycle for a sustainable nuclear energy supply in China

    Institute of Scientific and Technical Information of China (English)

    XU Mi

    2007-01-01

    From the thermal reactor to the fast reactor and then to the fusion reactor; this is the three-step strategy that has been decided for a sustainable nuclear energy supply in China. As the main thermal reactor type, the commercialized development phase of the pressurized water reactor (PWR) has been stepped up. The development of the fast reactor (FBR) is still in the early stage, marked by China experimental fast reactor (CEFR), which is currently under construction. According to the strategy study on the fast reactor development in China, its engineering development will be divided into three steps: the CEFR with a power of 65 MWt 20 Mwe; the China prototype fast reactor (CPFR) with a power of 1 500 MWt/600 Mwe; and the China demonstration fast reactor (CDFR) with a power of 2 500-3 750 MWt 1 000-1 500 Mwe. With regards to the fuel cycle, a 100 ta PWR spent fuel reprocessing pilot plant and a 500 kg/a MOX fabrication plant are under construction. A project involving the construction of an industrial reprocessing plant and an MOX fabrication plant are also under application phase.

  1. Parametric Analysis of PWR Spent Fuel Depletion Parameters for Long-Term-Disposal Criticality Safety

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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

  3. Spent Nuclear Fuel Transport Reliability Study

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  4. Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

    Directory of Open Access Journals (Sweden)

    Herrero J.J.

    2017-01-01

    Full Text Available In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

  5. Spent Nuclear Fuel Alternative Technology Decision Analysis

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Spent Nuclear Fuel Project Technical Databook

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Spent nuclear fuel rods encapsulated in copper

    Energy Technology Data Exchange (ETDEWEB)

    Hanes, H.D.

    1984-04-01

    Using hot isostatic pressing, spent nuclear fuel rods and other radioactive wastes can be encapsulated in solid copper. The copper capsule which is formed is free of pores and cracks, and is highly resistant to attack by reducing ground waters. Such capsules should contain radioactive materials safely for hundreds of thousands of years in underground storage.

  8. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Spent nuclear fuel project product specification

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. MACSTOR{trademark}: Dry spent fuel storage for the nuclear power industry

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F.E.; Pattantyus, P. [AECL Candu, Montreal, Quebec (Canada); Hanson, A.S. [Transnuclear, Inc., Hawthorne, NY (United States)

    1993-12-31

    Safe storage of spent fuel has long been an area of critical concern for the nuclear power industry. As fuel pools fill up and re-racking possibilities become exhausted, power plant operators will find that they must ship spent fuel assemblies off-site or develop new on-site storage options. Many utility companies are turning to dry storage for their spent fuel assemblies. The MACSTOR (Modular Air-cooled Canister STORage) concept was developed with this in mind. Derived from AECL`s successful vertical loading, concrete silo program for storing CANDU nuclear spent fuel, MACSTOR was developed for light water reactor spent fuel and was subjected to full scale thermal testing. The MACSTOR Module is a monolithic, shielded concrete vault structure than can accommodate up to 24 spent fuel canisters. Each canister holds 12 PWR or 32 PWR previously cooled spent fuel assemblies with burn-up rates as high as 45,000 MWD/MTU. The structure is passively cooled by natural convection through an array of inlet and outlet gratings and galleries serving a central plenum where the (vertically) stored canisters are located. The canisters are continuously monitored by means of a pressure monitoring system developed by TNI. The MACSTOR system includes the storage module(s), an overhead gantry system for cask handling, a transfer cask for moving fuel from wet to dry storage and a cask transporter. The canister and transfer cask designs are based on Transnuclear transport cask designs and proven hot cell transfer cask technology, adapted to requirements for on-site spent fuel storage. This Modular Air Cooled System has a number of inherent advantages: efficient use of construction materials and site space; cooling is virtually impossible to impede; has the ability to monitor fuel confinement boundary integrity during storage; the fuel canisters may be used for both storage and transport and canisters utilize a flanged, ASME-III closure system that allows for easy inspection.

  11. Nuclear spent fuel management scenarios. Status and assessment report

    Energy Technology Data Exchange (ETDEWEB)

    Dufek, J.; Arzhanov, V.; Gudowski, W. [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Nuclear and Reactor Physics

    2006-06-15

    The strategy for management of spent nuclear fuel from the Swedish nuclear power programme is interim storage for cooling and decay for about 30 years followed by direct disposal of the fuel in a geologic repository. In various contexts it is of interest to compare this strategy with other strategies that might be available in the future as a result of ongoing research and development. In particular partitioning and transmutation is one such strategy that is subject to considerable R and D-efforts within the European Union and in other countries with large nuclear programmes. To facilitate such comparisons for the Swedish situation, with a planned phase out of the nuclear power programme, SKB has asked the team at Royal Inst. of Technology to describe and explore some scenarios that might be applied to the Swedish programme. The results of this study are presented in this report. The following scenarios were studied by the help of a specially developed computer programme: Phase out by 2025 with direct disposal. Burning plutonium and minor actinides as MOX in BWR. Burning plutonium and minor actinides as MOX in PWR. Burning plutonium and minor actinides in ADS. Combined LWR-MOX plus ADS. For the different scenarios nuclide inventories, waste amounts, costs, additional electricity production etc have been assessed. As a general conclusion it was found that BWR is more efficient for burning plutonium in MOX fuel than PWR. The difference is approximately 10%. Furthermore the BWR produces about 10% less americium inventory. An ADS reactor park can theoretically in an ideal case burn (transmute) 99% of the transuranium isotopes. The duration of such a scenario heavily depends on the interim time needed for cooling the spent fuel before reprocessing. Assuming 10 years for cooling of nuclear fuel from ADS, the duration will be at least 200 years under optimistic technical assumptions. The development and use of advanced pyro-processing with an interim cooling time of only

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-01

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

  13. Development of self-interrogation neutron resonance densitometry (SINRD) to measure U-235 and Pu-239 content in a PWR spent fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Lafleur, Adrienne M [Los Alamos National Laboratory; Charlton, William S [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory

    2009-01-01

    The use of Self-Interrogation Neutron Resonance Densitometry (SINRD) to measure the {sup 235}U and {sup 239}Pu content in a PWR spent fuel assembly was investigated via Monte Carlo N-Particle eXtended transport code (MCNPX) simulations. The sensitivity of SINRD is based on using the same fissile materials in the fission chambers as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n, f) reaction peaks in fission chamber. These simulations utilize the {sup 244}Cm spontaneous fission neutrons to self-interrogate the fuel pins. The amount of resonance absorption of these neutrons in the fuel can be measured using {sup 235}U and {sup 239}Pu fission chambers placed adjacent to the assembly. We used ratios of different fission chambers to reduce the sensitivity of the measurements to extraneous material present in fuel. The development of SINRD to measure the fissile content in spent fuel is of great importance to the improvement of nuclear safeguards and material accountability. Future work includes the use of this technique to measure the fissile content in FBR spent fuel and heavy metal product from reprocessing methods.

  14. Antineutrino monitoring of spent nuclear fuel

    CERN Document Server

    Brdar, Vedran; Kopp, Joachim

    2016-01-01

    Military and civilian applications of nuclear energy have left a significant amount of spent nuclear fuel over the past 70 years. Currently, in many countries world wide, the use of nuclear energy is on the rise. Therefore, the management of highly radioactive nuclear waste is a pressing issue. In this letter, we explore antineutrino detectors as a tool for monitoring and safeguarding nuclear waste material. We compute the flux and spectrum of antineutrinos emitted by spent nuclear fuel elements as a function of time, and we illustrate the usefulness of antineutrino detectors in several benchmark scenarios. In particular, we demonstrate how a measurement of the antineutrino flux can help to re-verify the contents of a dry storage cask in case the monitoring chain by conventional means gets disrupted. We then comment on the usefulness of antineutrino detectors at long-term storage facilities such as Yucca mountain. Finally, we put forward antineutrino detection as a tool in locating underground "hot spots" in ...

  15. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Spent nuclear fuel project integrated schedule plan

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. A present status for dry storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K. S.; Lee, J. C.; Park, H. Y.; Seo, K. S

    2003-04-01

    National policy for management of a spent nuclear fuel does not establish in Korea yet. A storage capacity of a storage pool that is to store the spent nuclear fuel will be exceeded an amount of accumulation from the first Woljin nuclear power plant in 2007. Therefore it is necessary that dry storage facility is secured to store safely the spent nuclear fuel on site of the nuclear power plant until national policy for a back-end spent nuclear fuel cycle is established. In order to store safely spent nuclear fuel, it is important that the present status and technology on dry storage of spent nuclear fuel is looked over. Therefore, the present status on dry storage of spent nuclear fuel was analyzed so as to develop dry storage system and choose a proper dry storage method domestic.

  18. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  19. French nuclear plants PWR vessel integrity assessment and life management

    Energy Technology Data Exchange (ETDEWEB)

    Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)

    2001-07-01

    The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

  20. Spent Nuclear Fuel Alternative Technology Risk Assessment

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Report on interim storage of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Spent nuclear fuel project product specification

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Material accountancy measurement techniques in dry-powdered processing of nuclear spent fuels.

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, S. F.

    1999-03-24

    The paper addresses the development of inductively coupled plasma-mass spectrometry (ICPMS), thermal ionization-mass spectrometry (TIMS), alpha-spectrometry, and gamma spectrometry techniques for in-line analysis of highly irradiated (18 to 64 GWD/T) PWR spent fuels in a dry-powdered processing cycle. The dry-powdered technique for direct elemental and isotopic accountancy assay measurements was implemented without the need for separation of the plutonium, uranium and fission product elements in the bulk powdered process. The analyses allow the determination of fuel burn-up based on the isotopic composition of neodymium and/or cesium. An objective of the program is to develop the ICPMS method for direct fissile nuclear materials accountancy in the dry-powdered processing of spent fuel. The ICPMS measurement system may be applied to the KAERI DUPIC (direct use of spent PWR fuel in CANDU reactors) experiment, and in a near-real-time mode for international safeguards verification and non-proliferation policy concerns.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

    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.

  5. Spent Nuclear Fuel Vibration Integrity Study

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-01

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

  6. Characterization plan for Hanford spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Spent nuclear fuel project technical databook

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Advanced waste forms from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ackerman, J.P.; McPheeters, C.C.

    1995-12-31

    More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed.

  9. Design of a Prototype Differential Die-Away Instrument Proposed for Swedish Spent Nuclear Fuel Characterization

    Science.gov (United States)

    Martinik, Tomas; Henzl, Vladimir; Grape, Sophie; Jansson, Peter; Swinhoe, Martyn T.; Goodsell, Alison V.; Tobin, Stephen J.

    2016-06-01

    As part of the United States (US) Department of Energy's Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) project, the traditional Differential Die-Away (DDA) method that was originally developed for waste drum assay has been investigated and modified to provide a novel application to characterize or verify spent nuclear fuel (SNF). Following the promising, yet largely theoretical and simulation based, research of physics aspects of the DDA technique applied to SNF assay during the early stages of the NGSI-SF project, the most recent effort has been focused on the practical aspects of developing the first fully functional and deployable DDA prototype instrument for spent fuel. As a result of the collaboration among US research institutions and Sweden, the opportunity to test the newly proposed instrument's performance with commercial grade SNF at the Swedish Interim Storage Facility (Clab) emerged. Therefore the design of this instrument prototype has to accommodate the requirements of the Swedish regulator as well as specific engineering constrains given by the unique industrial environment. Within this paper, we identify key components of the DDA based instrument and we present methodology for evaluation and the results of a selection of the most relevant design parameters in order to optimize the performance for a given application, i.e. test-deployment, including assay of 50 preselected spent nuclear fuel assemblies of both pressurized (PWR) as well as boiling (BWR) water reactor type.

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

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H.K.

    1981-09-01

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

  11. Sensitivity and parametric evaluations of significant aspects of burnup credit for PWR spent fuel packages

    Energy Technology Data Exchange (ETDEWEB)

    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. Proving test on the seismic reliability of nuclear power plant: PWR reactor containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi; Yoshikawa, Teiichi; Ohno, Tokue; Yoshikawa, Eiji.

    1989-01-01

    Seismic reliability proving tests of nuclear power plant facilities are carried out by the Nuclear Power Engineering Test Center, using the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry. In 1982, the seismic reliability proving test of a PWR containment vessel was conducted using a test component of reduced scale 1/3.7. As a result of this test, the test component proved to have structural soundness against earthquakes, and at the same time its stable function was proved by leak tests which were carried out before and after the vibration test. In 1983, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. The seismic analysis and evaluation on the actual containment vessel were then performed using these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed.

  13. Sensitivity Analysis of Depletion Parameters for Heat Load Evaluation of PWR Spent Fuel Storage Pool

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Young; Lee, Un Chul [Seoul National University, Seoul (Korea, Republic of)

    2011-12-15

    As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.

  14. Mitsubishi PWR nuclear fuel with advanced design features

    Energy Technology Data Exchange (ETDEWEB)

    Kaua Goe, Toshiy Uki; Nuno kawa, Koi Chi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2008-10-15

    In the last few decades, the global warming has been a big issue. As the breakthrough in this crisis, advanced operations of the water reactor such as higher burn up, longer cycle, and up rating could be effective ways. From this viewpoint, Mitsubishi Heavy Industries (MHI) has developed the fuel for burn up extension, whose assembly burn-up limit is 55GWd/t(A), with the original and advanced designs such as corrosion resistant cladding material MDA, and supplied to Japanese PWR utilities. On the other hand, MHI intends to supply more advanced fuel assemblies not only to domestic market but to the global market. Actually MHI has submitted the application for standard design certification of USA . Advanced Pressurized Water Reactor on Jan. 2nd 2008. The fuel assembly for US APWR is 17x17 type with active fuel length of 14ft, characterized with three features, to {sup E}nhance Fuel Economy{sup ,} {sup E}nable Flexible Core Operation{sup ,} and to {sup I}mprove Reliability{sup .} MHI has also been conducting development activities for more advanced products, such as 70GWd/t(A) burn up limit fuel with cladding, guide thimble and spacer grid made from M-MDATM alloy that is new material with higher corrosion resistance, such as 12ft and 14ft active length fuel, such as fuel with countermeasure against grid fretting, debris fretting, and IRI. MHI will present its activities and advanced designs.

  15. Microbiology of spent nuclear fuel storage basins.

    Science.gov (United States)

    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.

  16. Spent Nuclear Fuel (SNF) Project Product Specification

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  18. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    Energy Technology Data Exchange (ETDEWEB)

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

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

    Science.gov (United States)

    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.

  1. Storage and Reprocessing of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    Addressing the problem of waste, especially high-level waste (HLW), is a requirement of the nuclear fuel cycle that cannot be ignored. We explore the two options employed currently, long-term storage and reprocessing.

  2. Railroad transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wooden, D.G.

    1986-03-01

    This report documents a detailed analysis of rail operations that are important for assessing the risk of transporting high-level nuclear waste. The major emphasis of the discussion is towards ''general freight'' shipments of radioactive material. The purpose of this document is to provide a basis for selecting models and parameters that are appropriate for assessing the risk of rail transportation of nuclear waste.

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies Phase I Study

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Y S; Sitaraman, S

    2008-12-24

    A novel methodology to detect diversion of spent fuel from Pressurized Water Reactors (PWR) has been developed in order to address a long unsolved safeguards verification problem for international safeguards community such as International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers. Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed and validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. A concept was developed to build a practical field device, Partial Defect Detector (PDET), which will be completely portable and will use standard radiation measuring devices already in use at the IAEA. The use of the device will not require any information provided

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

    CERN Document Server

    Jonkmans, G; Jewett, C; Thompson, M

    2012-01-01

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

  6. Alternative Measuring Approaches in Gamma Scanning on Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sihm Kvenangen, Karen

    2007-06-15

    In the future, the demand for energy is predicted to grow and more countries plan to utilize nuclear energy as their source of electric energy. This gives rise to many important issues connected to nuclear energy, such as finding methods that can verify that the spent nuclear fuel has been handled safely and used in ordinary power producing cycles as stated by the operators. Gamma ray spectroscopy is one method used for identification and verification of spent nuclear fuel. In the specific gamma ray spectroscopy method called gamma scanning the gamma radiation from the fission products Cs-137, Cs-134 and Eu-154 are measured in a spent fuel assembly. From the results, conclusions can be drawn about the fuels characteristics. This degree project examines the possibilities of using alternative measuring approaches when using the gamma scanning method. The focus is on examining how to increase the quality of the measured data. How to decrease the measuring time as compared with the present measuring strategy, has also been investigated. The main part of the study comprises computer simulations of gamma scanning measurements. The simulations have been validated with actual measurements on spent nuclear fuel at the central interim storage, Clab. The results show that concerning the quality of the measuring data the conventional strategy is preferable, but with other starting positions and with a more optimized equipment. When focusing on the time aspect, the helical measuring strategy can be an option, but this needs further investigation.

  7. Spent nuclear fuel project design basis capacity study

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Dissolution of spent nuclear fuel in carbonate-peroxide solution

    Science.gov (United States)

    Soderquist, Chuck; Hanson, Brady

    2010-01-01

    This study shows that spent UO2 fuel can be completely dissolved in a room temperature carbonate-peroxide solution apparently without attacking the metallic Mo-Tc-Ru-Rh-Pd fission product phase. In parallel tests, identical samples of spent nuclear fuel were dissolved in nitric acid and in an ammonium carbonate, hydrogen peroxide solution. The resulting solutions were analyzed for strontium-90, technetium-99, cesium-137, europium-154, plutonium, and americium-241. The results were identical for all analytes except technetium, where the carbonate-peroxide dissolution had only about 25% of the technetium that the nitric acid dissolution had.

  10. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  11. Common cause evaluations in applied risk analysis of nuclear power plants. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Taniguchi, T.; Ligon, D.; Stamatelatos, M.

    1983-04-01

    Qualitative and quantitative approaches were developed for the evaluation of common cause failures (CCFs) in nuclear power plants and were applied to the analysis of the auxiliary feedwater systems of several pressurized water reactors (PWRs). Key CCF variables were identified through a survey of experts in the field and a review of failure experience in operating PWRs. These variables were classified into categories of high, medium, and low defense against a CCF. Based on the results, a checklist was developed for analyzing CCFs of systems. Several known techniques for quantifying CCFs were also reviewed. The information provided valuable insights in the development of a new model for estimating CCF probabilities, which is an extension of and improvement over the Beta Factor method. As applied to the analysis of the PWR auxiliary feedwater systems, the method yielded much more realistic values than the original Beta Factor method for a one-out-of-three system.

  12. The Source Term Calculation and Analysis of PWR Spent Fuel%压水堆乏燃料源项计算与分析

    Institute of Scientific and Technical Information of China (English)

    苏卓; 邹树梁; 于涛; 谢金森

    2011-01-01

    In spent fuel reprocessing plant,the highly radioactive environment will have a certain radiation damage on monitoring equipments.Therefore,the using life of the equipments will be affected and the system reliability reduced.In this article,we use ORIGEN2 program to calculate the components of PWR fuel,and obtain a group of data about the important radionuclides in spent fuel components,such like radioactivity,photon energy spectra etc.The calculated results are accurate and credible,and it can provide initial source term data for shield design of electronic monitoring equipment in the first-side processing of spent fuel.%乏燃料后处理车间的高放射性环境会对监测设备产生一定的辐照损伤,影响其使用寿命,降低系统可靠性.本文使用ORIGEN2程序对压水堆燃料组件进行计算,得出一组乏燃料组件中重要核素的放射性活度、光子能谱等数据,计算结果准确可信,可为乏燃料首端处理中电子监测设备的屏蔽设计提供初始源项数据.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, Rodney C.

    2003-09-14

    The successful disposal of spent nuclear fuel (SNF) is one of the most serious challenges to the successful completion of the nuclear fuel cycle and the future of nuclear power generation. In the United States, 21 percent of the electricity is generated by 107 commercial nuclear power plants (NPP), each of which generates 20 metric tons of spent nuclear fuel annually. In 1996, the total accumulation of spent nuclear fuel was 33,700 metric tons of heavy metal (MTHM) stored at 70 sites around the country. The end-of-life projection for current nuclear power plants (NPP) is approximately 86,000 MTHM. In the proposed nuclear waste repository at Yucca Mountain over 95% of the radioactivity originates from spent nuclear fuel. World-wide in 1998, approximately 130,000 MTHM of SNF have accumulated, most of it located at 236 NPP in 36 countries. Annual production of SNF is approximately 10,000 MTHM, containing about 100 tons of ''reactor grade'' plutonium. Any reasonable increase in the proportion of energy production by NPP, i.e., as a substitute for hydrocarbon-based sources of energy, will significantly increase spent nuclear fuel production. Spent nuclear fuel is essentially UO{sub 2} with approximately 4-5 atomic percent actinides and fission product elements. A number of these elements have long half-lives hence, the long-term behavior of the UO{sub 2} is an essential concern in the evaluation of the safety and risk of a repository for spent nuclear fuel. One of the unique and scientifically most difficult aspects of the successful disposal of spent nuclear fuel is the extrapolation of short-term laboratory data (hours to years) to the long time periods (10{sup 3} to 10{sup 5} years) as required by the performance objectives set in regulations, i.e. 10 CFR 60. The direct verification of these extrapolations or interpolations is not possible, but methods must be developed to demonstrate compliance with government regulations and to satisfy the

  14. Estimating Source Terms for Diverse Spent Nuclear Fuel Types

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Spent nuclear fuel canister storage building conceptual design report

    Energy Technology Data Exchange (ETDEWEB)

    Swenson, C.E. [Westinghouse Hanford Co., Richland, WA (United States)

    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.

  17. Spent nuclear fuel project design basis capacity study

    Energy Technology Data Exchange (ETDEWEB)

    Cleveland, K.J.

    1998-07-22

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

  18. Review of partitioning proposals for spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bowersox, D.F.

    1976-07-01

    The initial phase of a study about recovery of valuable fission products from spent nuclear fuels has been to review various partitioning proposals. This report briefly describes the aqueous Purex process, the salt transport process, melt refining, fluoride volatility process, and gravimetric separations. All these processes appear to be possible technically, but further research will be necessary to determine which are most feasible. This review includes general recommendations for experimental research and development of several partitioning options.

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

    Energy Technology Data Exchange (ETDEWEB)

    Shen, E.J., Westinghouse Hanford

    1996-12-06

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

  20. Standard guide for drying behavior of spent nuclear fuel

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

  1. Analysis of nuclear characteristics and fuel economics for PWR core with homogeneous thorium fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joo, H. K.; Noh, J. M.; Yoo, J. W.; Song, J. S.; Kim, J. C.; Noh, T. W

    2000-12-01

    The nuclear core characteristics and economics of an once-through homogenized thorium cycle for PWR were analyzed. The lattice code, HELIOS has been qualified against BNL and B and W critical experiments and the IAEA numerical benchmark problem in advance of the core analysis. The infinite multiplication factor and the evolution of main isotopes with fuel burnup were investigated for the assessment of depletion charateristics of thorium fuel. The reactivity of thorium fuel at the beginning of irradiation is smaller than that of uranium fuel having the same inventory of {sup 235}U, but it decrease with burnup more slowly than in UO{sub 2} fuel. The gadolinia worth in thorium fuel assembly is also slightly smaller than in UO{sub 2} fuel. The inventory of {sup 233}U which is converted from {sup 232}Th is proportional to the initial mass of {sup 232}Th and is about 13kg per one tones of initial heavy metal mass. The followings are observed for thorium fuel cycle compared with UO{sub 2} cycle ; shorter cycle length, more positive MTC at EOC, more negative FTC, similar boron worth and control rod. Fuel economics of thorium cycle was analyzed by investigating the natural uranium requirements, the separative work requirements, and the cost for burnable poison rods. Even though less number of burnable poison rods are required in thorium fuel cycle, the costs for the natural uranium requirements and the separative work requirements are increased in thorium fuel cycle. So within the scope of this study, once through cycle concept, homogenized fuel concept, the same fuel management scheme as uranium cycle, the thorium fuel cycle for PWR does not have any economic incentives in preference to uranium.

  2. Studies on spent nuclear fuel evolution during storage

    Energy Technology Data Exchange (ETDEWEB)

    Rondinella, V.V.; Wiss, T.A.G.; Papaioannou, D.; Nasyrow, R. [European Commission Joint Research Centre, Karlsruhe (Germany). Inst. for Transuranium Elements

    2015-07-01

    Initially conceived to last only a few decades (40 years in Germany), extended storage periods have now to be considered for spent nuclear fuel due to the expanding timeline for the definition and implementation of the disposal in geologic repository. In some countries, extended storage may encompass a timeframe of the order of centuries. The safety assessment of extended storage requires predicting the behavior of the spent fuel assemblies and the package systems over a correspondingly long timescale, to ensure that the mechanical integrity and the required level of functionality of all components of the containment system are retained. Since no measurement of ''old'' fuel can cover the ageing time of interest, spent fuel characterization must be complemented by studies targeting specific mechanisms that may affect properties and behavior of spent fuel during extended storage. Tests conducted under accelerated ageing conditions and other relevant simulations are useful for this purpose. During storage, radioactive decay determines the overall conditions of spent fuel and generates heat that must be dissipated. Alpha-decay damage and helium accumulation are key processes affecting the evolution of properties and behavior of spent fuel. The radiation damage induced by a decay event during storage is significantly lower than that caused by a fission during in-pile operation: however, the duration of the storage is much longer and the temperature levels are different. Another factor potentially affecting the mechanical integrity of spent fuel rods during storage and handling / transportation is the behavior of hydrogen present in the cladding. At the Institute for Transuranium Elements, part of the Joint Research Centre of the European Commission, spent fuel alterations as a function of time and activity are monitored at different scales, from the microstructural level (defects and lattice parameter swelling) up to macroscopic properties such as

  3. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor [Zagreb Univ. (Croatia). Faculty of Electrical Engineering and Computing

    2013-12-15

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Spent Nuclear Fuel Option Study on Hybrid Reactor for Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    DUPIC nuclear fuel can be used in hybrid reactor by compensation of subcritical level through (U-10Zr) fuel. Energy production performance of Hyb-WT with DUPIC is grateful because it has high EM factor and performs waste transmutation at the same time. However, waste transmutation performance should be improved by different fissile fuel instead of (U-10Zr) fuel. SNF (Spent Nuclear Fuel) disposal is one of the problems in the nuclear industry. FFHR (Fusion-Fission Hybrid Reactor) is one of the most attractive option on reuse of SNF as a waste transmutation system. Because subcritical system like FFHR has some advantages compared to critical system. Subcritical systems have higher safety potential than critical system. Also, there is suppressed excess reactivity at BOC (Beginning of Cycle) in critical system, on the other hand there is no suppressed reactivity in subcritical system. Our research team could have designed FFHR for waste transmutation; Hyb-WT. Various researches have been conducted on fuel and coolant option for optimization of transmutation performance. However, Hyb-WT has technical disadvantage. It is required fusion power (Pfus) which is the key design parameter in FFHR is increased for compensation of decreasing subcritical level. As a result, structure material integrity is damaged under high irradiation condition by increasing Pfus. Also, deep burn of reprocessed SNF is limited by weakened integrity of structure material. Therefore, in this research, SNF option study will be conducted on DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactor) fuel, TRU fuel and DUPIC + TRU mixed fuel for optimization of Hyb-WT performance. Goal of this research is design check for low required fusion power and high waste transmutation. In this paper, neutronic analysis is conducted on Hyb-WT with DUPIC nuclear fuel. When DUPIC nuclear fuel is loaded in fast neutron system, supplement fissile materials need to be loaded together for compensation of low criticality

  6. Spent nuclear fuel for disposal in the KBS-3 repository

    Energy Technology Data Exchange (ETDEWEB)

    Grahn, Per; Moren, Lena; Wiborgh, Maria

    2010-12-15

    The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility. The report provides input to the assessment of the long-term safety, SR-Site as well as to the operational safety report, SR-Operation. The report presents the spent fuel to be deposited, and the requirements on the handling and selection of fuel assemblies for encapsulation that follows from that it shall be deposited in the KBS-3 repository. An overview of the handling and a simulation of the encapsulation and the resulting canisters to be deposited are presented. Finally, the initial state of the encapsulated spent nuclear fuel is given. The initial state comprises the radionuclide inventory and other data required for the assessment of the long-term safety

  7. Radiological consequences of accidents during disposal of spent nuclear fuel in a deep borehole

    Energy Technology Data Exchange (ETDEWEB)

    Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

    2013-07-15

    . The estimated maximum annual effective dose from one canister containing one PWR element that leaks due to damages created in conjunction with the accident is 5 mSv, primarily from {sup 137}Cs and {sup 107m}Ag. For one canister containing two BWR elements, the corresponding dose will be slightly less than 4 mSv. These doses are higher than the average background exposure of the Swedish population that amounts to about 3 mSv/year. Damage of multiple canisters will increase dose proportionally. The Swedish Radiation Safety Authority has issued regulations stipulating that the risk for harmful effects in conjunction with radioactive waste disposal should be less than 10{sup -6} per year. This corresponds to an annual effective dose of 1. 10{sup -5} Sv. In order for a facility for disposal in deep boreholes to meet this criterion, the probability of an accident in which one canister containing one PWR element is damaged at the time of the accident must be lower than 0.26 %, corresponding to 3{Chi} 10{sup -5} per disposal hole. In accidents involving damaging of a canister, a need to handle contaminated borehole mud may arise. Calculations in the current study indicate that such contaminated mud should be handled in tanks with extra shielding. It is concluded that necessary preparedness for accidents of the type described above is an obvious point of concern in any future planning of a facility for disposal of spent nuclear fuel in deep boreholes.

  8. Towards implementation of spent nuclear fuel management in Finland

    Energy Technology Data Exchange (ETDEWEB)

    Eero, Patrakka

    2007-07-01

    The final disposal of spent fuel from the Finnish nuclear power plants will be implemented by Posiva Oy, a company owned jointly by the two nuclear power producers Fortum Oyj and Teollisuuden Voima Oy. Preparations for nuclear waste management were started already in the 1970s. Potential sites for the disposal of spent fuel were screened in the 1980s, followed by detailed site investigations in the 1990s. In May 2001, the Finnish Parliament ratified the Decision-in-Principle that was a prerequisite for the selection of Olkiluoto as the site of the final disposal facility. The disposal project has progressed to the next stage -- constructing an underground characterization facility, known as ONKALO, at Olkiluoto, which has also been designed to serve as an access route to the repository when constructed. Work on the entire final disposal project is progressing so that disposal can commence in 2020. ONKALO will be used to obtain further information for the repository design, and it will also enable final disposal technology to be tested under actual conditions. Once ONKALO has been completed, work will start on building the encapsulation plant and final disposal repository in the 2010s. (auth)

  9. 78 FR 61401 - Entergy Nuclear Operations, Inc.; Big Rock Point; Independent Spent Fuel Storage Installation

    Science.gov (United States)

    2013-10-03

    ... Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001..., and 10 CFR part 50, allows ENO to possess and store spent nuclear fuel at the permanently shutdown and... Director, Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety...

  10. 78 FR 77606 - Security Requirements for Facilities Storing Spent Nuclear Fuel

    Science.gov (United States)

    2013-12-24

    ... COMMISSION 10 CFR Parts 72 and 73 RIN 3150-AI78 Security Requirements for Facilities Storing Spent Nuclear... requirements for storing spent nuclear fuel (SNF) in an independent spent fuel storage installation (ISFSI), and for storing SNF and/or high-level radioactive waste (HLW) in a monitored retrievable storage...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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.

  13. Development of dry storage technology of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Maruoka, Kunio [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Nuclear Energy Systems Engineering Center; Murakami, Kazuo; Yokoyama, Takeshi; Natsume, Tomohiro; Irino, Mitsuhiro

    1998-07-01

    The increasing demand for storage of spent fuel assemblies generated by commercial nuclear power plants is the urgent subject to solve. The dry storage system is as economically more advantageous than the pool storage system, and so, Mitsubishi Heavy Industries, Ltd. has developed the metal storage cask suited to small and medium storage capacity under 2000MTU - 3000MTU. For large scale capacity, the new `Mitsubishi Vault Storage System` has been developed, and it provides a safe and economical solution. Technical study concerning cooling ability was performed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  15. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  16. A Review on Sabotage against Transportation of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sungyeol; Lim, Jihwan [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    This report assesses the risk of routine transportation including cask response to an impact or fire accidents. In addition, we have still found the non-negligible difference among the studies for scenarios, approaches, and data. In order to evaluate attack cases on the same basis and reflect more realistic situations, at this moment, it is worthwhile to thoroughly review and analyze the existing studies and to suggest further development directions. In Section 2, we compare scenarios of terror attacks against spent fuel storage and transportation. Section 3 compares target scenarios, capabilities, and limitations of assessment methods. In addition, we collect and compare modeling data used for previous studies to analyze gaps and uncertainties in the existing studies. According to the long term management strategy for spent fuels in Korea, they will be transported from the spent fuel pools in each nuclear power plant to the central interim storage facility. The government should not be the only ones contributing to this dialogue. This dialogue that needs to happen should work both ways, with the government presenting their information and statistics and the public relaying their concerns for the government to review.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

    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.

  19. Spent nuclear fuel recycling with plasma reduction and etching

    Science.gov (United States)

    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.

  20. Physical modeling of spent-nuclear-fuel container

    Directory of Open Access Journals (Sweden)

    Wang Liping

    2012-11-01

    Full Text Available A new physical simulation model was developed to simulate the casting process of the ductile iron heavy section spent-nuclear-fuel container. In this physical simulation model, a heating unit with DR24 Fe-Cr-Al heating wires was used to compensate the heat loss across the non-natural surfaces of the sample, and a precise and reliable casting temperature controlling/monitoring system was employed to ensure the thermal behavior of the simulated casting to be similar to the actual casting. Also, a mould system was designed, in which changeable mould materials can be used for both the outside and inside moulds for different applications. The casting test was carried out with the designed mould and the cooling curves of central and edge points at different isothermal planes of the casting were obtained. Results show that for most isothermal planes, the temperature control system can keep the temperature differences within 6 ℃ between the edge points and the corresponding center points, indicating that this new physical simulation model has high simulation accuracy, and the mould developed can be used for optimization of casting parameters of spent-nuclear-fuel container, such as composition of ductile iron, the pouring temperature, the selection of mould material and design of cooling system. In addition, to maintain the spheroidalization of the ductile iron, the force-chilling should be used for the current physical simulation to ensure the solidification of casting in less than 2 h.

  1. Spent nuclear fuel recycling with plasma reduction and etching

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. A study on the direct use of spent PWR fuel in CANDU reactors -Fuel management and safety analysis-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Boh Wook; Choi, Hang Bok; Lee, Yung Wook; Cho, Jae Sun; Huh, Chang Wook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The reference DUPIC fuel composition was determined based on the reactor safety, thermal-hydraulics, economics, and refabrication aspects. The center pin of the reference DUPIC fuel bundle is poisoned with natural dysprosium. The worst LOCA analysis has shown that the transient power and heat deposition of the reference DUPIC core are the same as those of natural uranium CANDU core. The intra-code comparison has shown that the accuracy of DUPIC physics code system is comparable to the current CANDU core design code system. The sensitivity studies were performed for the refuelling schemes of DUPIC core and the 2-bundle shift refuelling scheme was selected as the standard refuelling scheme of the DUPIC core. The application of 4-bundle shift refuelling scheme will be studied in parallel as the auto-refuelling method is improved and the reference core parameters of the heterogeneous DUPIC core are defined. The heterogeneity effect was analyzed in a preliminary fashion using 33 fuel types and the random loading strategy. The refuelling simulation has shown that the DUPIC core satisfies the current CANDU 6 operating limits of channel and bundle power regardless of the fuel composition heterogeneity. The 33 fuel types used in the heterogeneity analysis was determined based on the initial enrichment and discharge burnup of the PWR fuel. 90 figs, 62 tabs, 63 refs. (Author).

  3. Challenges in spent nuclear fuel final disposal:conceptual design models

    Institute of Scientific and Technical Information of China (English)

    Mukhtar Ahmed RANA

    2008-01-01

    The disposal of spent nuclear fuel is a long-standing issue in nuclear technology. Mainly, UO2 and metallic U are used as a fuel in nuclear reactors. Spent nuclear fuel contains fission products and transuranium elements, which would remain radioactive for 104 to 108 years. In this brief communication, essential concepts and engineering elements related to high-level nuclear waste disposal are described. Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste. Notions of physical and chemical barriers to contain nuclear waste are highlightened. Concerns regarding integrity, self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed. The question of retrievability of spent nuclear fuel after disposal is considered.

  4. Nuclear Data Library Effects on Fast to Thermal Flux Shapes Around PWR Control Rod Tips

    Science.gov (United States)

    Vasiliev, A.; Ferroukhi, H.; Zhu, T.; Pautz, A.

    2014-04-01

    The development of a high-fidelity computational scheme to estimate the accumulated fluence at the tips of PWR control rods (CR) has been initiated at the Paul Scherrer Institut (PSI). Both the fluence from high-energy (E>1 MeV) neutrons as well as for the thermal range (E<0.625 eV) are required as these affect the CR integrity through stresses/strains induced by coupled clad embrittlement / absorber swelling phenomena. The concept of the PSI scheme under development is to provide from validated core analysis models, the volumetric neutron source to a full core MCNPX model that is then used to compute the neutron fluxes. A particular aspect that needs scrutiny is the ability of the MCNPX-based calculation methodology to accurately predict the flux shapes along the control rod surfaces, especially for fully withdrawn CRs. In that case, the tip is located a short distance above the core/reflector interface and since this situation corresponds to a large part of reactor operation, the accumulated fluence will highly depend on the achieved calculation accuracy and precision in this non-fueled zone. The objective of the work presented in this paper is to quantify the influence of nuclear data on the calculated fluxes at the CR tips by (1) conducting a systematic comparison of modern neutron cross-section libraries, including JENDL-4.0, JEFF-3.1.1 and ENDF/B-VII.0, and (2) by quantifying the uncertainties in the neutron flux calculations with the help of available neutron cross-section variances/covariances data. For completeness, the magnitude of these nuclear data-based uncertainties is also assessed in relation to the influence from other typical sources of modeling uncertainties/biases.

  5. Seismic analysis of spent nuclear fuel storage racks

    Energy Technology Data Exchange (ETDEWEB)

    Shah, S.J.; Biddle, J.R.; Bennett, S.M.; Schechter, C.B. [Framatome Cogema Fuels, Lynchburg, VA (United States); Harstead, G.A. [Harstead Engineering Associates, Inc., Old Tappan, NJ (United States); Marquet, F. [ATEA/FRAMATOME, Carquefou (France)

    1996-06-01

    In many nuclear power plants, existing storage racks are being replaced with high-density racks to accommodate the increasing inventory of spent fuel. In the hypothetical design considered here, the high-density arrangement of fuel assemblies, or consolidated fuel canisters, is accomplished through the use of borated stainless steel (BSS) plates acting as neutron absorbers. No structural benefit from the BSS is assumed. This paper describes the methods used to perform seismic analysis of high density spent fuel storage racks. The sensitivity of important parameters such as the effect of variation of coefficients of friction between the rack legs and the pool floor and fuel loading conditions (consolidated and unconsolidated) are also discussed in the paper. Results of this study are presented. The high-density fuel racks are simply supported by the pool floor with no structural connections to adjacent racks or to the pool walls or floor. Therefore, the racks are free standing and may slide and tip. Several time history, nonlinear, seismic analyses are required to account for variations in the coefficient of friction, rack loading configuration, and the type of the seismic event. This paper presents several of the mathematical models usually used. Friction cannot be precisely predicted, so a range of friction coefficients is assumed. The range assumed for the analysis is 0.2 to 0.8. A detailed model representing a single rack is used to evaluate the 3-D loading effects. This model is a controlling case for the stress analysis. A 2-D multi-rack model representing a row of racks between the spent fuel pool walls is used to evaluate the change in gaps between racks. The racks are normally analyzed for the fuel loading conditions of consolidated, full, empty, and half-loaded with fuel assemblies.

  6. Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

    Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.

  7. NEPA implementation: The Department of Energy`s program to manage spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shipler, D.B.

    1994-05-01

    The Department of Energy (DOE) is implementing the National Environmental Protection Act (NEPA) in its management of spent nuclear fuel. The DOE strategy is to address the short-term safety concerns about existing spent nuclear fuel, to study alternatives for interim storage, and to develop a long-range program to manage spent nuclear fuel. This paper discusses the NEPA process, the environmental impact statements for specific sites as well as the overall program, the inventory of DOE spent nuclear fuel, the alternatives for managing the fuel, and the schedule for implementing the program.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Advantages on dry interim storage for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, L.S. [Centro Tecnologico da Marinha em Sao Paulo, Av. Professor Lineu Prestes 2468, 05508-900 Sao Paulo (Brazil); Rzyski, B.M. [IPEN/ CNEN-SP, 05508-000 Sao Paulo (Brazil)]. e-mail: romanato@ctmsp.mar.mil.br

    2006-07-01

    When the nuclear fuel lose its ability to efficiently create energy it is removed from the core reactor and moved to a storage unit waiting for a final destination. Generally, the spent nuclear fuel (SNF) remains inside concrete basins with water within the reactors facility for the radioactive activity decay. Water cools the generated heat and shields radioactivity emissions. After some period of time in water basins the SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing installations, or still wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet installations, depending on the method adopted by the nuclear power plant or other plans of the country. In many SNF wet storage sites the capacity can be fulfilled very quickly. If so, additional area or other alternative storage system should be given. There are many options to provide capacity increase in the wet storage area, but dry storages are worldwide preferred since it reduces corrosion concerns. In the wet storage the temperature and water purity should be constantly controlled whereas in the dry storage the SNF stands protected in specially designed canisters. Dry interim storages are practical and approved in many countries especially that have the 'wait and see' philosophy (wait to see new technologies development). This paper shows the advantages of dry interim storages sites in comparison with the wet ones and the nowadays problems as terrorism. (Author)

  10. End crop sealing method for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yamanaka, Kiyoshi

    1998-12-04

    End crops of spent nuclear fuels and glass materials are sealed in a corrosion and heat resistant vessel having an upper portion opened, and the corrosion and heat resistant vessel is heated from outside to melt the glass materials and they are solidified to form glass solidification products in which the end crops and radioactive materials deposited on the end crops are sealed. Then, the opened portion is closed. Since the end crops and radioactive materials deposited on end crops are sealed as glass solidification products in the vessel, sealing property for radioactive materials can be enhanced. Accordingly, radioactive materials can be prevented from transferring to the outside upon storing wastes or processing them into underground. In addition, since a large quantity of end crops can be filled into the corrosion and heat resistant vessel to form high level wastes, the space for the storage of wastes and processing facilities can be reduced. (T.M.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Tsutomu; Komatsu, Kazunori; Takahashi, A. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    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.

  13. Optimization of thermal efficiency of nuclear central power like as PWR; Otimizacao da eficiencia termica de uma usina nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lapa, Nelbia da Silva

    2005-10-15

    The main purpose of this work is the definition of operational conditions for the steam and power conservation of Pressurized Water Reactor (PWR) plant in order to increase its system thermal efficiency without changing any component, based on the optimization of operational parameters of the plant. The thermal efficiency is calculated by a thermal balance program, based on conservation equations for homogeneous modeling. The circuit coefficients are estimated by an optimization tool, allowing a more realistic thermal balance for the plans under analysis, as well as others parameters necessary to some component models. With the operational parameter optimization, it is possible to get a level of thermal efficiency that increase capital gain, due to a better relationship between the electricity production and the amount of fuel used, without any need to change components plant. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Applying fast calorimetry on a spent nuclear fuel calorimeter

    Energy Technology Data Exchange (ETDEWEB)

    Liljenfeldt, Henrik [Swedish Nuclear Fuel and Waste Management (Sweden); Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Uppsala Univ. (Sweden)

    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.

  16. Validation of spent nuclear fuel nuclide composition data using percentage differences and detailed analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Cheol [Chung-Ang Univ., Seoul (Korea, Republic of). School of Energy Systems Engineering

    2017-06-15

    Nuclide composition data of spent nuclear fuels are important in many nuclear engineering applications. In reactor physics, nuclear reactor design requires the nuclide composition and the corresponding cross sections. In analyzing the radiological health effects of a severe accident on the public and the environment, the nuclide composition in the reactor inventory is among the important input data. Nuclide composition data need to be provided to analyze the possible environmental effects of a spent nuclear fuel repository. They will also be the basis for identifying the origin of unidentified spent nuclear fuels or radioactive materials.

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

    Science.gov (United States)

    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.

  18. Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

    Energy Technology Data Exchange (ETDEWEB)

    Mayers, J.B.; Soth, L.G.

    1978-04-01

    The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction.

  19. 76 FR 35137 - Vulnerability and Threat Information for Facilities Storing Spent Nuclear Fuel and High-Level...

    Science.gov (United States)

    2011-06-16

    ... Storing Spent Nuclear Fuel and High-Level Radioactive Waste AGENCY: U.S. Nuclear Regulatory Commission... Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than...-based security regulations for Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) storage...

  20. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  1. Extended calculations of OECD/NEA phase II-C burnup credit criticality benchmark problem for PWR spent fuel transport cask by using MCNP-4B2 code and JENDL-3.2 library

    Energy Technology Data Exchange (ETDEWEB)

    Kuroishi, Takeshi; Hoang, Anh Tuan; Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    The reactivity effect of the asymmetry of axial burnup profile in burnup credit criticality safety is studied for a realistic PWR spent fuel transport cask proposed in the current OECD/NEA Phase II-C benchmark problem. The axial burnup profiles are simulated in 21 material zones based on in-core flux measurements varying from strong asymmetry to more or less no asymmetry. Criticality calculations in a 3-D model have been performed using the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculation conditions are determined with consideration of the axial fission source convergence. Calculations are carried out not only for cases proposed in the benchmark but also for additional cases assuming symmetric burnup profile. The actinide-only approach supposed for first domestic introduction of burnup credit into criticality evaluation is also considered in addition to the actinide plus fission product approach adopted in the benchmark. The calculated results show that k{sub eff} and the end effect increase almost linearly with increasing burnup axial offset that is defined as one of typical parameters showing the intensity of axial burnup asymmetry. The end effect is more sensitive to the asymmetry of burnup profile for the higher burnup. For an axially distributed burnup, the axial fission source distribution becomes strongly asymmetric as its peak shifts toward the top end of the fuel's active zone where the local burnup is less than that of the bottom end. The peak of fission source distribution becomes higher with the increase of either the asymmetry of burnup profile or the assembly-averaged burnup. The conservatism of the assumption of uniform axial burnup based on the actinide-only approach is estimated quantitatively in comparison with the k{sub eff} result calculated with experiment-based strongest asymmetric axial burnup profile with the actinide plus fission product approach. (author)

  2. Deep geological disposal system development; mechanical structural stability analysis of spent nuclear fuel disposal canister under the internal/external pressure variation

    Energy Technology Data Exchange (ETDEWEB)

    Kwen, Y. J.; Kang, S. W.; Ha, Z. Y. [Hongik University, Seoul (Korea)

    2001-04-01

    This work constitutes a summary of the research and development work made for the design and dimensioning of the canister for nuclear fuel disposal. Since the spent nuclear fuel disposal emits high temperature heats and much radiation, its careful treatment is required. For that, a long term(usually 10,000 years) safe repository for spent fuel disposal should be securred. Usually this repository is expected to locate at a depth of 500m underground. The canister construction type introduced here is a solid structure with a cast iron insert and a corrosion resistant overpack, which is designed for spent nuclear fuel disposal in a deep repository in the crystalline bedrock, which entails an evenly distributed load of hydrostatic pressure from undergroundwater and high pressure from swelling of bentonite buffer. Hence, the canister must be designed to withstand these high pressure loads. Many design variables may affect the structural strength of the canister. In this study, among those variables array type of inner baskets and thicknesses of outer shell and lid and bottom are tried to be determined through the mechanical linear structural analysis, thicknesses of outer shell is determined through the nonlinear structural analysis, and the bentonite buffer analysis for the rock movement is conducted through the of nonlinear structural analysis Also the thermal stress effect is computed for the cast iron insert. The canister types studied here are one for PWR fuel and another for CANDU fuel. 23 refs., 60 figs., 23 tabs. (Author)

  3. INEL integrated spent nuclear fuel consolidation task team report

    Energy Technology Data Exchange (ETDEWEB)

    Henry, R.N.; Clark, J.H.; Chipman, N.A. [and others

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-01

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

  5. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    Energy Technology Data Exchange (ETDEWEB)

    N. E. Pettit

    2001-07-13

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident.

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

    Energy Technology Data Exchange (ETDEWEB)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J. [INEEL (US); Brey, R.F. [ISU (US); 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Rob L [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  8. 78 FR 66858 - Waste Confidence-Continued Storage of Spent Nuclear Fuel

    Science.gov (United States)

    2013-11-07

    ...; ] NUCLEAR REGULATORY COMMISSION 10 CFR Part 51 RIN 3150-AJ20 Waste Confidence--Continued Storage of Spent Nuclear Fuel AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule; extension of comment period. SUMMARY: On September 13, 2013, the U. S. Nuclear Regulatory Commission (NRC) published for public...

  9. A robotized surface workstation for manipulation, filling and closing of packaging containers for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bartos, Pavel [FITE a.s., Ostrava-Marianske Hory (Czech Republic); Haladova, Petra [Robotsystem, LLC/Moravian Research, LLC, Ostrava-Moravska (Czech Republic); Otcenasek, Petr

    2016-01-15

    Options for the handling of spent nuclear fuel are described and a packaging cask for an underground repository is presented as also a robotic surface workplace for the repository. The potential for the closing the nuclear fuel cycle is discussed. Currently, a team of Czech experts is developing a project of fully robotic technology for manipulation and storage of packaging casks for spent nuclear fuel in host rock of underground repository.

  10. Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method; Sondas para inspecao de trocadores de calor instalados em usinas nucleares tipo PWR pelo metodo de correntes parasitas

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alonso F.O. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Enghenharia Mecanica]. E-mail: kauzz21@yahoo.com; Alencar, Donizete A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mail: daa@cdtn.br

    2007-07-01

    From all non destructive examination methods usable to perform integrity evaluation of critical equipment installed at nuclear power plants (NPP), eddy current test (ET) may be considered the most important one, when examining heat exchangers. For its application, special probes and reference calibration standards are employed. In pressurized water reactor (PWR) NPPs, a particularly critical equipment is the steam generator (SG), a huge heat exchanger that contains thousands of U-bend thin wall tubes. Due to its severe working conditions (pressure and temperature), that component is periodically examined by means of ET. In this paper a revision of the operating fundamentals of the main ET probes, used to perform SG inspections is presented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    CLEVELAND, K.J.

    2000-08-17

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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. Environmental Impact Statement. March 2011. Interim storage, encapsulation and final disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    2011-07-01

    An Environmental Impact Statement (EIS) shall be prepared and submitted along with applications for permissibility and a licence under the Environmental Code and a licence under the Nuclear Activities Act for new nuclear facilities. This Environmental Impact Statement has been prepared by Svensk Kaernbraenslehantering AB (the Swedish Nuclear Fuel and Waste Management Co, SKB) to be included in the licence applications for continued operation of Clab (central interim storage facility for spent nuclear fuel) in Simpevarp in Oskarshamn Municipality and construction and operation of facilities for encapsulation (integrated with Clab) and final disposal of spent nuclear fuel in Forsmark in Oesthammar Municipality

  16. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Modelling of thermally driven groundwater flow in a facility for disposal of spent nuclear fuel in deep boreholes

    Energy Technology Data Exchange (ETDEWEB)

    Marsic, Nico; Grundfelt, Bertil [Kemakta Konsult AB, Stockholm (Sweden)

    2013-09-15

    In this report calculations are presented of buoyancy driven groundwater flow caused by the emission of residual heat from spent nuclear fuel deposited in deep boreholes from the ground surface in combination with the natural geothermal gradient. This work has been conducted within SKB's programme for evaluation of alternative methods for final disposal of spent nuclear fuel. The basic safety feature of disposal of spent nuclear fuel in deep boreholes is that the groundwater at great depth has a higher salinity, and hence a higher density, than more superficial groundwater. The result of this is that the deep groundwater becomes virtually stagnant. The study comprises analyses of the effects of different inter-borehole distances as well as the effect of different permeabilities in the backfill and sealing materials in the borehole and of different shapes of the interface between fresh and saline groundwater. The study is an update of a previous study published in 2006. In the present study, the facility design proposed by Sandia National Laboratories has been studied. In this design, steel canisters containing two BWR elements or one PWR element are stacked on top of each other between 3 and 5 kilometres depth. In order to host all spent fuel from the current Swedish nuclear programme, about 80 such holes are needed. The model used in this study comprises nine boreholes spaced 100 metres alternatively 50 metres apart in a 3{Chi}3 matrix. In one set of calculations the salinity in the groundwater was assumed to increase from zero above 700 metres depth to 10% by weight at 1500 metres depth and below. In another set, a sharper salinity gradient was applied in which the salinity increased from 0 to 10% between 1400 and 1500 metres depth. A geothermal gradient of 16 deg C/km was applied. The heat output from the spent fuel was assumed to decrease by time in manner consistent with the radioactive decay in the fuel. When the inter-borehole distance decreased from

  18. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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

  19. Calculation of source term in spent PWR fuel assemblies for dry storage and shipping cask design; Calculo de los terminos fuente de combustibles irradiados PWR para el diseno de contenedores de almacenamiento y transporte

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, J. L.; Lopez, J.

    1986-07-01

    Using the ORIGEN-2 Coda, the decay heat and neutron and photon sources for an irradiated PWR fuel element have been calculated. Also, parametric studies on the behaviour of the magnitudes with the burn-up, linear heat power and irradiation and cooling times were performed. Finally, a comparison between our results and other design calculations shows a good agreement and confirms the validity of the used method. (Author) 6 refs.

  20. Power Flattening and Rejuvenation of PWR Spent Fuel Blanket for Hybrid Fusion-Fission Reactor%功率展平的压水堆乏燃料发电包层中子学初步研究

    Institute of Scientific and Technical Information of China (English)

    马续波; 陈义学; 王继亮; 王悦; 韩静茹; 陆道纲

    2011-01-01

    The hybrid fusion-fission reactor has advantages of breeding of the nuclear fuel and transmutation of the long-life nuclear waste and having inherent safety. Meanwhile, the engineering and technological demand of hybrid reactor is significantly reduced comparing with that of pure fusion reactor. A generating electricity blanket concept using the PWR spent fuel directly was proposed, which was based on ITER parameter level achieved. Different volume fractions of the fuel in blanket enabled to realize a power flattening in the fissile zone. The results show that the peak-to-average power factor becomes less than no power flattening, and the output power of the fuel zone raises more than 21. 7%. At the end of the operation, the maximum fuel enrichment is 5. 23%. The blanket is feasible from the neutronics viewpoint.%聚变裂变混合堆在增殖核燃料、嬗变长寿命核废料及固有安全性等方面具有较大优势,同时,它比纯聚变堆在工程及技术方面要求低,因此较聚变堆更易实现.本工作基于目前国际聚变实验堆(ITER)所能达到的技术水平,提出一种直接利用乏燃料进行发电的聚变裂变混合堆包层概念,利用在不同位置放置不同乏燃料体积分数的方法对燃料增殖区实现了功率展平.计算结果表明:功率展平后的包层功率不均匀系数更小,且包层中燃料区的能量输出要比不展平情况下的能量输出高约21.7%.燃料富集度到运行末期最大可达5.23%.从中子学角度初步论证了该包层的可行性.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Spent nuclear fuel project high-level information management plan

    Energy Technology Data Exchange (ETDEWEB)

    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

  4. Standard PWR for Italy

    Energy Technology Data Exchange (ETDEWEB)

    Negroni, A.; Velona, F. (Ente Nazionale per l' Energia Elettrica, Rome (Italy))

    1983-03-01

    A description is given of the general design for the standard PWR which will be used in the seven to eight nuclear power stations provided for in the Italian national energy plan. Special features to meet Italian conditions include double containment and a common foundation mat for the reactor, auxiliary and fuel buildings.

  5. Spent Fuel Source Term Calculation of Daya Bay Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    XU; Zhi-long; WAN; Hai-xia; LI; Long; WU; Xiao-chun; SHAO; Jing; LIU; Li-li; ZHANG; Jing

    2013-01-01

    The spent fuel of nuclear power plant should be transported to reprocessing plant for reprocessing after reserving for a period of time.Before that,safety analysis and environmental impact assessment should be carried on to the transportation process,which need radioactive source term calculation and analysis.The task of Daya Bay Nuclear Power Plant spent fuel source term calculation includes estimation of

  6. Spent nuclear fuel project detonation phenomena of hydrogen/oxygen in spent fuel containers

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, T.D.

    1996-09-30

    Movement of Spent N Reactor fuels from the Hanford K Basins near the Columbia River to Dry interim storage facility on the Hanford plateau will require repackaging the fuel in the basins into multi-canister overpacks (MCOs), drying of the fuel, transporting the contained fuel, hot conditioning, and finally interim storage. Each of these functions will be accomplished while the fuel is contained in the MCOs by several mechanisms. The principal source of hydrogenand oxygen within the MCOs is residual water from the vacuum drying and hot conditioning operations. This document assesses the detonation phenomena of hydrogen and oxygen in the spent fuel containers. Several process scenarios have been identified that could generate detonation pressures that exceed the nominal 10 atmosphere design limit ofthe MCOS. Only 42 grams of radiolized water are required to establish this condition.

  7. State of the art report: Underwater examination techniques for spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Yong Bum

    1997-06-01

    In these days, much efforts are being put to increase the final discharge burnup of PWR fuels. Therefore, the necessity of the inspection of irradiated nuclear fuels assembly during the the refueling outage is greatly increased to evaluate the safe operation and soundness of fuel assemblies in their next cycles in core, and apply the results for safe operation and effective core management. The necessity to evaluate the irradiation performance of indigenous nuclear fuels pushes the relative researchers to the development of on-site fuel inspection techniques and devices which can perform the underwater inspection and measurement of irradiated nuclear fuels during refueling outage. To ensure the technologies, the status of in situ underwater fuel inspection techniques and equipment were investigated and reviewed. Those information provides the fuel inspection capability to evaluate and certificate the performance and integrity of nuclear fuels which leads to the safe operation of NPP. (author). 49 refs., 52 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Development of nuclear spent fuel Maritime transportation scenario

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Min; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2014-08-15

    Spent fuel transportation of South Korea is to be conducted through near sea because it is able to ship a large amount of the spent fuel far from the public comparing to overland transportation. The maritime transportation is expected to be increased and its risk has to be assessed. For the risk assessment, this study utilizes the probabilistic safety assessment (PSA) method and the notions of the combined event. Risk assessment of maritime transportation of spent fuel is not well developed in comparison with overland transportation. For the assessment, first, the transportation scenario should be developed and categorized. Categories are assorted into the locations, release aspects and exposure aspects. This study deals with accident that happens on voyage and concentrated on ship-ship collision. The collision accident scenario is generated with event tree analysis. The scenario will be exploited for the maritime transportation risk model which includes consequence and accident probability.

  10. Advantages of dry hardened cask storage over wet storage for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Romanato, Luiz Sergio, E-mail: romanato@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil). Dept. da Qualidade

    2011-07-01

    Pools are generally used to store and maintain spent nuclear fuel assemblies for cooling, after removed from reactors. After three to five years stored in the pools, spent fuel can be reprocessed or sent to a final disposition in a geological repository and handled as radioactive waste or sent to another site waiting for future solution. Spent fuel can be stored in dry or wet installations, depending on the method adopted by the nuclear plant. If this storage were exclusively wet, at the installation decommissioning in the future, another solution for storage will need to be found. Today, after a preliminary cooling, the spent fuel assemblies can be removed from the pool and sent to dry hardened storage installations. This kind of storage does not need complex radiation monitoring and it is safer than wet storage. Brazil has two nuclear reactors in operation, a third reactor is under construction and they use wet spent fuel storage . Dry hardened casks use metal or both metal and concrete for radiation shielding and they are safe, especially during an earthquake. An earthquake struck Japan on March 11, 2011 damaging Fukushima Daiichi nuclear power plant. The occurrence of earthquakes in Brazil is very small but dry casks can resist to other events, including terrorist acts, better than pools. This paper shows the advantages of dry hardened cask storage in comparison with the wet storage (water pools) for spent nuclear fuel. (author)

  11. Analysis of preliminary design concept of stainless steel container for disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, K.S.; Ku, J.H.; Park, J.H.; Choi, J.W. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-02-01

    This report represents the structural, thermal and radiation shielding analysis of the basic concepts of the disposal container, that could accommodate PWR and CANDU fuels of which physical dimensions and shapes are quite different each other, with respect to the emplacement modes. Basic concepts of the disposal containers for the vertical horehole and the drift emplacement modes are proposed with their maximum allowable thermal loading. Appropriate thickness of the container to withstand the expected external pressure in the underground repository system was delivered by the structural analyses. The thermal analysis of the container containing spent fuels showed that the internal maximum temperatures of all container concepts did not reach the constraint values. Radiation dose rate from the container with 10cm thickness wall were also less than the established constraint value. (author). 9 refs., 33 figs., 12 tabs.

  12. Site selection - siting of the final repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    2011-03-15

    SKB has selected Forsmark as the site for the final repository for spent nuclear fuel. The site selection is the end result of an extensive siting process that began in the early 1990s. The strategy and plan for the work was based on experience from investigations and development work over a period of more than ten years prior to then. This document describes the siting work and SKB's choice of site for the final repository. It also presents the information on which the choice was based and the reasons for the decisions made along the way. The document comprises Appendix PV to applications under the Nuclear Activities Act and the Environmental Code for licences to build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to build and operate a final repository for spent nuclear fuel in Forsmark in Oesthammar Municipality

  13. Analysis of Spent Nuclear Fuel Imaging Using Multiple Coulomb Scattering of Cosmic Muons

    CERN Document Server

    Chatzidakis, Stylianos; Tsoukalas, Lefteri H

    2016-01-01

    Cosmic ray muons passing through matter lose energy from inelastic collisions with electrons and are deflected from nuclei due to multiple Coulomb scattering. The strong dependence of scattering on atomic number Z and the recent developments on position sensitive muon detectors indicate that multiple Coulomb scattering could be an excellent candidate for spent nuclear fuel imaging. Muons present significant advantages over existing monitoring and imaging techniques and can play a central role in monitoring nuclear waste and spent nuclear fuel stored in dense well shielded containers. The main purpose of this paper is to investigate the applicability of multiple Coulomb scattering for imaging of spent nuclear fuel dry casks stored within vertical and horizontal commercial storage dry casks. Calculations of muon scattering were performed for various scenarios, including vertical and horizontal fully loaded dry casks, half loaded dry casks, dry casks with one row of fuel assemblies missing, dry casks with one fu...

  14. Realistic bandwidth estimation in the theoretically predicted radionuclide inventory of PWR-UO2 spent fuel derived from reactor design and operating data

    Energy Technology Data Exchange (ETDEWEB)

    Fast, Ivan

    2017-06-01

    Nuclear energy for power generation produces heat-generating high- and intermediate level radioactive waste (HLW and ILW) for which a safe solution for the handling and disposal has to be found. Currently, many European countries consider the final disposal of HLW and ILW in deep geological formations as the most preferable option. In Germany the main stream of HLW and ILW include spent fuel assemblies from nuclear power plants (NPPs), the vitrified waste and compacted metallic waste of the fuel assembly structural parts originate from reprocessing plants. An important task that occurs within the framework of the Product Quality Control (PQC) of nuclear waste is the assessment of the compliance of any reprocessed waste product inventory with the prescribed limits for each relevant radionuclide (RN). The PQC task is to verify the required quality and safety of nuclear waste prior to transportation to a German repository and to avert the disposal of non-conform waste packages. The verification is usually based on comparing the declared radionuclide inventory of the waste with the presumed or expected composition, which is estimated, based on the known history of the waste and its processing. The difficulty of such estimations for radioactive components from nuclear fuel assemblies is that reactor design parameters and operating histories can have a significant influence on the nuclide inventory of any individual fuel assembly. Thus, knowledge of these parameters is a key issue to determine the realistic concentration ranges, or bandwidths, of the radionuclide inventory. As soon as a governmental decision on the construction of a high-level waste repository will be made, comprehensive radionuclide inventories of the wastes assigned for the deposition will be required. The list of final repository relevant radionuclide is based on the safety assessment for this particular repository, thus it is likely to comprise more-or-less the same radionuclides that need to be

  15. Methodology to evaluate the crack growth rate by stress corrosion cracking in dissimilar metals weld in simulated environment of PWR nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Paula, Raphael G.; Figueiredo, Celia A.; Rabelo, Emerson G., E-mail: raphaelmecanica@gmail.com, E-mail: caf@cdtn.br, E-mail: egr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2013-07-01

    Inconel alloys weld metal is widely used to join dissimilar metals in nuclear reactors applications. It was recently observed failures of weld components in plants, which have triggered an international effort to determine reliable data on the stress corrosion cracking behavior of this material in reactor environment. The objective of this work is to develop a methodology to determine the crack growth rate caused by stress corrosion in Inconel alloy 182, using the specimen (Compact Tensile) in simulated PWR environment. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  17. Decommissioning of the BR3 PWR[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Massaut, V.

    1998-07-01

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific program, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1997 are summarized.

  18. Simulation of alpha dose for predicting radiolytic species at the surface of spent nuclear fuel pellets

    OpenAIRE

    Becker Frank; Kienzler Bernhard

    2014-01-01

    In many countries, spent nuclear fuel is considered as a waste form to be disposed of in underground disposal. Under deep host rock conditions, a reducing environment prevails. In the case of water contact, long-term radionuclide release from the fuel depends on dissolution processes of the UO2 matrix. The dissolution rate of irradiated UO2 is controlled by oxidizing processes facilitated by dissolved species formed by alpharadiolysis of water in contact with spent nuc...

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  20. Near-field chemistry of the spent nuclear fuel repository; Kemialliset vuorovaikutukset kaeytetyn ydinpolttoaineen loppusijoitustilan laehialueella

    Energy Technology Data Exchange (ETDEWEB)

    Kumpulainen, H.; Lehikoinen, J.; Muurinen, A.; Ollila, K. [VTT Chemical Technology, Espoo (Finland). Industrial Physics

    1998-07-01

    Factors affecting near-field chemistry of the spent nuclear fuel repository as well as the involved mutual interactions are described on the basis of literature. The most important processes in the near-field (spent-fuel, canister and bentonite) are presented. The related examples on near-field chemistry models shed light on the extensive problematics of near-field chemistry. (authors) 80 refs.

  1. Roles and effects of pyroprocessing for spent nuclear fuel management in South Korea

    OpenAIRE

    Ahn, J

    2014-01-01

    Republic of Korea (ROK) changed its spent nuclear fuel policy from the once-through usage and direct disposal to a total system approach that includes pyroprocessing, sodium-cooled fast reactors, and a two-tier geological repository to achieve a breakthrough for domestic deadlock situation and thus enable sustainable utilization of nuclear power, but caused disagreement in the bilateral negotiation with the United States (US) for the Nuclear Cooperation Agreement. Analysis has revealed that t...

  2. Spent nuclear fuel discharges from US reactors 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-05-05

    This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

  3. 75 FR 45167 - Notice of Public Workshop on a Potential Rulemaking for Spent Nuclear Fuel Reprocessing Facilities

    Science.gov (United States)

    2010-08-02

    ... civilian nuclear power globally and close the nuclear fuel cycle through reprocessing spent fuel and... Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor... regulations in 10 CFR Part 171, ``Annual Fees for Reactor Licenses and Fuel Cycle Licenses and......

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Identification of unknown nuclear material

    Energy Technology Data Exchange (ETDEWEB)

    Nicolaou, G. [University of Thrace, Department of Electrical and Computer Engineering, Laboratory of Nuclear Technology, Kimmerria Campus, 67100 Xanthi (Greece)

    2010-07-01

    Aim: provenance determination of unknown nuclear material: - demonstrated for spent nuclear fuel; - information sought for unknown: fuel type, reactor type where fuel was irradiated, final burnup; Using an isotopic finger-printing method: - U, Pu or Pu isotopics or fission products; - simulations of fuel evolution during irradiation, using ORIGEN; - multivariate statistical tools. Fuel considered: simulated commercial spent fuel for a range of burnups: - PWR UO{sub 2} 3.1% and 3.5% {sup 235}U, - PWR thermal MOX, - BWR UO{sub 2} 3.2% {sup 235}U, - CANDU-N natural U, - CANDU-S UO{sub 2} 3.2% {sup 235}U, - fast Reactor MOX; simulated commercial spent fuel for a range of burnups: - PWR UO{sub 2} 3.1% and 3.5% {sup 235}U, - PWR thermal MOX, - BWR UO{sub 2} 3.2% {sup 235}U, - CANDU-N natural U, - CANDU-S UO{sub 2} 3.2% {sup 235}U, - fast Reactor MOX; 'unknown' spent fuel: - PWR 1: UO{sub 2} 3.1% {sup 235}U (26 GWd/t), - PWR 2: UO{sub 2} 3.1% {sup 235}U (32 GWd/t). Procedures: U, Pu or Pu isotopic compositions or fission products: - isotopic composition of unknown spent fuel, - simulated for commercial spent fuel from a range of nuclear power reactors {yields} comparison of compositions through factor analysis {yields} unknown has the provenance of the commercial spent fuel with which it exhibits the most similar composition. In conclusion: different reactor-fuel types well resolved; fuel and reactor type accurately predicted; burnup predicted to within 5% of declared; different reactor-fuel types. (authors)

  6. CONSIDERATIONS REGARDING ROK SPENT NUCLEAR FUEL MANAGEMENT OPTIONS

    Directory of Open Access Journals (Sweden)

    CHAIM BRAUN

    2013-08-01

    Pyroprocessing is but an enabling technology on the path leading to fissile material recycling and burning in future SFRs. In this regard we discuss two SFR options under development in the U.S.: the Super Prism and the Travelling Wave Reactor (TWR. We note that the U.S. is further along in reactor development than the ROK. The ROK though has acquired more experience, recently in investigating fuel recycling options for SFRs. We thus call for two complementary joint R&D project to be conducted by U.S. and ROK scientists. One leading to the development of a demonstration centralized away-from-reactors spent fuel storage facility. The other involve further R&D on a combined SFR-fuel cycle complex based on the reactor and fuel cycle options discussed in the paper.

  7. Redox Chemistry in Radiation Induced Dissolution of Spent Nuclear Fuel : from Elementary Reactions to Predictive Modeling

    OpenAIRE

    Roth, Olivia

    2008-01-01

    The focus of this doctoral thesis is the redox chemistry involved in radiation induced oxidative dissolution of spent nuclear fuel and UO2 (as a model substance for spent nuclear fuel). It is shown that two electron oxidants are more efficient than one electron oxidants in oxidative dissolution of UO2 at low oxidant concentrations. Furthermore, it is shown that H2O2 is the only oxidant that has to be taken into account in radiation induced dissolution of UO2 under deep repository conditions (...

  8. Chemical Reactivity Testing for the National Spent Nuclear Fuel Program. Quality Assurance Project Plan

    Energy Technology Data Exchange (ETDEWEB)

    Newsom, H.C.

    1999-01-24

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of Work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, QA-101PD, revision 1, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted.

  9. German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

    Directory of Open Access Journals (Sweden)

    A. Schwenk-Ferrero

    2013-01-01

    Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

  10. Existing Condition Analysis of Dry Spent Fuel Storage Technology

    Institute of Scientific and Technical Information of China (English)

    LI Ning; XU Lan; HAO Jian-sheng

    2016-01-01

    As in some domestic nuclear power plants, spent fuel pools near capacity, away-from-reactor type storage should be arranged to transfer spent fuel before the pool capacity is full and the plants can operate in safety. This study compares the features of wet and dry storage technology, analyzes the actualities of foreign dry storage facilities and then introduces structural characteristics of some foreign dry storage cask. Finally, a glance will be cast on the failure of away-from-reactor storage facilities of Pressurized Water Reactor(PWR)to meet the need of spent-fuel storage. Therefore, this study believes dry storage will be a feasible solution to the problem.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

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

  14. Storage facilities of spent nuclear fuel in dry for Mexican nuclear facilities; Instalaciones de almacenamiento de combustible nuclear gastado en seco para instalaciones nucleares mexicanas

    Energy Technology Data Exchange (ETDEWEB)

    Salmeron V, J. A.; Camargo C, R.; Nunez C, A.; Mendoza F, J. E.; Sanchez J, J., E-mail: juan.salmeron@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    In this article the relevant aspects of the spent fuel storage and the questions that should be taken in consideration for the possible future facilities of this type in the country are approached. A brief description is proposed about the characteristics of the storage systems in dry, the incorporate regulations to the present Nuclear Regulator Standard, the planning process of an installation, besides the approaches considered once resolved the use of these systems; as the modifications to the system, the authorization periods for the storage, the type of materials to store and the consequent environmental impact to their installation. At the present time the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) considers the possible generation of two authorization types for these facilities: Specific, directed to establish a new nuclear installation with the authorization of receiving, to transfer and to possess spent fuel and other materials for their storage; and General, focused to those holders that have an operation license of a reactor that allows them the storage of the nuclear fuel and other materials that they possess. Both authorizations should be valued according to the necessities that are presented. In general, this installation type represents a viable solution for the administration of the spent fuel and other materials that require of a temporary solution previous to its final disposal. Its use in the nuclear industry has been increased in the last years demonstrating to be appropriate and feasible without having a significant impact to the health, public safety and the environment. Mexico has two main nuclear facilities, the nuclear power plant of Laguna Verde of the Comision Federal de Electricidad (CFE) and the facilities of the TRIGA Reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) that will require in a future to use this type of disposition installation of the spent fuel and generated wastes. (Author)

  15. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix C, Savannah River Site Spent Nuclear Fuel Mangement Program

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The US Department of Energy (DOE) is engaged in two related decision making processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad programmatic decisions that will have applicability across the DOE complex and describes in detail the purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document, which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program, supports Volume 1 of the EIS. Following the introduction, Chapter 2 contains background information related to the SRS and the framework of environmental regulations pertinent to spent nuclear fuel management. Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement at the SRS, and summarizes their potential environmental consequences. Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel activities could affect. Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel management alternative and describes cumulative impacts. The chapter also contains information on unavoidable adverse impacts, commitment of resources, short-term use of the environment and mitigation measures.

  16. Simulations of H 2O 2 concentration profiles in the water surrounding spent nuclear fuel

    Science.gov (United States)

    Nielsen, Fredrik; Lundahl, Karin; Jonsson, Mats

    2008-01-01

    A simple mathematical model describing the hydrogen peroxide concentration profile in water surrounding a spent nuclear fuel pellet as a function of time has been developed. The water volume is divided into smaller elements, and the processes that affect hydrogen peroxide concentration are applied to each volume element. The model includes production of H 2O 2 from α-radiolysis, surface reaction between H 2O 2 and UO 2 and diffusion. Simulations show that the surface concentration of H 2O 2 increases fairly rapidly and approaches the steady-state concentration. The time to reach steady-state is sufficiently short to be neglected compared to the times of interest when simulating spent fuel dissolution under deep repository conditions. Consequently, the steady-state approach can be used to estimate the rate for radiation-induced spent nuclear fuel dissolution.

  17. Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

    CERN Document Server

    Poulson, D; Guardincerri, E; Morris, C L; Bacon, J D; Plaud-Ramos, K; Morley, D; Hecht, A

    2016-01-01

    Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This paper describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, the cask contents can be confirmed with high confidence in less than two days exposure. Similar results can be obtained by moving a smaller detector to view the cask from multiple angles.

  18. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 3, Calculated activity profiles of spent nuclear fuel assembly hardware for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. 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 that is also radioactive and required 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 Volume 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.

  19. Spent fuel assembly hardware: Characterization and 10 CFR 61 classification for waste disposal: Volume 2, Calculated activity profiles of spent nuclear fuel assembly hardware for pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Short, S.M.; Luksic, A.T.; Lotz, T.L.; Schutz, M.E.

    1989-06-01

    Consolidation of spent fuel is under active consideration as the US Department of Energy plans to dispose of spent fuel as required by the Nuclear Waste Policy Act of 1982. 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 present 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.

  20. A Historical Review of the Safe Transport of Spent Nuclear Fuel, Rev. 1

    Energy Technology Data Exchange (ETDEWEB)

    Connolly, Kevin J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pope, Ronald [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-01

    This report is a revision to M3 milestone M3FT-16OR090402028 for the former Nuclear Fuels Storage and Transportation Planning Project (NFST), “Safety Record of SNF Shipments.” The US Department of Energy (DOE) has since established the Office of Integrated Waste Management (IWM), which builds on the work begun by NFST, to develop an integrated waste management system for spent nuclear fuel (SNF), including the developm

  1. Literature on fabrication of tungsten for application in pyrochemical processing of spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Edstrom, C.M.; Phillips, A.G.; Johnson, L.D.; Corle, R.R.

    1980-10-11

    The pyrochemical processing of nuclear fuels requires crucibles, stirrers, and transfer tubing that will withstand the temperature and the chemical attack from molten salts and metals used in the process. This report summarizes the literature that pertains to fabrication (joining, chemical vapor deposition, plasma spraying, forming, and spinning) is the main theme. This report also summarizes a sampling of literature on molbdenum and the work previously performed at Argonne National Laboratory on other container materials used for pyrochemical processing of spent nuclear fuels.

  2. Neutron spectrometry at the interim storage facility for spent nuclear fuel

    CERN Document Server

    Králik, M; Studeny, J

    2002-01-01

    Dosimetric characteristics of neutron and photon components of mixed fields around casks for spent nuclear fuel have been determined at various places at the dry interim storage facility. The results obtained with metrological grade instruments were compared with data provided by usual survey meters for both neutrons and photons.

  3. Status of DOE efforts to renew acceptance of foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Head, C.R.

    1997-08-01

    This presentation summarizes the efforts being made by the Department of Energy to renew acceptance of spent nuclear fuel shipments from foreign research reactors. The author reviews the actions undertaken in this process in a fairly chronological manner, through the present time, as well as the development of an environmental impact statement to support the proposed actions.

  4. 76 FR 34007 - Draft Regulatory Basis for a Potential Rulemaking on Spent Nuclear Fuel Reprocessing Facilities

    Science.gov (United States)

    2011-06-10

    ...: NUREG-1909, a white paper authored by the Advisory Committee on Nuclear Waste and Materials, titled... waste through developing more sophisticated reprocessing technologies. During the Bush Administration... reprocessing spent fuel and deploying fast reactors to burn long-lived actinides. In response to these...

  5. Study of minimum-weight highway transporters for spent nuclear fuel casks: Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Hoess, J.A.; Drago, V.J.

    1989-05-01

    There are federal and state limits on the maximum tractor-trailer- payload combination and individual axle loads permissible on US highways. These can generally be considered as two sets, i.e., legal-weight and overweight limits. The number of individual shipments required will decrease as the capacity of the spent nuclear fuel cask increases. Thus, there is an incentive for identifying readily available minimum-weight tractors and trailers capable of safely and reliably transporting as large a cask as possible without exceeding the legal gross combination weight (GCW) of 80,000 lb or selected overweight GCW limit of 110,000 lb. This study identifies options for commercially available heavy-duty on-highway tractors and trailers for transporting proposed future loaded spent nuclear fuel casks. Loaded cask weights of 56,000 and 80,000 lb were selected as reference design points for the legal-weight and overweight transporters, respectively. The technical data on tractor and trailer characteristics obtained indicate that it is possible to develop a tractor-trailer combination, tailored for spent nuclear fuel transportation service, utilizing existing technology and commercially available components, capable of safely and reliably transporting 56,000 and 80,000-lb spent nuclear fuel casks without exceeding GCWs of 80,000 and 10,000 lb, respectively. 4 figs., 14 tabs.

  6. Fission product partitioning in aerosol release from simulated spent nuclear fuel

    NARCIS (Netherlands)

    Di Lemma, F.G.; Colle, J.Y.; Rasmussen, G.; Konings, R.J.M.

    2015-01-01

    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

  7. Comparison of nuclear data uncertainty propagation methodologies for PWR burn-up simulations

    CERN Document Server

    Diez, Carlos Javier; Hoefer, Axel; Porsch, Dieter; Cabellos, Oscar

    2014-01-01

    Several methodologies using different levels of approximations have been developed for propagating nuclear data uncertainties in nuclear burn-up simulations. Most methods fall into the two broad classes of Monte Carlo approaches, which are exact apart from statistical uncertainties but require additional computation time, and first order perturbation theory approaches, which are efficient for not too large numbers of considered response functions but only applicable for sufficiently small nuclear data uncertainties. Some methods neglect isotopic composition uncertainties induced by the depletion steps of the simulations, others neglect neutron flux uncertainties, and the accuracy of a given approximation is often very hard to quantify. In order to get a better sense of the impact of different approximations, this work aims to compare results obtained based on different approximate methodologies with an exact method, namely the NUDUNA Monte Carlo based approach developed by AREVA GmbH. In addition, the impact ...

  8. Near-term improvements for nuclear power plant control room annunciator systems. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rankin, W.L.; Duvernoy, E.G.; Ames, K.R.; Morgenstern, M.H.; Eckenrode, R.J.

    1983-04-01

    This report sets forth a basic design philosophy with its associated functional criteria and design principles for present-day, hard-wired annunciator systems in the control rooms of nuclear power plants. It also presents a variety of annunciator design features that are either necessary for or useful to the implementation of the design philosophy. The information contained in this report is synthesized from an extensive literature review, from inspection and analysis of control room annunciator systems in the nuclear industry and in related industries, and from discussions with a variety of individuals who are knowledgeable about annunciator systems, nuclear plant control rooms, or both. This information should help licensees and license applicants in improving their hard-wired, control room annunciator systems as outlined by NUREG-0700.

  9. Separation of the rare-earth fission product poisons from spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  10. Separation of the rare-earth fission product poisons from spent nuclear fuel

    Science.gov (United States)

    Christian, Jerry D.; Sterbentz, James W.

    2016-08-30

    A method for the separation of the rare-earth fission product poisons comprising providing a spent nuclear fuel. The spent nuclear fuel comprises UO.sub.2 and rare-earth oxides, preferably Sm, Gd, Nd, Eu oxides, with other elements depending on the fuel composition. Preferably, the provided nuclear fuel is a powder, preferably formed by crushing the nuclear fuel or using one or more oxidation-reduction cycles. A compound comprising Th or Zr, preferably metal, is provided. The provided nuclear fuel is mixed with the Th or Zr, thereby creating a mixture. The mixture is then heated to a temperature sufficient to reduce the UO.sub.2 in the nuclear fuel, preferably to at least to 850.degree. C. for Th and up to 600.degree. C. for Zr. Rare-earth metals are then extracted to form the heated mixture thereby producing a treated nuclear fuel. The treated nuclear fuel comprises the provided nuclear fuel having a significant reduction in rare-earths.

  11. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  12. 78 FR 40200 - Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...

    Science.gov (United States)

    2013-07-03

    ... COMMISSION Duke Energy Carolinas, LLC, Oconee Nuclear Station Units 1, 2, and 3; Independent Spent Fuel...) for an exemption request submitted by Duke Energy Carolinas, LLC, on August 13, 2012 for the Oconee Nuclear Station Independent Spent Fuel Storage Facility (ISFSI). ] ADDRESSES: Please refer to Docket...

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

    Energy Technology Data Exchange (ETDEWEB)

    McIntosh, Kathryn G., E-mail: kmcintosh@lanl.gov; Reilly, Sean D.; Havrilla, George J.

    2015-08-01

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

  14. Qualification of NEXUS/ANC Nuclear Design System for PWR Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Mayhue, Larry; Milanova, Radka; Huria, Harish; Zhang, Baocheng; Franceschini, Fausto; Ouisloumen, Mohamed [Westinghouse Electric Company, Pittsburgh, PA (United States); Mueller, Erwin; Forslun Guimaraes, Petri [Westinghouse Electric Company, Vaesteraas (Sweden)

    2008-07-01

    NEXUS is a new cross section and nuclear data generation system for core simulators developed by Westinghouse. This system generates once-through, full temperature range nuclear data for both PWRs and BWRs. The system has been implemented for PWRs in the NEXUS/ANC code system. A brief description of the methodology and the codes comprising this system is presented. The qualification for NEXUS/ANC has been completed and a summary of some of the results is presented for 10 plants and 45 cycles of operation. These results include startup data and at-power axial offset performance. Results for low temperature calculations are also presented. The NEXUS/ANC system includes new methodology to cover the operation of AP1000 plants including a new pin power recovery method and a method to capture the effects of control rod depletion. A brief summary of these methods is also presented. (authors)

  15. PWR Cross Section Libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, Carolyn [Texas A& M University; Ilas, Germina [ORNL

    2012-01-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% 235U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time.

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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

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

  17. Research and development for decontamination system of spent resin in Hanbit Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Gi Hong [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2015-12-15

    When reactor coolant leaks occur due to cracks of a steam generator tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000-7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In

  18. Research on advanced aqueous reprocessing of spent nuclear fuel: literature study

    Energy Technology Data Exchange (ETDEWEB)

    Van Hecke, K.; Goethals, P.

    2006-07-15

    The goal of the partitioning and transmutation strategy is to reduce the radiotoxicity of spent nuclear fuel to the level of natural uranium in a short period of time (about 1000 years) and thus the required containment period of radioactive material in a repository. Furthermore, it aims to reduce the volume of waste requiring deep geological disposal and hence the associated space requirements and costs. Several aqueous as well as pyrochemical separation processes have been developed for the partitioning of the long-lived radionuclides from the remaining of the spent fuel. This report aims to describe and compare advanced aqueous reprocessing methods.

  19. Effect of burn-up and high burn-up structure on spent nuclear fuel alteration

    Energy Technology Data Exchange (ETDEWEB)

    Clarens, F.; Gonzalez-Robles, E.; Gimenez, F. J.; Casas, I.; Pablo, J. de; Serrano, D.; Wegen, D.; Glatz, J. P.; Martinez-Esparza, A.

    2009-07-01

    In this report the results of the experimental work carried out within the collaboration project between ITU-ENRESA-UPC/CTM on spent fuel (SF) covering the period 2005-2007 were presented. Studies on both RN release (Fast Release Fraction and matrix dissolution rate) and secondary phase formation were carried out by static and flow through experiments. Experiments were focussed on the study of the effect of BU with two PWR SF irradiated in commercial reactors with mean burn-ups of 48 and 60 MWd/KgU and; the effect of High Burn-up Structure (HBS) using powdered samples prepared from different radial positions. Additionally, two synthetic leaching solutions, bicarbonate and granitic bentonite ground wa ter were used. Higher releases were determined for RN from SF samples prepared from the center in comparison with the fuel from the periphery. However, within the studied range, no BU effect was observed. After one year of contact time, secondary phases were observed in batch experiments, covering the SF surface. Part of the work was performed for the Project NF-PRO of the European Commission 6th Framework Programme under contract no 2389. (Author)

  20. Physical protection of nuclear facilities. Quarterly progress report, July--September 1978. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, L.D. (ed.)

    1979-01-01

    Major activities during the fourth quarter of FY78 included (1) the vital area analysis of operational reactors and characterization of the Standardized Nuclear Unit Power Plant System (SNUPPS), (2) the algorithm development of a new pathfinding computer code, (3) the completion of contractor-supported work for the component generic data base, (4) the refinement of tests related to human parameters modeling, and (5) the addition of improvements to and demonstration of the Safeguards Automated Facility Evaluation (SAFE), Safeguards Network Analysis Procedure (SNAP), and Fixed-Site Neutralization Model (FSNM) methodologies.

  1. Dose reduction and optimization studies (ALARA) at nuclear power facilities. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Baum, J.W.; Meinhold, C.B.

    1983-01-01

    Brookhaven National Laboratory (BNL) has been commissioned by the Nuclear Regulatory Commission (NRC) to study dose-reduction techniques and effectiveness of as low as reasonably achievable (ALARA) planning at LWR plants. These studies have the following objectives: identify high-dose maintenance tasks; identify dose-reduction techniques; examine incentives for dose reduction; evaluate cost-effectiveness and optimization of dose-reduction techniques; and compile an ALARA handbook on data, engineering modifications, cost-effectiveness calculations, and other information of interest to ALARA practioners.

  2. Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology

    Directory of Open Access Journals (Sweden)

    Eun-Young Choi

    2015-12-01

    Full Text Available The electrochemical reduction process has been used to reduce spent oxide fuel to a metallic form using pyroprocessing technology for a closed fuel cycle in combination with a metal-fuel fast reactor. In the electrochemical reduction process, oxides fuels are loaded at the cathode basket in molten Li2O–LiCl salt and electrochemically reduced to the metal form. Various approaches based on thermodynamic calculations and experimental studies have been used to understand the electrode reaction and efficiently treat spent fuels. The factors that affect the speed of the electrochemical reduction have been determined to optimize the process and scale-up the electrolysis cell. In addition, demonstrations of the integrated series of processes (electrorefining and salt distillation with the electrochemical reduction have been conducted to realize the oxide fuel cycle. This overview provides insight into the current status of and issues related to the electrochemical processing of spent nuclear fuels.

  3. Performance assessment of self-interrogation neutron resonance densitometry for spent nuclear fuel assay

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei, E-mail: huj1@ornl.gov [Reactor and Nuclear Systems Division, Oak Ridge National Laboratory, 1 Bethel Valley Road, PO Box 2008, MS-6172, Oak Ridge, TN 37831-6172 (United States); Tobin, Stephen J.; LaFleur, Adrienne M.; Menlove, Howard O.; Swinhoe, Martyn T. [Nuclear Engineering and Nonproliferation Division, Los Alamos National Laboratory (United States)

    2013-11-21

    Self-Interrogation Neutron Resonance Densitometry (SINRD) is one of several nondestructive assay (NDA) techniques being integrated into systems to measure spent fuel as part of the Next Generation Safeguards Initiative (NGSI) Spent Fuel Project. The NGSI Spent Fuel Project is sponsored by the US Department of Energy's National Nuclear Security Administration to measure plutonium in, and detect diversion of fuel pins from, spent nuclear fuel assemblies. SINRD shows promising capability in determining the {sup 239}Pu and {sup 235}U content in spent fuel. SINRD is a relatively low-cost and lightweight instrument, and it is easy to implement in the field. The technique makes use of the passive neutron source existing in a spent fuel assembly, and it uses ratios between the count rates collected in fission chambers that are covered with different absorbing materials. These ratios are correlated to key attributes of the spent fuel assembly, such as the total mass of {sup 239}Pu and {sup 235}U. Using count rate ratios instead of absolute count rates makes SINRD less vulnerable to systematic uncertainties. Building upon the previous research, this work focuses on the underlying physics of the SINRD technique: quantifying the individual impacts on the count rate ratios of a few important nuclides using the perturbation method; examining new correlations between count rate ratio and mass quantities based on the results of the perturbation study; quantifying the impacts on the energy windows of the filtering materials that cover the fission chambers by tallying the neutron spectra before and after the neutrons go through the filters; and identifying the most important nuclides that cause cooling-time variations in the count rate ratios. The results of these studies show that {sup 235}U content has a major impact on the SINRD signal in addition to the {sup 239}Pu content. Plutonium-241 and {sup 241}Am are the two main nuclides responsible for the variation in the count

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

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

  5. 400-MWe Consolidated Nuclear Steam System (CNSS). 1200-MWt Phase 2A interim studies. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The Phase 2A interim studies of the Consolidated Nuclear Steam System (CNSS) consisted of a number of separate task studies addressing the design concepts developed during the Phase 1 study reported in BAW--1445. The purpose of the interim studies was to better establish overall concept feasibility from both a hardware and economic standpoint, to make modification and additions to the design where appropriate, and to understand and reduce the technical risks in critical areas of the design. The work on these task studies included input from Barberton, Mt. Vernon, and the Alliance Research Center as well as United Engineers and Constructors (UE and C). The UE and C work was carried out under a separate DOE contract.

  6. Disposal of Steam Generators from Decommissioning of PWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Walberg, Mirko; Viermann, Joerg; Beverungen, Martin [GNS Gesellschaft fuer Nuklear-Service mbH, Hollestrasse 7A, 45127 Essen (Germany); Kemp, Lutz [Kernkraftwerk Stade GmbH and Co.oHG, Bassenflether Chaussee, 21683 Stade (Germany); Lindstroem, Anders [Studsvik Nuclear AB, SE-611 82 Nykoeping (Sweden)

    2008-07-01

    Amongst other materials remarkable amounts of radioactively contaminated or activated scrap are generated from the dismantling of Nuclear Power Plants. These scrap materials include contaminated pipework, fittings, pumps, the reactor pressure vessel and other large components, most of them are heat exchangers. Taking into account all commercial and technical aspects an external processing and subsequent recycling of the material might be an advantageous option for many of these components. The disposal of steam generators makes up an especially challenging task because of their measures, their weight and compared to other heat exchangers high radioactive inventory. Based on its experiences from many years of disposal of smaller components of NPP still in operation or under decommissioning GNS and Studsvik Nuclear developed a concept for disposal of steam generators, also involving experiences made in Sweden. The concept comprises transport preparations and necessary supporting documents, the complete logistics chain, steam generator treatment and the processing of arising residues and materials not suitable for recycling. The first components to be prepared, shipped and treated according to this concept were four steam generators from the decommissioning of the German NPP Stade which were removed from the plant and shipped to the processing facility during the third quarter of 2007. Although the plant had undergone a full system decontamination, due to the remaining contamination in a number of plugged tubes the steam generators had to be qualified as industrial packages, type 2 (IP-2 packages), and according to a special requirement of the German Federal Office for Radiation Protection a license for a shipment under special arrangement had to be applied for. The presentation gives an overview of the calculations and evidences required within the course of the IP-2 qualification, additional requirements of the competent authorities during the licensing procedure as

  7. Assessment of the risk of transporting spent nuclear fuel by truck

    Energy Technology Data Exchange (ETDEWEB)

    Elder, H.K.

    1978-11-01

    The assessment includes the risks from release of spent fuel materials and radioactive cask cavity cooling water due to transportation accidents. The contribution to the risk of package misclosure and degradation during normal transport was also considered. The results of the risk assessment have been related to a time in the mid-1980's, when it is projected that nuclear plants with an electrical generating capacity of 100 GW will be operating in the U.S. For shipments from reactors to interim storage facilities, it is estimated that a truck carrying spent fuel will be involved in an accident that would not be severe enough to result in a release of spent fuel material about once in 1.1 years. It was estimated that an accident that could result in a small release of radioactive material (primarily contaminated cooling water) would occur once in about 40 years. The frequency of an accident resulting in one or more latent cancer fatalities from release of radioactive materials during a truck shipment of spent fuel to interim storage was estimated to be once in 41,000 years. No accidents were found that would result in acute fatalities from releases of radioactive material. The risk for spent fuel shipments from reactors to reprocessing plants was found to be about 20% less than the risk for shipments to interim storage. Although the average shipment distance for the reprocessing case is larger, the risk is somewhat lower because the shipping routes, on average, are through less populated sections of the country. The total risk from transporting 180-day cooled spent fuel by truck in the reference year is 4.5 x 10/sup -5/ fatalities. An individual in the population at risk would have one chance in 6 x 10/sup 11/ of suffering a latent cancer fatality from a release of radioactive material from a truck carrying spent fuel in the reference year. (DLC)

  8. SACSESS – the EURATOM FP7 project on actinide separation from spent nuclear fuels

    Directory of Open Access Journals (Sweden)

    Bourg Stéphane

    2015-12-01

    Full Text Available Recycling of actinides by their separation from spent nuclear fuel, followed by transmutation in fast neutron reactors of Generation IV, is considered the most promising strategy for nuclear waste management. Closing the fuel cycle and burning long-lived actinides allows optimizing the use of natural resources and minimizing the long-term hazard of high-level nuclear waste. Moreover, improving the safety and sustainability of nuclear power worldwide. This paper presents the activities striving to meet these challenges, carried out under the Euratom FP7 collaborative project SACSESS (Safety of Actinide Separation Processes. Emphasis is put on the safety issues of fuel reprocessing and waste storage. Two types of actinide separation processes, hydrometallurgical and pyrometallurgical, are considered, as well as related aspects of material studies, process modeling and the radiolytic stability of solvent extraction systems. Education and training of young researchers in nuclear chemistry is of particular importance for further development of this field.

  9. Derivation of correction factor to be applied for calculated results of PWR fuel isotopic composition by ORIGEN2 code

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Nomura, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murazaki, Minoru [Tokyo Nuclear Service Inc., Tokyo (Japan); Mochizuki, Hiroki [The Japan Research Institute Ltd., Tokyo (Japan)

    2001-11-01

    For providing conservative PWR spent fuel compositions from the view point of nuclear criticality safety, correction factors applicable for result of burnup calculation by ORIGEN2 were evaluated. Its conservativeness was verified by criticality calculations using MVP. To calculate these correction factors, analyses of spent fuel isotopic composition data were performed by ORIGEN2. Maximum or minimum value of the ratio of calculation result to experimental data was chosen as correction factor. These factors are given to each set of fuel assembly and ORIGEN2 library. They could be considered as the re-definition of recommended isotopic composition given in Nuclear Criticality Safety Handbook. (author)

  10. Development of U-frame bending system for studying the vibration integrity of spent nuclear fuel

    Science.gov (United States)

    Wang, Hong; Wang, Jy-An John; Tan, Ting; Jiang, Hao; Cox, Thomas S.; Howard, Rob L.; Bevard, Bruce B.; Flanagan, Michelle

    2013-09-01

    A bending fatigue system developed to evaluate the response of spent nuclear fuel rods to vibration loads is presented. A U-frame testing setup is used for imposing bending loads on the fuel rod specimen. The U-frame setup consists of two rigid arms, side connecting plates to the rigid arms, and linkages to a universal testing machine. The test specimen's curvature is obtained through a three-point deflection measurement method. The tests using surrogate specimens with stainless steel cladding revealed increased flexural rigidity under unidirectional cyclic bending, significant effect of cladding-pellets bonding on the response of surrogate rods, and substantial cyclic softening in reverse bending mode. These phenomena may cast light on the expected response of a spent nuclear fuel rod. The developed U-frame system is thus verified and demonstrated to be ready for further pursuit in hot-cell tests.

  11. Calibration of burnup monitor of spent nuclear fuel installed at Rokkasho reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Oeda, Kaoru; Matoba, Masaru; Wakabayashi, Genichiro [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Naito, Hirofumi; Hirota, Masanari [Nuclear Fuel Industries Ltd., Tokyo (Japan); Morizaki, Hidetoshi; Kumanomido, Hironori; Natsume, Koichiro [Toshiba Corp., Tokyo (Japan)

    2001-05-01

    The spent nuclear fuel storage pool of Rokkasho reprocessing plant adopts the burnup credit' conception. Spent fuel assemblies are measured every one by one, by burnup monitors, and stored to a storage rack which is designed with specified residual enrichment. For nuclear criticality control, it is necessary for the burnup monitor that the measured value includes a kind of margin, which consists of errors of the monitor. In this paper, we describe the error of the burnup monitors, and the way of taking of the margin. From the result of calibration of the burnup monitor carried out from July through November, 1999, we describe that the way of taking of the margin is validated. And comments about possibility of error reduction are remarked. (author)

  12. Plan for characterization of K Basin spent nuclear fuel and sludge

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, L.A.; Marschman, S.C.

    1995-06-01

    This plan outlines a characterization program that supports the accelerated Path Forward scope and schedules for the Spent Nuclear Fuel stored in the Hanford K Basins. This plan is driven by the schedule to begin fuel transfer by December 1997. The program is structured for 4 years and is limited to in-situ and laboratory examinations of the spent nuclear fuel and sludge in the K East and K West Basins. The program provides bounding behavior of the fuel, and verification and acceptability for three different sludge disposal pathways. Fuel examinations are based on two shipping campaigns for the K West Basin and one from the K East Basin. Laboratory examinations include physical condition, hydride and oxide content, conditioning testing, and dry storage behavior.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ludewigt, Bernhard A; Quiter, Brian J.; Ambers, Scott D.

    2011-01-14

    The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity of the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size

  14. STRUCTURAL CALCULATIONS FOR THE CODISPOSAL OF TRIGA SPENT NUCLEAR FUEL IN A WASTE PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    S. Mastilovic

    1999-07-28

    The purpose of this analysis is to determine the structural response of a TRIGA Department of Energy (DOE) spent nuclear fuel (SNF) codisposal canister placed in a 5-Defense High Level Waste (DHLW) waste package (WP) and subjected to a tipover design basis event (DBE) dynamic load; the results will be reported in terms of displacements and stress magnitudes. This activity is associated with the WP design.

  15. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    PICKETT, W.W.

    2000-09-22

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. Because this sub-project is still in the construction/start-up phase, all verification activities have not yet been performed (e.g., canister cover cap and welding fixture system verification, MCO Internal Gas Sampling equipment verification, and As-built verification.). The verification activities identified in this report that still are to be performed will be added to the start-up punchlist and tracked to closure.

  16. Cooling Performance Evaluation of the Hybrid Heat Pipe for Spent Nuclear Fuel Dry Storage Cask

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yeong Shin; Bang, In Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    To evaluate the concept of the cooling device, 2-step CFD analysis was conducted for the cooling performance of hybrid heat pipe, which consists of single fuel assembly model and full scope dry cask model. As a passive cooling device of the metal cask for dry storage of spent nuclear fuel, hybrid heat pipe was applied to DPC developed in Korea. Hybrid heat pipe is the heat pipe containing neutron absorber can be used as a passive cooling in nuclear application with both decay heat removal and control the reactivity. In this study, 2-step CFD analysis was performed to find to evaluate the heat pipe-based passive cooling system for the application to the dry cask. Only spent fuel pool cannot satisfy the demands for high burnup fuel and large amount of spent fuel. Therefore, it is necessary to prepare supplement of the storage facilities. As one of the candidate of another type of storage, dry storage method have been preferred due to its good expansibility of storage capacity and easy long-term management. Dry storage uses the gas or air as coolant with passive cooling and neutron shielding materials was used instead of water in wet storage system. It is relatively safe and emits little radioactive waste for the storage. As short term actions for the limited storage capacity of spent fuel pool, it is considered to use dry interim/long term storage method to increase the capacity of spent nuclear fuel storage facilities. For 10-year cooled down spent fuel in the pool storage, fuel rod temperature inside metal cask is expected over 250 .deg. C in simulation. Although it satisfied the criteria that cladding temperature of the spent fuel should keep under 400 .deg. C during storage period, high temperature inside cask can accelerate the thermal degradation of the structural materials consisting metal cask and fuel assembly as well as limitation of the storage capacity of metal cask. In this paper, heat pipe-based cooling device for the dry storage cask was suggested for

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  18. Treatment and recycling of spent nuclear fuel. Actinide partitioning - Application to waste management

    Energy Technology Data Exchange (ETDEWEB)

    Abonneau, E.; Baron, P.; Berthon, C.; Berthon, L.; Beziat, A.; Bisel, I.; Bonin, L.; Bosse, E.; Boullis, B.; Broudic, J.C.; Charbonnel, M.C.; Chauvin, N.; Den Auwer, C.; Dinh, B.; Duhamet, J.; Escleine, J.M.; Grandjean, S.; Guilbaud, P.; Guillaneux, D.; Guillaumont, D.; Hill, C.; Lacquement, J.; Masson, M.; Miguirditchian, M.; Moisy, P.; Pelletier, M.; Ravenet, A.; Rostaing, C.; Royet, V.; Ruas, A.; Simoni, E.; Sorel, C.; Vaudano, A.; Venault, L.; Warin, D.; Zaetta, A.; Pradel, P.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Forestier, A.; Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Latge, C.; Limoge, Y.; Madic, C.; Santarini, G.; Seiler, J.M.; Sollogoob, P.; Vernaz, E.; Bazile, F.; Parisot, J.P.; Finot, P.; Roberts, J.F

    2008-07-01

    subsequent to its in-reactor dwell time, spent fuel still contains large amounts of materials that are recoverable, for value-added energy purposes (uranium, plutonium), together with fission products, and minor actinides, making up the residues from nuclear reactions. The treatment and recycling of spent nuclear fuel, as implemented in France, entail that such materials be chemically partitioned. The development of the process involved, and its deployment on an industrial scale stand as a high achievement of French science, and technology. Treatment and recycling allow both a satisfactory management of nuclear waste to be implemented, and substantial savings, in terms of fissile material. Bolstered of late as it has been, due to spectacularly skyrocketing uranium prices, this strategy is bound to become indispensable, with the advent of the next generation of fast reactors. This Monograph surveys the chemical process used for spent fuel treatment, and its variants, both current, and future. It outlines currently ongoing investigations, setting out the challenges involved, and recent results obtained by CEA. (authors)

  19. Rod internal pressure of spent nuclear fuel and its effects on cladding degradation during dry storage

    Science.gov (United States)

    Kim, Ju-Seong; Hong, Jong-Dae; Yang, Yong-Sik; Kook, Dong-Hak

    2017-08-01

    Temperature and hoop stress limits have been used to prevent the gross rupture of spent nuclear fuel during dry storage. The stress due to rod internal pressure can induce cladding degradation such as creep, hydride reorientation, and delayed hydride cracking. Creep is a self-limiting phenomenon in a dry storage system; in contrast, hydride reorientation and delayed hydride cracking are potential degradation mechanisms activated at low temperatures when the cladding material is brittle. In this work, a conservative rod internal pressure and corresponding hoop stress were calculated using FRAPCON-4.0 fuel performance code. Based on the hoop stresses during storage, a study on the onset of hydride reorientation and delayed hydride cracking in spent nuclear fuel was conducted under the current storage guidelines. Hydride reorientation is hard to occur in most of the low burn-up fuel while some high burn-up fuel can experience hydride reorientation, but their effect may not be significant. On the other hand, delayed hydride cracking will not occur in spent nuclear fuel from pressurized water reactor; however, there is a lack of confirmatory data on threshold intensity factor for delayed hydride cracking and crack size distribution in the fuel.

  20. Projects of Modifications of design for mitigation of accidents outside the design Bases on nuclear Central PWR Siemens-KWU and Westinghouse; Proyectos de Modificaciones de Sieno para Mitigacion de Accidentes fuera de la Bases de Diseno en Centrales Nucleares PWR Siemens-KWU y Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez Gonzalez, G.; Cano Rodriguez, L. A.; Arguello Tara, A.

    2014-07-01

    Following the accident at the Japanese Fukushima-Daiichi NPP, the different regulators of nuclear power generation have required numerous reports regarding the evaluation and modification of the capacity of the plants to face accidents with severities beyond that established in their Design Bases. Under this new scenario, with multiple new demands and commitments, EA has carried out the required works for the implementation of strategies to mitigate the consequences of beyond Design Basis accidents for utilities owning Siemens-KWU and Westinghouse PWR nuclear power plants. (Author)

  1. Validation of gadolinium burnout using PWR benchmark specification

    Energy Technology Data Exchange (ETDEWEB)

    Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2014-07-01

    Graphical abstract: - Highlights: • We present methodology for validation of gadolinium burnout in PWR. • We model 17 × 17 PWR fuel assembly using MCB code. • We demonstrate C/E ratios of measured and calculated concentrations of Gd isotopes. • The C/E for Gd154, Gd156, Gd157, Gd158 and Gd160 shows good agreement of ±10%. • The C/E for Gd152 and Gd155 shows poor agreement below ±10%. - Abstract: The paper presents comparative analysis of measured and calculated concentrations of gadolinium isotopes in spent nuclear fuel from the Japanese Ohi-2 PWR. The irradiation of the 17 × 17 fuel assembly containing pure uranium and gadolinia bearing fuel pins was numerically reconstructed using the Monte Carlo Continuous Energy Burnup Code – MCB. The reference concentrations of gadolinium isotopes were measured in early 1990s at Japan Atomic Energy Research Institute. It seems that the measured concentrations were never used for validation of gadolinium burnout. In our study we fill this gap and assess quality of both: applied numerical methodology and experimental data. Additionally we show time evolutions of infinite neutron multiplication factor K{sub inf}, FIMA burnup, U235 and Gd155–Gd158. Gadolinium-based materials are commonly used in thermal reactors as burnable absorbers due to large neutron absorption cross-section of Gd155 and Gd157.

  2. Choice of method - evaluation of strategies and systems for disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    2010-10-15

    This report deals with the question of how the Swedish spent nuclear fuel is to be disposed of. What are the requirements? What are the alternatives? In the main chapter of the report, an evaluation is made of the KBS-3 method compared with other strategies and systems for final disposal of spent nuclear fuel. An appendix to the report presents in general terms how the KBS-3 method has developed from the end of the 1970s up to today. The report is one of a number of supporting documents for SKB's applications for construction and operation of the final repository for spent nuclear fuel. In parallel with and as a basis for the present report, SKB has prepared the reports Principer, strategier och system foer slutligt omhaendertagande av anvaent kaernbraensle ('Principles, strategies and systems for final disposal of spent nuclear fuel') /Grundfelt 2010a/, Jaemfoerelse mellan KBS-3-metoden och deponering i djupa borrhaal foer slutlig foervaring av anvaent kaernbraensle ('Comparison between the KBS-3 method and deposition in deep boreholes for final disposal of spent nuclear fuel') /Grundfelt 2010b/ and Utvecklingen av KBS-3- metoden. Genomgaang av forskningsprogram, saekerhetsanalyser, myndighetsgranskningar samt SKB:s internationella forskningssamarbete ('Development of the KBS-3 method. Review of research programmes, safety assessments, regulatory reviews and SKB's international research cooperation') /SKB 2010a/. The reports are in Swedish, but contain summaries in English. The first report is an update of the comprehensive account of alternative methods presented by SKB in 2000. The second report presents a comparison between the KBS-3 method and the Deep Boreholes concept, plus a status report on research and development in the area of Deep Boreholes. The last report describes how the KBS-3 method has been developed from the end of the 1970s up to today. It further describes how the method has been further developed and

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Study on the fire-protection-system for interim storage facilities of spent nuclear fuel and transportation ships

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. O; Choi, M. H.; Lee, S. C. and others [Dongbang Electron Industry Corporation, Seoul (Korea, Republic of)

    1993-12-15

    This study consists of : the fire risk and it's fire protection for the storage facilities and transportation equipments of dangerous goods, the fire risk and it's fire protection for the interim storage facilities of spent nuclear fuel, the fire risk and it's fire protection for the dangerous goods transportation ships, the necessary equipment for safety of ships and regulations of fire fighting equipment for ships, technical specification of spent nuclear fuel transportation ships which are operated in foreign countries, draft of fire protection guideline for interim storage facilities of spent nuclear fuel, inspection items of fire fighting equipment, scope of education and training. On the basis of the aforementioned, a draft of fire protection guideline for interim storage facilities of spent nuclear fuel is proposed and the regulations for ship engaged in the a carrage of dangerous goods that should be considered in design and operation stage are proposed.

  6. H 2 inhibition of radiation induced dissolution of spent nuclear fuel

    Science.gov (United States)

    Trummer, Martin; Roth, Olivia; Jonsson, Mats

    2009-01-01

    In order to elucidate the effect of noble metal clusters in spent nuclear fuel on the kinetics of radiation induced spent fuel dissolution we have used Pd particle doped UO 2 pellets. The catalytic effect of Pd particles on the kinetics of radiation induced dissolution of UO 2 during γ-irradiation in HCO3- containing solutions purged with N 2 and H 2 was studied in this work. Four pellets with Pd concentrations of 0%, 0.1%, 1% and 3% were produced to mimic spent nuclear fuel. The pellets were placed in 10 mM HCO3- aqueous solutions and γ-irradiated, and the dissolution of UO22+ was measured spectrophotometrically as a function of time. Under N 2 atmosphere, 3% Pd prevent the dissolution of uranium by reduction with the radiolytically produced H 2, while the other pellets show a rate of dissolution of around 1.6 × 10 -9 mol m -2 s -1. Under H 2 atmosphere already 0.1% Pd effectively prevents the dissolution of uranium, while the rate of dissolution for the pellet without Pd is 1.4 × 10 -9 mol m -2 s -1. It is also shown in experiments without radiation in aqueous solutions containing H 2O 2 and O 2 that ɛ-particles catalyze the oxidation of the UO 2 matrix by these molecular oxidants, and that the kinetics of the catalyzed reactions is close to diffusion controlled.

  7. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository - Volume 3: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, L.L.; Wilson, J.R. (INEEL); Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K. (SNL); Rath, J.S. (New Mexico Engineering Research Institute)

    1998-10-01

    The United States Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  9. Safety case for the disposal of spent nuclear fuel at Olkiluoto - Synthesis 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a spent nuclear fuel repository. Consistent with the Government Decisions-in- Principle, this foresees a repository developed in bedrock at the Olkiluoto site according to the KBS-3 method, designed to accept spent nuclear fuel from the lifetime operations of the Olkiluoto and Loviisa reactors. Synthesis 2012 presents a synthesis of Posiva Oy's Safety Case 'TURVA-2012' portfolio. It summarises the design basis for the repository at the Olkiluoto site, the assessment methodology and key results of performance and safety assessments. It brings together all the lines of argument for safety, evaluation of compliance with the regulatory requirements, and statement of confidence in long-term safety and Posiva's safety analyses. The TURVA-2012 safety case demonstrates that the proposed repository design provides a safe solution for the disposal of spent nuclear fuel, and that the performance and safety assessments are fully consistent with all the legal and regulatory requirements related to long-term safety as set out in Government Decree 736/2008 and in guidance from the nuclear regulator - the STUK. Moreover, Posiva considers that the level of confidence in the demonstration of safety is appropriate and sufficient to submit the construction licence application to the authorities. The assessment of long-term safety includes uncertainties, but these do not affect the basic conclusions on the long-term safety of the repository. (orig.)

  10. Analysis of Spent Nuclear Fuel Imaging Using Multiple Coulomb Scattering of Cosmic Muons

    Science.gov (United States)

    Chatzidakis, Stylianos; Choi, Chan K.; Tsoukalas, Lefteri H.

    2016-12-01

    Cosmic ray muons passing through matter lose energy from inelastic collisions with electrons and are deflected from nuclei due to multiple Coulomb scattering. The strong dependence of scattering on atomic number Z and the recent developments on position sensitive muon detectors indicate that multiple Coulomb scattering could be an excellent candidate for spent nuclear fuel imaging. Muons present significant advantages over existing monitoring and imaging techniques and can play a central role in monitoring nuclear waste and spent nuclear fuel stored in dense well shielded containers. The main purpose of this paper is to investigate the applicability of multiple Coulomb scattering for imaging of spent nuclear fuel dry casks stored within vertical and horizontal commercial storage dry casks. Calculations of muon scattering were performed for various scenarios, including vertical and horizontal fully loaded dry casks, half loaded dry casks, dry casks with one row of fuel assemblies missing, dry casks with one fuel assembly missing and empty dry casks. Various detector sizes (1.2 m ×1.2 m, 2.4 m ×2.4 m and 3.6 m ×3.6 m) and number of muons (105, 5 · 105, 106 and 107) were used to assess the effect on image resolution. The Point-of-Closest-Approach (PoCA) algorithm was used for the reconstruction of the stored contents. The results demonstrate that multiple Coulomb scattering can be used to successfully reconstruct the dry cask contents and allow identification of all scenarios with the exception of one fuel assembly missing. In this case, an indication exists that a fuel assembly is not present; however, the resolution of the imaging algorithm was not enough to identify exact location.

  11. Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages, Phase 1. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, C.A.

    1978-08-01

    The objective of this project is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements toward reducing its overall duration. Thus, the plant's availability for power production would be increased. Revisions in procedures and some equipment modifications were implemented and evaluated during the Zion Unit 2 refueling/maintenance outage beginning in January 1977. Analysis of the observed data has identified benefits available through improved refueling equipment and also areas where additional new, innovative refueling, or refueling-related equipment should be beneficial. A number of specific design concepts are recommended as a result of Phase 1. In addition, a new master planning mechanism is described for implementation during subsequent planned outages at Zion Station. This final report describes the recommended conceptual designs and planning mechanism and assesses their impact upon future outages. Their effect on savings in refueling time, labor, and radiation exposure is discussed. The estimated economic payoff for these concepts was found to be of such significance that an additional phase of the program is warranted. During this extended phase, a more detailed engineering study should be undertaken to determine the cost of implementation along with more specific estimates of the benefits for PWR plants already in operation or under construction.

  12. External cost assessment for nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Byung Heung [Korea National University of Transportation, Chungju (Korea, Republic of); Ko, Won Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    Nuclear power is currently the second largest power supply method in Korea and the number of nuclear power plants are planned to be increased as well. However, clear management policy for spent fuels generated from nuclear power plants has not yet been established. The back-end fuel cycle, associated with nuclear material flow after nuclear reactors is a collection of technologies designed for the spent fuel management and the spent fuel management policy is closely related with the selection of a nuclear fuel cycle. Cost is an important consideration in selection of a nuclear fuel cycle and should be determined by adding external cost to private cost. Unlike the private cost, which is a direct cost, studies on the external cost are focused on nuclear reactors and not at the nuclear fuel cycle. In this research, external cost indicators applicable to nuclear fuel cycle were derived and quantified. OT (once through), DUPIC (Direct Use of PWR SF in CANDU), PWR-MOX (PWR PUREX reprocessing), and Pyro-SFR (SFR recycling with pyroprocessing) were selected as nuclear fuel cycles which could be considered for estimating external cost in Korea. Energy supply security cost, accident risk cost, and acceptance cost were defined as external cost according to precedent and estimated after analyzing approaches which have been adopted for estimating external costs on nuclear power generation.

  13. SPENT NUCLEAR FUEL STORAGE BASIN WATER CHEMISTRY: ELECTROCHEMICAL EVALUATION OF ALUMINUM CORROSION

    Energy Technology Data Exchange (ETDEWEB)

    Hathcock, D

    2007-10-30

    The factors affecting the optimal water chemistry of the Savannah River Site spent fuel storage basin must be determines in order to optimize facility efficiency, minimize fuel corrosion, and reduce overall environmental impact from long term spent nuclear fuel storage at the Savannah River Site. The Savannah River National Laboratory is using statistically designed experiments to study the effects of NO{sub 3}{sup -}, SO{sub 4}{sup 2-}, and Cl{sup -} concentrations on alloys commonly used not only as fuel cladding, but also as rack construction materials The results of cyclic polarization pitting and corrosion experiments on samples of Al 6061 and 1100 alloys will be used to construct a predictive model of the basin corrosion and its dependence on the species in the basin. The basin chemistry model and corrosion will be discussed in terms of optimized water chemistry envelope and minimization of cladding corrosion.

  14. Searching for Methods on Evaluation Alternatives and Studying Decision Making System Regarding Enhancing Publicity of Nuclear Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seongkyung [Myongji Univ., Seoul (Korea, Republic of); Choi, Seungho; Kim, Hyerim; Song, Jiyeon [DOMO Communication Consulting, Seoul (Korea, Republic of); Lee, Yoonsup; Sohn, Seohyun [Fleishman-Hillard Korea, Seoul (Korea, Republic of)

    2013-05-15

    This study was done in order to anticipate the aspect of publicity enhancement on nuclear spent fuel so that it can find the evaluation methods of alternative ways of management which could applied actually and make the decision making system of Publicity Enhancement Committee in advance. In Korea, the nuclear spent fuel is temporarily stored inside of the nuclear facility field, and it is expected that Gori nuclear facility is going to be saturated since 2016 but the solutions are still incomplete. The problem of management of nuclear spent fuel is an important issue in terms of not only the nuclear power policy but also of safe management of the already made nuclear spent fuel. This study has its meaning to draw the evaluation criteria of the management alternatives on nuclear spent fuel which can be applied in Korean case, and to find the necessity of verifying the evaluation of management alternatives through Publicity Enhancement because of different stands according to the interests. As a result, rather than technological engineering safety evaluation, qualitative analysis in terms of social costs, quantitative evaluation in terms of economic costs, this study advises the methods of public hearings and citizen juries which are effective, which makes it meaningful.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

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

  16. Modeling of molecular and particulate transport in dry spent nuclear fuel canisters

    Science.gov (United States)

    Casella, Andrew M.

    2007-09-01

    The transportation and storage of spent nuclear fuel is one of the prominent issues facing the commercial nuclear industry today, as there is still no general consensus regarding the near- and long-term strategy for managing the back-end of the nuclear fuel cycle. The debate continues over whether the fuel cycle should remain open, in which case spent fuel will be stored at on-site reactor facilities, interim facilities, or a geologic repository; or if the fuel cycle should be closed, in which case spent fuel will be recycled. Currently, commercial spent nuclear fuel is stored at on-site reactor facilities either in pools or in dry storage containers. Increasingly, spent fuel is being moved to dry storage containers due to decreased costs relative to pools. As the number of dry spent fuel containers increases and the roles they play in the nuclear fuel cycle increase, more regulations will be enacted to ensure that they function properly. Accordingly, they will have to be carefully analyzed for normal conditions, as well as any off-normal conditions of concern. This thesis addresses the phenomena associated with one such concern; the formation of a microscopic through-wall breach in a dry storage container. Particular emphasis is placed on the depressurization of the canister, release of radioactivity, and plugging of the breach due to deposition of suspended particulates. The depressurization of a dry storage container upon the formation of a breach depends on the temperature and quantity of the fill gas, the pressure differential across the breach, and the size of the breach. The first model constructed in this thesis is capable of determining the depressurization time for a breached container as long as the associated parameters just identified allow for laminar flow through the breach. The parameters can be manipulated to quantitatively determine their effect on depressurization. This model is expanded to account for the presence of suspended particles. If

  17. Cost estimations for deep disposal of spent nuclear fuels; Kostnadsberaekning av djupfoervaring av det anvanda kaernbraenslet

    Energy Technology Data Exchange (ETDEWEB)

    Palmqvist, K.; Wallroth, T. [BERGAB - Berggeologiska Undersoekningar AB, Goeteborg (Sweden); Green, L.; Joensson, Lars [Peab Berg AB, Goeteborg (Sweden)

    1999-10-01

    According to the Act on the Financing of Future Expenses for Spent Nuclear Fuel etc. (Financing Act), the Swedish Nuclear Fuel and Waste Management Co. (SKB) must submit, every year, to the Swedish Nuclear Power Inspectorate (SKI), a cost estimate for the management of spent nuclear fuel and for the decommissioning and dismantling of the nuclear power plants. After SKI has examined and evaluated the cost estimates, SKI must submit a proposal to the Government concerning the fee which should be paid by the nuclear power companies per kWh of generated electricity. According to the Financing Act, the reactor owners must pledge collateral in the event that the accumulated fees should be found to be insufficient as a result of early closure of reactors or as a result of underestimating the future expenses of managing the spent nuclear fuel and of decommissioning and dismantling the reactors. The future total expenses resulting from the Financing Act are estimated at about SEK 48 billion at the January 1998 price level. Of this amount, the cost of the final disposal of spent nuclear fuel in SKB's programme is expected to amount to about SEK 12 billion. SKB's estimate comprises the cost of siting, construction and operation of a deep repository for spent nuclear fuel, based on the KBS-3 concept, and a rock cavern for other long-lived waste which SKB plans to locate next to the spent fuel repository. The cost estimate also includes the dismantling and closure of the facility once all of the fuel and the long-lived waste are deposited. The calculations are based on all of the fuel, which will be generated through the operation of the 12 Swedish reactors during a period of 25 years and for every additional year of operation. At the beginning of 1998, SKI commissioned BERGAB to evaluate the cost estimate for the deep disposal of the spent nuclear fuel. The task was divided into two stages, namely a study which was submitted in June 1998 concerning the technical

  18. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Complementary considerations 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Complementary Considerations sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective of enhancing confidence in the outcomes of the safety assessment for a spent nuclear fuel repository to be constructed at Olkiluoto, Finland. The main emphasis in this report is on the evidence and understanding that can be gained from observations at the site, including its regional geological environment, and from natural and anthropogenic analogues for the repository, its components and the processes that affect safety. In particular, the report addresses diverse and less quantifiable types of evidence and arguments that are enclosed to enhance confidence in the outcome of the safety assessment. These complementary considerations have been described as evaluations, evidence and qualitative supporting arguments that lie outside the scope of the other reports of the quantitative safety assessment. The experience with natural analogues for the long-term durability of the materials involved and the extent of processes provides high confidence in our understanding of the disposal system and its evolution. For each engineered barrier and key process, there is increasing analogue evidence to support the conceptual models and parameters. Regarding the suitability of the Olkiluoto site to host a spent fuel repository, a number of factors have been identified that indicate the suitability of crystalline host rock in general, and that of the Olkiluoto site in particular. The report also provides radiation background information for the use of complementary indicators, which aid in putting the results of the safety analysis presented in Assessment of Radionuclide Release Scenarios for the Repository System and Biosphere Assessment in a broader perspective to show that the radiation originating from a spent nuclear fuel repository remains in most cases much below natural background radiation or that caused by non-nuclear industries. (orig.)

  19. Technical data summary supporting the spent nuclear fuel environment impact statement, March 1994

    Energy Technology Data Exchange (ETDEWEB)

    Geddes, R.L.; Claxton, R.E.; Lengel, J.D. [and others

    1994-03-01

    This report has been compiled by the WSRC Nuclear Materials Processing Division`s Planning Section at the request of the Office of Spent Fuel Management and Special Projects (EM-37) to support issuance of the Spent Nuclear Fuel Environmental Impact Statement. Savannah River Site input data evaluates five programmatic options (including {open_quotes}No Action{close_quotes}) ranging up to transfer of all DOE responsibility spent fuel to the SRS. For each option, a range of management/disposition scenarios has been examined. Each case summary provides information relative to the technical proposal, technical issues, environmental impacts, and projected costs for a forty year period (FY-35) when it is assumed that the material will be dispositioned from the SRS. The original issue of the report which was prepared under severe time constraints contained many simplifications and assumptions. Although the revisions have corrected some of the shortcomings of the original report, it is still highly recommended that significant additional study be performed before basing key decisions upon the data contained in this report. The data represents the best effort by a significant group of technical personnel familiar with nuclear materials processing, handling, and storage; but it is likely that careful scrutiny will reveal numerous discrepancies, inconsistencies and omissions. Nor does this report attempt to analyze every potential disposal pathway, but probably establishes the bounds for the most of the viable pathways. The bulk of the effort went into defining the engineering approaches necessary to execute the various mission scenarios which were changed since the last revision. The decision to limit reprocessing to only SRS aluminum clad required a major alteration of the TDS. Collection and/or calculation of much of the various waste, emission, and utility consumption data, so important to an EIS, has been updated since the last revision, but not thoroughly completed.

  20. NDE of copper canisters for long-term storage of spent nuclear fuel from the Swedish nuclear power plants

    Science.gov (United States)

    Stepinski, Tadeusz

    2003-07-01

    Sweden has been intensively developing methods for long term storage of spent fuel from the nuclear power plants for twenty-five years. A dedicated research program has been initiated and conducted by the Swedish company SKB (Swedish Nuclear Fuels and Waste Management Co.). After the interim storage SKB plans to encapsulate spent nuclear fuel in copper canisters that will be placed at a deep repository located in bedrock. The canisters filled with fuel rods will be sealed by an electron beam weld. This paper presents three complementary NDE techniques used for assessing the sealing weld in copper canisters, radiography, ultrasound, and eddy current. A powerful X-ray source and a digital detector are used for the radiography. An ultrasonic array system consisting of a phased ultrasonic array and a multi-channel electronics is used for the ultrasonic examination. The array system enables electronic focusing and rapid electronic scanning eliminating the use of a complicated mechanical scanner. A specially designed eddy current probe capable of detecting small voids at the depth up to 4 mm in copper is used for the eddy current inspection. Presently, all the NDE techniques are verified in SKB's Canister Laboratory where full scale canisters are welded and examined.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

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

  2. Technical Basis Spent Nuclear Fuel (SNF) Project Radiation and Contamination Trending Program

    Energy Technology Data Exchange (ETDEWEB)

    ELGIN, J.C.

    2000-10-02

    This report documents the technical basis for the Spent Nuclear Fuel (SNF) Program radiation and contamination trending program. The program consists of standardized radiation and contamination surveys of the KE Basin, radiation surveys of the KW basin, radiation surveys of the Cold Vacuum Drying Facility (CVD), and radiation surveys of the Canister Storage Building (CSB) with the associated tracking. This report also discusses the remainder of radiological areas within the SNFP that do not have standardized trending programs and the basis for not having this program in those areas.

  3. Design Verification Report Spent Nuclear Fuel (SNF) Project Canister Storage Building (CSB)

    Energy Technology Data Exchange (ETDEWEB)

    BAZINET, G.D.

    2000-11-03

    The Sub-project W379, ''Spent Nuclear Fuel Canister Storage Building (CSB),'' was established as part of the Spent Nuclear Fuel (SNF) Project. The primary mission of the CSB is to safely store spent nuclear fuel removed from the K Basins in dry storage until such time that it can be transferred to the national geological repository at Yucca Mountain Nevada. This sub-project was initiated in late 1994 by a series of studies and conceptual designs. These studies determined that the partially constructed storage building, originally built as part of the Hanford Waste Vitrification Plant (HWVP) Project, could be redesigned to safely store the spent nuclear fuel. The scope of the CSB facility initially included a receiving station, a hot conditioning system, a storage vault, and a Multi-Canister Overpack (MCO) Handling Machine (MHM). Because of evolution of the project technical strategy, the hot conditioning system was deleted from the scope and MCO welding and sampling stations were added in its place. This report outlines the methods, procedures, and outputs developed by Project W379 to verify that the provided Structures, Systems, and Components (SSCs): satisfy the design requirements and acceptance criteria; perform their intended function; ensure that failure modes and hazards have been addressed in the design; and ensure that the SSCs as installed will not adversely impact other SSCs. The original version of this document was prepared by Vista Engineering for the SNF Project. The purpose of this revision is to document completion of verification actions that were pending at the time the initial report was prepared. Verification activities for the installed and operational SSCs have been completed. Verification of future additions to the CSB related to the canister cover cap and welding fixture system and MCO Internal Gas Sampling equipment will be completed as appropriate for those components. The open items related to verification of those

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

    Science.gov (United States)

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

    1982-01-19

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

  5. Integrated data base report--1995: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    The information in this report summarizes the U.S. Department of Energy (DOE) data base for inventories, projections, and characteristics of domestic spent nuclear fuel and radioactive waste. This report is updated annually to keep abreast of continual waste inventory and projection changes in both the government and commercial sectors. Baseline information is provided for DOE program planning purposes and to support DOE program decisions. Although the primary purpose of this document is to provide background information for program planning within the DOE community, it has also been found useful by state and local governments, the academic community, and some private citizens.

  6. A structural analysis on the KN-12 spent nuclear fuel transport casks

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Lee, Jae Hyung; Na, Jae Yun [Chungnam National Univ., Taejon (Korea, Republic of)

    2001-08-15

    In this study, safety of the spent nuclear fuel cask KN-12 which is developed in 2000 is evaluated for hypothetical accidents conditions such as free drop, puncture, fire accident and water immersion. Finite element code ABAQUS/Explicit is used to compare with safety analysis report of the GNB in which analysis is performed with LS-DYNA3D for hypothetical accident conditions. Through this study, the safety of KN-12 is evaluated by comprehensive structural analysis. The capability and technological advancement of Korean community on the analysis and structural assessment of the cask will be improved. Also people's anxiety about radioactive dangers will be eliminated.

  7. Potential corrosion and degradation mechanisms of Zircaloy cladding on spent nuclear fuel in a tuff repository

    Energy Technology Data Exchange (ETDEWEB)

    Rothman, A.J.

    1984-09-01

    A literature review and analysis were made of corrosion and degradation processes applicable to Zircaloy cladding on spent nuclear fuel in a tuff repository. In particular, lifetime sought for the Zircaloy is 10,000 years. Among the potential failure mechanisms examined were: oxidation by steam, air, and water, including the effects of ions whose presence is anticipated in the water; mechanical overload; stress (creep) rupture; stress-corrosion cracking (SCC); and delayed failure due to hydride cracking. The conclusion is that failure due to oxidation is not credible, although a few experiments are suggested to confirm the effect of aqueous fluoride on the Zircaloy cladding. Mechanical overload is not a problem, and failure from stress-rupture does not appear likely based on a modified Larson-Miller analysis. Analysis shows that delayed hydride cracking is not anticipated for the bulk of spent fuel pins. However, for a minority of pins under high stress, there is some uncertainty in the analysis as a result of: (1) uncertainty about crack depths in spent fuel claddings and (2) the effect of slow cooling on the formation of radially oriented hydride precipitates. Experimental resolution is called for. Finally, insufficient information is currently available on stress-corrosion cracking. While evidence is presented that SCC failure is not likely to occur, it is difficult to demonstrate this conclusively because the process is not clearly understood and data are limited. Further experimental work on SCC susceptibility is especially needed.

  8. Precisely determined the spent nuclear fuel antineutrino flux and spectrum for Daya Bay antineutrino experiment

    CERN Document Server

    Ma, X B; Chen, Y X; Zhong, W L; An, F P

    2015-01-01

    Spent nuclear fuel (SNF) antineutrino flux is an important source of uncertainties for a reactor neutrino flux prediction. However, if one want to determine the contribution of spent fuel, many data are needed, such as the amount of spent fuel in the pool, the time after discharged from the reactor core, the burnup of each assembly, and the antineutrino spectrum of the isotopes in the spend fuel. A method to calculate the contribution of SNF is proposed in this study. In this method, reactor simulation code verified by experiment have been used to simulate the fuel depletion by taking into account more than 2000 isotopes and fission products, the quantity of SNF in each six spend fuel pool, and the antineutrino spectrum of SNF varying with time after SNF discharged from core. Results show that the contribution of SNF to the total antineutrino flux is about 0.26%~0.34%, and the shutdown impact is about 20%. The SNF spectrum would distort the softer part of antineutrino spectra, and the maximum contribution fro...

  9. Experimental observations on electrorefining spent nuclear fuel in molten LiCl-KCl/liquid cadmium system.

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, T. A.; Laug, D. V.; Li, S. X.; Sofu, T.

    1999-07-14

    Argonne National Laboratory (ANL) is currently performing a demonstration program for the Department of Energy (DOE) which processes spent nuclear fuel from the Experimental Breeder Reactor (EBR-II). One of the key steps in this demonstration program is electrorefining of the spent fuel in a molten LiCl-KCl/liquid cadmium system using a pilot scale electrorefiner (Mk-IV ER). This article summarizes experimental observations and engineering aspects for electrorefining spent fuel in the molten LiCl-KCl/liquid cadmium system. It was found that the liquid cadmium pool acted as an intermediate electrode during the electrorefining process in the ER. The cadmium level was gradually decreased due to its high vapor pressure and vaporization rate at the ER operational temperature. The low cadmium level caused the anode assembly momentarily to touch the ER vessel hardware, which generated a periodic current change at the salt/cathode interface and improved uranium recovery efficiency for the process. The primary current distributions calculated by numerical simulations were used in interpreting the experimental results.

  10. A study of antineutrino spectra from spent nuclear fuel at Daya Bay

    Institute of Scientific and Technical Information of China (English)

    ZHOU Bin; RUAN Xi-Chao; NIE Yang-Bo; ZHOU Zu-Ying; AN Feng-Peng; CAO Jun

    2012-01-01

    The Daya Bay Reactor Antineutrino Experiment is designed to determine the as yet unknown neutrino mixing angle,θ13,by measuring the disappearance of electron antineutrinos from several nuclear reactor cores.The projected sensitivity in sin2(2θ13) of better than 0.01 at a 90% CL should be achieved after three years of data-taking.Antineutrinos emitted from spent nuclear fuel (SNF) distort the soft part of the energy spectrum.In this article,a calculation of the antineutrino spectra from the long-life isotopes in SNF is performed.A non-equilibrium generation of long half-life isotopes during the running time of the reactor is also analyzed.Finally,we show that the antineutrino event rate contribution from SNF,which has been stored in the SNF pool for several years,may be non-negligible.

  11. Glutarimidedioxime. A complexing and reducing reagent for plutonium recovery from spent nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Xian, Liang [China Institute of Atomic Energy, Beijing (China). Radiochemistry Dept.; Tian, Guoxin [China Institute of Atomic Energy, Beijing (China). Radiochemistry Dept.; Lawrence Berkeley National Laboratory, Berkeley, CA (United States). Chemical Sciences Div.; Beavers, Christine M.; Teat, Simon J. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States). Advanced Light Source; Shuh, David K. [Lawrence Berkeley National Laboratory, Berkeley, CA (United States). Chemical Sciences Div.

    2016-04-04

    Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt-free complexing and reducing reagent, glutarimidedioxime (H{sub 2}A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1M HNO{sub 3} with H{sub 2}A. The complexation-reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu{sup IV} complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu{sup IV}, through the complexation-reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes.

  12. Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2016-12-01

    Radioactive iodine is the Achilles’ heel in the design for the safe geological disposal of spent UO2 nuclear fuel. Iodine’s high solubility and anticipated instant release during waste package compromise jeopardize performance assessment calculations. However, dissolution studies have indicated that the instant release fraction (IRF) of radioiodine (I) does not correlate with increasing fuel burn-up. In fact, there is a peak in the release iodine at around 50-60 Mwd/kgU and with increasing burn-up the instant release of iodine decreases. Detailed electron microscopy analysis of high burn-up fuel (~80 MWd/kgU) has revealed the presence of (Pd,Ag)(I,Br) nano-particles. As UO2 fuels are irradiated, the Ag and Pd content increases, from 239Pu fission, enabling radioiodine to be retained. The occurrence of these phases in nuclear fuels may have significant implications for the long-term behavior of iodine.

  13. Roles of Radiolytic and Externally Generated H2 in the Corrosion of Fractured Spent Nuclear Fuel.

    Science.gov (United States)

    Liu, Nazhen; Wu, Linda; Qin, Zack; Shoesmith, David W

    2016-11-15

    A 2-D model for the corrosion of spent nuclear fuel inside a failed nuclear waste container has been modified to determine the influence of various redox processes occurring within fractures in the fuel. The corrosion process is driven by reaction of the fuel with the dominant α radiolysis product, H2O2. A number of reactions are shown to moderate or suppress the corrosion rate, including H2O2 decomposition and a number of reactions involving dissolved H2 produced either by α radiolysis or by the corrosion of the steel container vessel. Both sources of H2 lead to the suppression of fuel corrosion, with their relative importance being determined by the radiation dose rate, the steel corrosion rate, and the dimensions of the fractures in the fuel. The combination of H2 from these two sources can effectively prevent corrosion when only micromolar quantities of H2 are present.

  14. 77 FR 4807 - Revised Fee Policy for Acceptance of Foreign Research Reactor Spent Nuclear Fuel From High-Income...

    Science.gov (United States)

    2012-01-31

    ... National Nuclear Security Administration Revised Fee Policy for Acceptance of Foreign Research Reactor... Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel'' (61 FR 25092, May..., Department of Energy. ACTION: Notice of a change in the fee policy. SUMMARY: This notice announces a...

  15. Site selection - location of the repository for spent nuclear fuel; Platsval - lokalisering av slutfoervaret foer anvaent kaernbraensle

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

    This document describes the localization work and SKB's choice of site for the repository. Furthermore, SKB's basis and rationale for the decisions taken during the work are reported. The document is Appendix PV of applications under the Nuclear Activities Act and the Environmental Code to both build and operate an encapsulation plant adjacent to the central interim storage facility for spent nuclear fuel in Oskarshamn, and to construct and operate a disposal facility for spent nuclear fuel at Forsmark in Oesthammar municipality

  16. Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Mueller, Don [ORNL; Goluoglu, Sedat [ORNL; Hollenbach, Daniel F [ORNL; Fox, Patricia B [ORNL

    2007-10-01

    The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform postclosure criticality calculations. The validation process applies the criticality analysis methodology approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report. The application systems for this validation consist of waste packages containing transport, aging, and disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types, initial enrichments, and burnup values that are expected from the waste stream and of varying degree of internal component degradation that may occur over the 10,000-year regulatory time period. The criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP) transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit (LBTL) functions for CSNF waste packages are determined based on the guidance in ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1. The development of this report is consistent with Test Plan for: Range of Applicability and Bias Determination for Postclosure Criticality. This calculation report has been developed in support of licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual TAD canister.

  17. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel G.; Morrow, Charles.

    2013-01-01

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  18. Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Yousry E-mail: gohar@anl.gov

    2001-11-01

    The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D-T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

  19. Development of melt dilute technology for disposition of aluminum based spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Swift, W.F. [Nuclear Material Management Division Westinghouse Savannah River Company, Savannah River Site Building 707-C, Aiken, SC 29808 (United States)

    2002-07-01

    The US Department of Energy (DOE) has for many years had a program for receipt and disposition of spent nuclear fuels of US origin from research reactors around the world. The research reactor spent nuclear fuel that consists of aluminum alloy composition has historically been returned to the Savannah River Site (SRS) and dispositioned via chemical reprocessing. In 1995, the DOE evaluated a number of alternatives to chemical reprocessing. In 2000, the DOE selected the melt-dilute alternative as the primary disposition path and direct disposal as the backup path. The melt-dilute technology has been developed from lab-scale demonstration up through the construction of a pilot-scale facility. The pilot-scale L-Area Experimental Facility (LEF) has been constructed and is ready for operation. The LEF will be used primarily, to confirm laboratory research on zeolite media for off- gas trapping and remote operability. Favorable results from the LEF are expected to lead to final design of the production melt-dilute facility identified as the Treatment and Storage Facility (TSF). This paper will describe the melt-dilute process and provide a status of the program development. (author)

  20. IMPACT ANALYSES AND TESTS OF CONCRETE OVERPACKS OF SPENT NUCLEAR FUEL STORAGE CASKS

    Directory of Open Access Journals (Sweden)

    SANGHOON LEE

    2014-02-01

    Full Text Available A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

  1. Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    D.R. Jackson; G.R. Kiebel

    1999-08-24

    Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Oxidative dissolution of spent nuclear fuel in aqueous alkaline solutions - An alternative to the Purex process?

    Energy Technology Data Exchange (ETDEWEB)

    Runde, Wolfgang; Peper, Shane; Brodnax, Lia; Crooks, William; Zehnder, Ralph; Jarvinen, Gordon

    2004-07-01

    As an alternative to acidic reprocessing of spent nuclear, oxidative dissolution of UO{sub 2} into aqueous alkaline solutions and subsequent separation of fission products is considered. The efficacy of such a method is limited by the kinetics of the UO{sub 2} dissolution and the capacity of alkaline solutions for dissolved U(VI) species. We performed a series of dissolution studies on UO{sub 2} and U{sub 3}O{sub 8} in aqueous alkaline solutions applying various oxidants. Among the oxidative agents commonly used to transform low-valence actinides into their higher oxidation states, H{sub 2}O{sub 2} has proven to be the most effective in basic media. Consequently, we investigated the dissolution of UO{sub 2} and U{sub 3}O{sub 8} in NaOH-H{sub 2}O{sub 2} and Na{sub 2}CO{sub 3}-H{sub 2}O{sub 2} solutions and determined the dissolution kinetics as a function of peroxide and hydroxide (carbonate) concentrations. Methods to remove fission products, e.g., Cs, Sr, Ba and Zr, from alkaline solutions will be evaluated based upon their decontamination factors. We will discuss the feasibility of using chemically oxidizing alkaline solutions as an alternative spent nuclear fuel reprocessing method based on results from experimental quantitative investigations. (authors)

  4. Unsteady heat conduction in the soil layers above underground repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talukder, N.K. [Clark Atlanta Univ., GA (United States)

    2000-04-01

    Due to radioactivity of spent nuclear fuel, a repository is expected to act as a heat source of exponentially decreasing intensity over hundreds of years. In case of underground emplacement of such a heat source, the temperature changes in the soil layers surrounding the heat source may have important implications such as evaporation of the water contained in the soil and its subsequent condensation. Assuming a uniformly distributed power generation over a horizontal, relatively thin, circular zone of several thousand meters diameter located several hundred meters below the ground surface, the temperature variations along the vertical centerline of the heating zone was estimated analytically under simplifying assumptions. Unsteady one-dimensional heat conduction in a semi-infinite solid with an exponentially decreasing heat flux at the proximal end was considered. The corresponding solution of the Fourier equation for heat conduction contains Error Functions of complex arguments. The evaluation of the Error Functions for discrete space and time parameters was performed applying analytical procedures and using standard tables. The results show temperature distributions in the soil at various time points over thousands of years after underground emplacement of spent nuclear fuel. (orig.)

  5. Chemical reactivity testing for the National Spent Nuclear Fuel Program. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Koester, L.W.

    2000-02-08

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, Y60-101PD, Quality Program Description, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted. The project consists of conducting three separate series of related experiments, ''Passivation of Uranium Hydride Powder With Oxygen and Water'', '''Passivation of Uranium Hydride Powder with Surface Characterization'', and ''Electrochemical Measure of Uranium Hydride Corrosion Rate''.

  6. Sampling and analysis plan for the preoperational environmental survey of the spent nuclear fuel project facilities

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL, R.M.

    1999-04-01

    This sampling and analysis plan will support the preoperational environmental monitoring for construction, development, and operation of the Spent Nuclear Fuel (SNF) Project facilities, which have been designed for the conditioning and storage of spent nuclear fuels; particularly the fuel elements associated with the operation of N-Reactor. The SNF consists principally of irradiated metallic uranium, and therefore includes plutonium and mixed fission products. The primary effort will consist of removing the SNF from the storage basins in K East and K West Areas, placing in multicanister overpacks, vacuum drying, conditioning, and subsequent dry vault storage in the 200 East Area. The primary purpose and need for this action is to reduce the risks to public health and safety and to the environment. Specifically these include prevention of the release of radioactive materials into the air or to the soil surrounding the K Basins, prevention of the potential migration of radionuclides through the soil column to the nearby Columbia River, reduction of occupational radiation exposure, and elimination of the risks to the public and to workers from the deterioration of SNF in the K Basins.

  7. Centrifugal microfluidic platform for radiochemistry: potentialities for the chemical analysis of nuclear spent fuels.

    Science.gov (United States)

    Bruchet, Anthony; Taniga, Vélan; Descroix, Stéphanie; Malaquin, Laurent; Goutelard, Florence; Mariet, Clarisse

    2013-11-15

    The use of a centrifugal microfluidic platform is for the first time reported as an alternative to classical chromatographic procedures for radiochemistry. The original design of the microfluidic platform has been thought to fasten and simplify the prototyping process with the use of a circular platform integrating four rectangular microchips made of thermoplastic. The microchips, dedicated to anion-exchange chromatographic separations, integrate a localized monolithic stationary phase as well as injection and collection reservoirs. The results presented here were obtained with a simplified simulated nuclear spent fuel sample composed of non-radioactive isotopes of Europium and Uranium, in proportion usually found for uranium oxide nuclear spent fuel. While keeping the analytical results consistent with the conventional procedure (extraction yield for Europium of ≈97%), the use of the centrifugal microfluidic platform allowed to reduce the volume of liquid needed by a factor of ≈250. Thanks to their unique "easy-to-use" features, centrifugal microfluidic platforms are potential successful candidates for the downscaling of chromatographic separation of radioactive samples (automation, multiplexing, easy integration in glove-boxes environment and low cost of maintenance).

  8. Lab-scale demonstration of the UREX+1a process using spent nuclear fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, C.; Vandegrift, G. F.; Regalbuto, M. C.; Bakel, A.; Bowers, D.; Gelia, A. V.; Hebden, A. S.; Maggos, L. E.; Stepinski, D.; Tsai, Y.; Laidler, J. J.; Chemical Engineering

    2007-01-01

    The Global Nuclear Energy Partnership (GNEP) is developing technologies to greatly expand repository capacity, improve proliferation resistance, and recover valuable energy that would otherwise be discarded; thus assuring a stable energy supply for the future. An important element of this initiative is the separation of key radionuclides followed by either superior waste-disposal forms and/or transmutation of long-lived isotopes. To that end, the GNEP is developing advanced fuel reprocessing systems that separate key radionuclides from spent fuel. One of these systems is the UREX+1a process. The UREX+1a process is a series of four solvent-extraction flowsheets that perform the following operations: (1) recovery of U and Tc (UREX), (2) recovery of Cs and Sr (CCD-PEG), and (3) recovery of TRU and rare earth elements (TRUEX), and (4) separation of TRU elements from the rare earths (TALSPEAK). This paper discusses the results of the demonstration of the UREX, TRUEX, and TALSPEAK processes using spent nuclear fuel, as well as future development needs and plans.

  9. Modification of spent filter treat method at Uljin nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jong rak; Ha, Jong Hyun [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

    1999-07-01

    Spent resin drying system(SRDS) and concentrated waste drying system (CWDS) have been introduced into Uljin NPP designed to treat spent resin, concentrated waste and spent filter at the waste cementation facility. Spent filter has been the only waste to treat at the cementation facility so that it is to become difficult to be provided the C-type containers needed to treat spent filter. To solve this problem, two programs including to change the spent filter container have been considered. One os to treat spent filter at the cementation facility using DOT - 17H drum, and another is to contain several spent filters in a High Integrated Container(HIC). (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-04

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

  11. The Decay of Communism: Managing Spent Nuclear Fuel in the Soviet Union, 1937-1991

    Energy Technology Data Exchange (ETDEWEB)

    Hoegselius, Per (History of Science and Technology, Royal Inst. of Technology, Stockholm (Sweden)), e-mail: perho@kth.se

    2010-09-15

    The historical evolution of spent nuclear fuel (SNF) decision-making in Western Europe and North America is already fairly well-known. For the former socialist countries of Eastern Europe, and in particular the Soviet Union, we know less. There have recently been several good studies of Soviet nuclear power history (e.g. Schmid 2004, 2006, Josephson 2005), but none of them has gone into any depth when it comes to SNF, but rather focused on nuclear power reactors, public acceptance, the role of the media, etc. There are also several good overviews available that problematize the radioactive legacy of the Soviet Union, including the SNF and waste issue, but these studies do not address the historical dynamics and evolution of SNF management over a longer period of time; in other words, they fail to explain how and why the present state of affairs have actually come into being. The aim of this paper is to provide historical insight into the dynamics of SNF decision-making in the Soviet Union, from the origins of nuclear engineering in the 1930s to the collapse of the country in 1991. The nuclear fuel system can be described as a large technical system with a variety of interrelated components. The system is 'large' both because it involves key links between geographically disperse activities, and because it involves a variety of technologies, organizations and people that influence the dynamics and evolution of the system. Soviet SNF history is of particular interest in this context, with a nuclear fuel system that was the most complex in the world. The USSR was a pioneer within nuclear power and developed a variety of reactor designs and technologies for uranium mining, conversion and enrichment, as well as for transport, treatment, storage and reprocessing of spent nuclear fuel. It explored both military and civil uses of the atom, and an enormous amount of people and organizations were involved in realizing highly ambitious nuclear programmes. The USSR is

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

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.

    2000-02-10

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

  13. Geological Disposal Options for the Radioactive Wastes from a Recycling Process of Spent Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. Y.; Choi, H. J.; Lee, M. S.; Jeong, J. T.; Choi, J. W.; Kim, S. K.; Cho, D. K.; Kuk, D. H.; Cha, J. H

    2008-10-15

    The electricity from the nuclear power plants is around 40 % of total required electricity in Korea and according to the energy development plan, the proportion will be raised about 60 % in near future. To implement this plan, the most important factor is the back-end fuel cycle, namely the safe management of the spent fuel or high level radioactive wastes from the nuclear power plants. Various researches are being carried out to manage the spent fuel effectively in the world. In our country, as one of the management alternatives which is more effective and non-proliferation, pyro-processing method is being developed actively to retrieve reusable uranium and TRU, and to reduce the volume of high level waste from a Nuclear power plant. This is a new dry recycling process. In this report, the amount of various wastes and their characteristics are estimated in a Pyro-process. Based on these information, the geological disposal alternatives are developed. According to the amount and the characteristics of each waste, the concepts of waste packages and the disposal container are developed. And also from the characteristics of the radioactivity and the heat generation, multi-layer of the depth is considered to dispose these wastes. The proposed various alternatives in this report can be used as input data for design of the deep geological disposal system. And they will be improved through the application of the real site data and safety assessment in the future. After then, the final disposal concept will be selected with various assessment and the optimization will be carried out.

  14. Characterising legacy spent nuclear fuel pond materials using microfocus X-ray absorption spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Bower, W.R. [Research Centre for Radwaste Disposal, School of Earth and Environmental Sciences, The University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom); National Nuclear Laboratory, Chadwick House, Warrington Road, Birchwood Park, Warrington, WA3 6AE (United Kingdom); Centre for Radiochemistry Research, Chemistry Building, The University of Manchester, Brunswick Street, Manchester M13 9PL (United Kingdom); Morris, K. [Research Centre for Radwaste Disposal, School of Earth and Environmental Sciences, The University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom); Mosselmans, J.F.W. [Diamond Light Source, Harwell Science and Innovation Campus, Didcot, Oxfordshire OX11 0DE (United Kingdom); Thompson, O.R. [National Nuclear Laboratory, Chadwick House, Warrington Road, Birchwood Park, Warrington, WA3 6AE (United Kingdom); Banford, A.W. [National Nuclear Laboratory, Chadwick House, Warrington Road, Birchwood Park, Warrington, WA3 6AE (United Kingdom); School of Chemical Engineering and Analytical Science, The University of Manchester, Oxford Road, Manchester, M13 9PL (United Kingdom); Law, K. [Centre for Radiochemistry Research, Chemistry Building, The University of Manchester, Brunswick Street, Manchester M13 9PL (United Kingdom); Pattrick, R.A.D., E-mail: richard.pattrick@manchester.ac.uk [Research Centre for Radwaste Disposal, School of Earth and Environmental Sciences, The University of Manchester, Oxford Road, Manchester M13 9PL (United Kingdom)

    2016-11-05

    Highlights: • A sample from a spent fuel pond wall has been analysed via X-ray spectroscopy. • Autoradiography shows a patchy distribution of radioactivity on the core face. • μXAS across a ‘hot spot’ showed Sr associates with the TiO{sub 2} pigment in the paint. • Original concrete coatings prove effective at limiting radionuclide migration. • Sorption studies show Sr immobilisation by the concrete and Cs by aggregate clasts. - Abstract: Analysis of a radioactive, coated concrete core from the decommissioned, spent nuclear fuel cooling pond at the Hunterston-A nuclear site (UK) has provided a unique opportunity to study radionuclides within a real-world system. The core, obtained from a dividing wall and sampled at the fill level of the pond, exhibited radioactivity (dominantly {sup 137}Cs and {sup 90}Sr) heterogeneously distributed across both painted faces. Chemical analysis of the core was undertaken using microfocus spectroscopy at Diamond Light Source, UK. Mapping of Sr across the surface coatings using microfocus X-ray fluorescence (μXRF) combined with X-ray absorption spectroscopy showed that Sr was bound to TiO{sub 2} particles in the paint layers, suggesting an association between TiO{sub 2} and radiostrontium. Stable Sr and Cs sorption experiments using concrete coupons were also undertaken to assess their interactions with the bulk concrete in case of a breach in the coating layers. μXRF and scanning electron microscopy showed that Sr was immobilized by the cement phases, whilst at the elevated experimental concentrations, Cs was associated with clay minerals in the aggregates. This study provides a crucial insight into poorly understood infrastructural contamination in complex systems and is directly applicable to the UK’s nuclear decommissioning efforts.

  15. Final disposal of spent nuclear fuel - regulatory system and roles of different actors during the decision process

    Energy Technology Data Exchange (ETDEWEB)

    2009-03-15

    In November 2006 Swedish Nuclear Fuels Co. applied for a license to build a plant for encapsulation of spent nuclear fuels at Oskarshamn, Sweden. The company also have plans to apply, in 2009, for a license to construct a underground repository for spent nuclear fuels. KASAM arranged a seminar in November 2006 in order to describe and discuss the licensing rules and regulations and the roles of different parties in the decision making. Another objective of the seminar was to point out possible ambiguities in this process. Another interesting question under discussion was in what ways the basic data for the decision should be produced. The seminar covered the part of the process beginning with the application for a license and ending with the government approval/rejection of the application. Most time was spent on the legal aspects of the process

  16. DOCUMENTATION OF NATIONAL WEATHER CONDITIONS AFFECTING LONG-TERM DEGRADATION OF COMMERCIAL SPENT NUCLEAR FUEL AND DOE SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE

    Energy Technology Data Exchange (ETDEWEB)

    W. L. Poe, Jr.; P.F. Wise

    1998-11-01

    The U.S. Department of Energy (DOE) is preparing a proposal to construct, operate 2nd monitor, and eventually close a repository at Yucca Mountain in Nye County, Nevada, for the geologic disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). As part of this effort, DOE has prepared a viability assessment and an assessment of potential consequences that may exist if the repository is not constructed. The assessment of potential consequences if the repository is not constructed assumes that all SNF and HLW would be left at the generator sites. These include 72 commercial generator sites (three commercial facility pairs--Salem and Hope Creek, Fitzpatrick and Nine Mile Point, and Dresden and Morris--would share common storage due to their close proximity to each other) and five DOE sites across the country. DOE analyzed the environmental consequences of the effects of the continued storage of these materials at these sites in a report titled Continued Storage Analysis Report (CSAR; Reference 1 ) . The CSAR analysis includes a discussion of the degradation of these materials when exposed to the environment. This document describes the environmental parameters that influence the degradation analyzed in the CSAR. These include temperature, relative humidity, precipitation chemistry (pH and chemical composition), annual precipitation rates, annual number of rain-days, and annual freeze/thaw cycles. The document also tabulates weather conditions for each storage site, evaluates the degradation of concrete storage modules and vaults in different regions of the country, and provides a thermal analysis of commercial SNF in storage.

  17. The Swedish Radiation Protection Institute's regulations concerning the final management of spent nuclear fuel and nuclear waste - with background and comments

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-11-01

    This report presents and comments on the Swedish Radiation Protection Institute's Regulations concerning the Protection of Human Health and the Environment in connection with the Final Management of Spent Nuclear Fuel or Nuclear Waste, SSI FS 1998: 1.

  18. Study of Compton suppression for use in spent nuclear fuel assay

    Science.gov (United States)

    Bender, Sarah

    The focus of this study has been to assess Compton suppressed gamma-ray detection systems for the multivariate analysis of spent nuclear fuel. This objective has been achieved using direct measurement of samples of irradiated fuel elements in two geometrical configurations with Compton suppression systems. In order to address the objective to quantify the number of additionally resolvable photopeaks, direct Compton suppressed spectroscopic measurements of spent nuclear fuel in two configurations were performed: as intact fuel elements and as dissolved feed solutions. These measurements directly assessed and quantified the differences in measured gamma-ray spectrum from the application of Compton suppression. Several irradiated fuel elements of varying cooling time from the Penn State Breazeale Reactor spent fuel inventory were measured using three Compton suppression systems that utilized different primary detectors: HPGe, LaBr3, and NaI(Tl). The application of Compton suppression using a LaBr3 primary detector to the measurement of the current core fuel element, which presented the highest count rate, allowed four additional spectral features to be resolved. In comparison, the HPGe-CSS was able to resolve eight additional photopeaks as compared to the standalone HPGe measurement. Measurements with the NaI(Tl) primary detector were unable to resolve any additional peaks, due to its relatively low resolution. Samples of Approved Test Material (ATM) commercial fuel elements were obtained from Pacific Northwest National Laboratory. The samples had been processed using the beginning stages of the PUREX method and represented the unseparated feed solution from a reprocessing facility. Compton suppressed measurements of the ATM fuel samples were recorded inside the guard detector annulus, to simulate the siphoning of small quantities from the main process stream for long dwell measurement periods. Photopeak losses were observed in the measurements of the dissolved ATM

  19. Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities

    Energy Technology Data Exchange (ETDEWEB)

    Garvin, L.J.

    1995-11-01

    This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

  20. Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

    Science.gov (United States)

    Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

    Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

  1. Characterising legacy spent nuclear fuel pond materials using microfocus X-ray absorption spectroscopy.

    Science.gov (United States)

    Bower, W R; Morris, K; Mosselmans, J F W; Thompson, O R; Banford, A W; Law, K; Pattrick, R A D

    2016-11-05

    Analysis of a radioactive, coated concrete core from the decommissioned, spent nuclear fuel cooling pond at the Hunterston-A nuclear site (UK) has provided a unique opportunity to study radionuclides within a real-world system. The core, obtained from a dividing wall and sampled at the fill level of the pond, exhibited radioactivity (dominantly (137)Cs and (90)Sr) heterogeneously distributed across both painted faces. Chemical analysis of the core was undertaken using microfocus spectroscopy at Diamond Light Source, UK. Mapping of Sr across the surface coatings using microfocus X-ray fluorescence (μXRF) combined with X-ray absorption spectroscopy showed that Sr was bound to TiO2 particles in the paint layers, suggesting an association between TiO2 and radiostrontium. Stable Sr and Cs sorption experiments using concrete coupons were also undertaken to assess their interactions with the bulk concrete in case of a breach in the coating layers. μXRF and scanning electron microscopy showed that Sr was immobilized by the cement phases, whilst at the elevated experimental concentrations, Cs was associated with clay minerals in the aggregates. This study provides a crucial insight into poorly understood infrastructural contamination in complex systems and is directly applicable to the UK's nuclear decommissioning efforts.

  2. Comparative analysis of radiation characteristics from various types of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kryuchkov, E.F.; Opalovsky, V.A.; Tikhomirov, G.V. [Moscow Engineering Physics Institute (State University) (Russian Federation)

    2003-07-01

    At the present time, in purposes of the most effective utilization of nuclear materials, new advanced fuel cycles are under development. These cycles imply application of uranium-plutonium, uranium-thorium and some other types of nuclear fuel. However, it is obvious that the parameters of new nuclear fuel (NF) types will be quite different from those for traditional NF types. These differences can affect significantly the conditions for storage, transportation and reprocessing of spent nuclear fuel (SNF). So, it is necessary to carry out a comparative analysis of radiation characteristics for various NF types at different stages of nuclear fuel cycle (NFC). The present paper addresses radiation properties of the following NF types: UO{sub 2}, UO{sub 2}-PuO{sub 2}, ThO{sub 2}-PaO{sub 2}-UO{sub 2}. Numerical studies have been carried out to determine radiation properties of these NF types at the following NFC stages: radiation properties of NF directly before and after irradiation in the reactor core, after different cooling time, radiation properties of uranium and plutonium fractions after chemical separation, radiation properties of NF re-fabricated for recycle, radiation properties of NF after the second and third recycles. The computer code package SCALE is used for evaluating the radiation properties of different SNF types. Finally, the following major conclusions can be made: 1) Correct description of SNF radiation and dosimetric properties requires available benchmark data on contents of heavy nuclides in SNF; 2) ThO{sub 2}-PaO{sub 2}-UO{sub 2} fuel demonstrates an important feature: internal transmutation of minor actinides provided the ultra-high fuel burn-up is achieved.

  3. Potential applications of sonochemistry in spent nuclear fuel reprocessing: a short review.

    Science.gov (United States)

    Nikitenko, S I; Venault, L; Pflieger, R; Chave, T; Bisel, I; Moisy, P

    2010-08-01

    The industrial treatment of spent nuclear fuel is based upon a hydrometallurgical process in nitric acid medium. In order to minimize the volume of radioactive waste it seems interesting to generate the reactive species in situ in such solutions using ultrasonic irradiation without addition of salt-forming reagents. This review summarizes for the first time the versatile sonochemical processes with uranium, neptunium and plutonium in homogeneous nitric acid solutions and heterogeneous systems. The dissolution of refractory solids, ultrasonically driven liquid-liquid extraction and the sonochemical degradation of the volatile products of organic solvent radiolysis issued from PUREX process are considered. Also the guidelines for required further work to ensure successful application of the studied processes at industrial scale are discussed.

  4. Potential applications of sono-chemistry in spent nuclear fuel reprocessing: A short review

    Energy Technology Data Exchange (ETDEWEB)

    Nikitenko, S. I.; Pflieger, R.; Chave, T. [CEA Marcoule, CNRS, UMII, ICSM, ENSCM, UMR 5257, F-30207 Bagnols Sur Ceze (France); Venault, L.; Bisel, I.; Moisy, P. [CEA Marcoule, DEN DRCP, F-30207 Bagnols Sur Ceze (France)

    2010-07-01

    The industrial treatment of spent nuclear fuel is based upon a hydrometallurgical process in nitric acid medium. In order to minimize the volume of radioactive waste it scorns interesting to generate the reactive species in situ in such solutions using ultrasonic irradiation without addition of salt-forming reagents. This review summarizes for the first time the versatile sono-chemical processes with uranium, neptunium and plutonium in homogeneous nitric acid solutions and heterogeneous systems. The dissolution of refractory solids, ultrasonically driven liquid-liquid extraction and the sono-chemical degradation of the volatile products of organic solvent radiolysis issued from PUREX process are considered. Also the guidelines for required further work to ensure successful application of the studied processes at industrial scale are discussed. (authors)

  5. Radiolytic and Thermal Processes Relevant to Dry Storage of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Theodore E. Madey

    2001-10-01

    Characterize the effects of temperature and radiation processes on the interactions of H20 with oxide surfaces. Our experiments focused on the fundamental interaction of H20 molecules with surfaces of U02. We characterized the surface chemistry of adsorbed H2O using thermal desorption methods and radiotracer methods, as well as x-ray photoelectron spectroscopy (XPS) and low energy ion scattering (LEIS). In parallel with these measurements of thermal effects, we examined the effects of secondary electrons and high-energy photons on hydrogen and oxygen generation and, and how this related to corrosion of spent nuclear fuel. These studies concentrated on neutral and ionic (cation and anion) desorption products of low-energy electron irradiation of water-covered UO2.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, C.N.

    1990-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Dewes, J.

    2014-02-24

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

  8. Spent nuclear fuel corrosion: The application of ICP-MS to direct actinide analysis

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R. [Caledon-Consult AB, Nykoeping (Sweden); Eklund, U.B. [Studsvik Nuclear AB, Nykoeping (Sweden)

    1995-01-01

    The ICP-MS technique has been applied to the analysis of the actinide contents of corrodant solutions from experiments performed to study the corrosion of spent nuclear fuel in simulated groundwaters. Analysis was performed directly on the solutions, without employing separation or isotope dilution techniques. The results from two analytical campaigns using natural indium and thorium internal standards are compared. Under both oxic and anoxic conditions, the U contents can be determined with good accuracy and precision. The same applies to Np and Pu under oxic conditions, where the solution concentrations range down to about 0.1 ppb. Under anoxic conditions, where solution concentrations are lower by one or two orders of magnitude, reasonable results for these two actinides can be obtained, but with much lower precision. Direct analysis of Am and Cm, however, gave unsatisfactory results, since the technique is limited by poor measurement statistics and background uncertainty.

  9. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    2000-02-03

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

  10. Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    1999-07-02

    This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, Cold Vacuum Drying Facility Design Requirements, Rev. 4, and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-22

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

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

  14. Available reprocessing and recycling services for research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tozser, Sandor; Marshall, Frances M.; Adelfang, Pablo; Bradley, Edward [International Atomic Energy Agency, Vienna (Austria); Budu, Madalina Elena [SOSNY Research and Development Company, Moscow (Russian Federation); Chiguer, Mustapha [AREVA, Paris La Defense (France)

    2016-03-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. In the future inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. The IAEA, based on the experience gained during the decade of international cooperation in supporting the objectives of the HEU take-back programmes, drew up a report presenting available reprocessing and recycling services for RR SNF. This paper gives an overview of the report, which will address all aspects of reprocessing and recycling services for RR SNF.

  15. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Features, events and processes 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Features, Events and Processes sits within Posiva Oy's Safety Case 'TURVA-2012' portfolio and has the objective of presenting the main features, events and processes (FEPs) that are considered to be potentially significant for the long-term safety of the planned KBS-3V repository for spent nuclear fuel at Olkiluoto. The primary purpose of this report is to support Performance Assessment, Formulation of Radionuclide Release Scenarios, Assessment of the Radionuclide Release Scenarios for the Repository System and Biosphere Assessment by ensuring that the scenarios are comprehensive and take account of all significant FEPs. The main FEPs potentially affecting the disposal system are described for each relevant subsystem component or barrier (i.e. the spent nuclear fuel, the canister, the buffer and tunnel backfill, the auxiliary components, the geosphere and the surface environment). In addition, a small number of external FEPs that may potentially influence the evolution of the disposal system are described. The conceptual understanding and operation of each FEP is described, together with the main features (variables) of the disposal system that may affect its occurrence or significance. Olkiluoto-specific issues are considered when relevant. The main uncertainties (conceptual and parameter/data) associated with each FEP that may affect understanding are also documented. Indicative parameter values are provided, in some cases, to illustrate the magnitude or rate of a process, but it is not the intention of this report to provide the complete set of numerical values that are used in the quantitative safety assessment calculations. Many of the FEPs are interdependent and, therefore, the descriptions also identify the most important direct couplings between the FEPs. This information is used in the formulation of scenarios to ensure the conceptual models and calculational cases are both comprehensive and representative. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  17. Cavern/Vault Disposal Concepts and Thermal Calculations for Direct Disposal of 37-PWR Size DPCs

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Clayton, Daniel James [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-03-01

    This report provides two sets of calculations not presented in previous reports on the technical feasibility of spent nuclear fuel (SNF) disposal directly in dual-purpose canisters (DPCs): 1) thermal calculations for reference disposal concepts using larger 37-PWR size DPC-based waste packages, and 2) analysis and thermal calculations for underground vault-type storage and eventual disposal of DPCs. The reader is referred to the earlier reports (Hardin et al. 2011, 2012, 2013; Hardin and Voegele 2013) for contextual information on DPC direct disposal alternatives.

  18. Risk Analysis on Safety Injection Test of PWR Nuclear Power Plant%压水堆核电厂安注试验风险研究

    Institute of Scientific and Technical Information of China (English)

    徐永华

    2012-01-01

    During the safety injection (SI) test in the PWR nuclear power plant, full water in pressurizer and high-high level, low-low level in SG may take place. This paper analyzes the risks and response measures in the SI test. Focusing on the full scale problem of pressurizer thermal calibration level gauge in the SI test of nuclear power plant, the test process is analyzed and the problems that should be noted in the SI test are summed up.%在压水堆核电厂安注试验期间,可能出现稳压器满水及主蒸汽发生器(SG)高高、低低液位等问题.本文分析了安注试验的风险及应对措施;针对某核电厂进行安注试验时稳压器热态标定液位计满量程问题,对试验过程进行分析,并总结出安注试验中应注意的问题.

  19. Nuclear Waste Risk Perceptions and Attitudes in Siting a Final Repository for Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sjoeberg, Lennart [Stockholm School of Economics (Sweden). Center for Risk Research

    2006-09-15

    some extent unexplained as well. The reasons for the development towards more positive attitudes cannot be determined with any certainty at this point. However, it should be mentioned that attitude to nuclear power per se became more positive during the studied period, both in the two municipalities and in the nation as a whole. Nuclear power attitude was strongly correlated with policy intention in the present data (0.67). Yet, adding nuclear power attitude to the analysis of covariance model mentioned above did not result in a substantially weaker effect of the time factor (or of gender). This result suggests that whatever factor accounts for the changes observed, beyond the ones measured and investigated here, probably are area specific and possibly connected with the SKB program of information - a question for future research to answer.

  20. Visualized Research on Primary Loop Simulation for PWR Nuclear Power Plant%压水堆核电厂一回路仿真可视化研究

    Institute of Scientific and Technical Information of China (English)

    肖瑶; 巫英伟; 苏光辉; 秋穗正

    2013-01-01

    In this study the main equipments and the primary loop of PWR nuclear power plant (NPP) were analyzed in detail.The model of point neutron dynamics,steam generator model with two-phase drift-flux governing equations,3-zone nonequilibrium pressurizer model and 4-quadrant main pump performance model were established.Based on the above models,a NPP simulation program was developed by using mixed programming with FORTRAN90 and Visual C++.The simulation program is of capability to achieve visualized simulation for the main equipments in primary loop and entire system of PWR nuclear power plant.It provides not only the visualized functions of real-time plotting,zooming,etc.,but also the output of numerical results with standard picture and/or text formatting files.Besides,the program was validated by comparing the calculation results of the program developed by authors and those of RELAP5/MOD3.0.%对压水堆核电厂一回路系统及主要设备进行了详细分析,建立了点堆中子动力学模型、两相漂移流蒸汽发生器模型、三区不平衡稳压器模型和主循环泵四象限特性模型,并以此为基础使用FORTRAN90语言和Visual C++语言通过混合编程的方法开发了核电厂仿真分析程序,实现了对压水堆核电厂一回路主要设备及全系统的可视化仿真计算.软件提供实时绘图、缩放等可视化功能,还提供了数据结果的标准图片格式和标准文本格式输出.通过将程序的计算结果与RELAP5/MOD3.0计算结果进行比较,对程序的可靠性进行了验证.

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

    Energy Technology Data Exchange (ETDEWEB)

    Patrick, W.C. (comp.)

    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.

  2. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

  3. Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  4. Sequestration of radioactive iodine in silver-palladium phases in commercial spent nuclear fuel

    Science.gov (United States)

    Buck, Edgar C.; Mausolf, Edward J.; McNamara, Bruce K.; Soderquist, Chuck Z.; Schwantes, Jon M.

    2016-12-01

    Radioactive iodine is the Achilles' heel in the design for the safe geological disposal of spent uranium oxide (UO2) nuclear fuel. Furthermore, iodine's high volatility and aqueous solubility were mainly responsible for the high early doses released during the accident at Fukushima Daiichi in 2011. Studies Kienzler et al., however, have indicated that the instant release fraction (IRF) of radioiodine (131/129I) does not correlate directly with increasing fuel burn-up. In fact, there is a peak in the release of iodine at around 50-60 MW d/kgU, and with increasing burn-up, the IRF of 131/129I decreases. The reasons for this decrease have not fully been understood. We have performed microscopic analysis of chemically processed high burn-up UO2 fuel (80 MW d/kgU) and have found recalcitrant nano-particles containing, Pd, Ag, I, and Br, possibly consistent with a high pressure phase of silver iodide in the undissolved residue. It is likely that increased levels of Ag and Pd from 239Pu fission in high burnup fuels leads to the formation of these metal halides. The occurrence of these phases in UO2 nuclear fuels may reduce the impact of long-lived 129I on the repository performance assessment calculations.

  5. Glutarimidedioxime: a complexing and reducing reagent for plutonium recovery from spent nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Xian, Liang [Radiochemistry Department, China Institute of Atomic Energy, Beijing (China); Tian, Guoxin [Radiochemistry Department, China Institute of Atomic Energy, Beijing (China); Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Beavers, Christine M.; Teat, Simon J. [Advanced Light Source, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Shuh, David K. [Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley, CA (United States)

    2016-04-04

    Efficient separation processes for recovering uranium and plutonium from spent nuclear fuel are essential to the development of advanced nuclear fuel cycles. The performance characteristics of a new salt-free complexing and reducing reagent, glutarimidedioxime (H{sub 2}A), are reported for recovering plutonium in a PUREX process. With a phase ratio of organic to aqueous of up to 10:1, plutonium can be effectively stripped from 30 % tributyl phosphate (TBP) in kerosene into 1 m HNO{sub 3} with H{sub 2}A. The complexation-reduction mechanism is illustrated with the combination of UV/Vis absorption spectra and the crystal structure of a Pu{sup IV} complex with the reagent. The fast stripping rate and the high efficiency for stripping Pu{sup IV}, through the complexation-reduction mechanism, is suitable for use in centrifugal contactors with very short contact/resident times, thereby offering significant advantages over conventional processes. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  6. Accurate determination of ⁴¹Ca concentrations in spent resins from the nuclear industry by accelerator mass spectrometry.

    Science.gov (United States)

    Nottoli, Emmanuelle; Bourlès, Didier; Bienvenu, Philippe; Labet, Alexandre; Arnold, Maurice; Bertaux, Maité

    2013-12-01

    The radiological characterisation of nuclear waste is essential for managing storage sites. Determining the concentration of Long-Lived RadioNuclides (LLRN) is fundamental for their long-term management. This paper focuses on the measurement of low (41)Ca concentrations in ions exchange resins used for primary fluid purification in Pressurised Water Reactors (PWR). (41)Ca concentrations were successfully measured by Accelerator Mass Spectrometry (AMS) after the acid digestion of resin samples, followed by radioactive decontamination and isobaric suppression through successive hydroxide, carbonate, nitrate and final CaF2 precipitations. Measured (41)Ca concentrations ranged from 0.02 to 0.03 ng/g, i.e. from 0.06 to 0.09 Bq/g. The (41)Ca/(60)Co activity ratios obtained were remarkably reproducible and in good agreement with the current ratio used for resins management.

  7. PWR ENDF/B-VII cross-section libraries for ORIGEN-ARP

    Energy Technology Data Exchange (ETDEWEB)

    McGraw, C. [Dept. of Nuclear Engineering, Texas A and M Univ., 3133 TAMU, College Station, TX 77843-3133 (United States); Ilas, G. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6172 (United States)

    2012-07-01

    New pressurized water reactor (PWR) cross-section libraries were generated for use with the ORIGEN-ARP depletion sequence in the SCALE nuclear analysis code system. These libraries are based on ENDF/B-VII nuclear data and were generated using the two-dimensional depletion sequence, TRITON/NEWT, in SCALE 6.1. The libraries contain multiple burnup-dependent cross-sections for seven PWR fuel designs, with enrichments ranging from 1.5 to 6 wt% {sup 235}U. The burnup range has been extended from the 72 GWd/MTU used in previous versions of the libraries to 90 GWd/MTU. Validation of the libraries using radiochemical assay measurements and decay heat measurements for PWR spent fuel showed good agreement between calculated and experimental data. Verification against detailed TRITON simulations for the considered assembly designs showed that depletion calculations performed in ORIGEN-ARP with the pre-generated libraries provide similar results as obtained with direct TRITON depletion, while greatly reducing the computation time. (authors)

  8. On the effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel

    Science.gov (United States)

    Trummer, Martin; Nilsson, Sara; Jonsson, Mats

    2008-08-01

    Radiation induced oxidative dissolution of UO 2 is a key process for the safety assessment of future geological repositories for spent nuclear fuel. This process is expected to govern the rate of radionuclide release to the biosphere. In this work, we have studied the catalytic effects of fission product noble metal inclusions on the kinetics of radiation induced dissolution of spent nuclear fuel. The experimental studies were performed using UO 2 pellets containing 0%, 0.1%, 1% and 3% Pd as a model for spent nuclear fuel. H 2O 2 was used as a model for radiolytical oxidants (previous studies have shown that H 2O 2 is the most important oxidant in such systems). The pellets were immersed in aqueous solution containing H 2O 2 and HCO3- and the consumption of H 2O 2 and the dissolution of uranium were analyzed as a function of H 2 pressure (0-40 bar). The noble metal inclusions were found to catalyze oxidation of UO 2 as well as reduction of surface bound oxidized UO 2 by H 2. In both cases the rate of the process increases with increasing Pd content. The reduction process was found to be close to diffusion controlled. This process can fully account for the inhibiting effect of H 2 observed in several studies on spent nuclear fuel dissolution.

  9. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Kaushik [ORNL; Scaglione, John M [ORNL

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  10. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, Kaushik [ORNL; Scaglione, John M [ORNL

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  11. Contribution of Energetically Reactive Surface Features to the Dissolution of CeO2 and ThO2 Analogues for Spent Nuclear Fuel Microstructures

    OpenAIRE

    Corkhill, C.; Myllykyla, E.; Bailey, D. J.; Thornber, S.M.; Qi, J.; Maldonado, P.; Stennett, M.C.; Hamilton, A.; Hyatt, N.C.

    2014-01-01

    In the safety case for the geological disposal of nuclear waste, the release of radioactivity from the repository is controlled by the dissolution of the spent fuel in groundwater. There remain several uncertainties associated with understanding spent fuel dissolution, including the contribution of energetically reactive surface sites to the dissolution rate. In this study, we investigate how surface features influence the dissolution rate of synthetic CeO2 and ThO2, spent nuclear fuel analog...

  12. Spent fuel characteristics & disposal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M.

    1996-06-01

    The fuel used in commercial nuclear power reactors is uranium, generally in the form of an oxide. The gas-cooled reactors developed in England use metallic uranium enclosed in a thin layer of Magnox. Since this fuel must be processed into a more stable form before disposal, we will not consider the characteristics of the Magnox spent fuel. The vast majority of the remaining power reactors in the world use uranium dioxide pellets in Zircaloy cladding as the fuel material. Reactors that are fueled with uranium dioxide generally use water as the moderator. If ordinary water is used, the reactors are called Light Water Reactors (LWR), while if water enriched in the deuterium isotope of hydrogen is used, the reactors are called Heavy Water reactors. The LWRs can be either pressurized reactors (PWR) or boiling water reactors (BWR). Both of these reactor types use uranium that has been enriched in the 235 isotope to about 3.5 to 4% total abundance. There may be minor differences in the details of the spent fuel characteristics for PWRs and BWRs, but for simplicity we will not consider these second-order effects. The Canadian designed reactor (CANDU) that is moderated by heavy water uses natural uranium without enrichment of the 235 isotope as the fuel. These reactors run at higher linear power density than LWRs and produce spent fuel with lower total burn-up than LWRs. Where these difference are important with respect to spent fuel management, we will discuss them. Otherwise, we will concentrate on spent fuel from LWRs.

  13. BEACON/MOD2A: a computer program for subcompartment analysis of nuclear reactor containment. A user's manual. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Wells, R. A.

    1979-03-01

    The BEACON code is a Best Estimate Advanced Containment code which being developed by EG and G, Idaho, Inc., at the Idaho National Engineering Laboratory. The program is designed to perform a best estimate analysis of the flow of a mixture of air, water, and steam in a nuclear reactor containment system under loss-of-coolant accident conditions. The code can simulate two-component, two-phase fluid flow in complex geometries using a combination of two-dimensional, one-dimensional, and lumped-parameter representations for the various parts of the system. The current version of BEACON, which is designated BEACON/MOD2A, contains mass and heat transfer models for wall film and for wall conduction. It is suitable for the evaluation of short term transients in PWR dry containment systems. This manual describes the models employed in BEACON/MOD2A and specifies code implementation requirements. It provides application information for input data preparation and for output data interpretation.

  14. On the performance of an artificial bee colony optimization algorithm applied to the accident diagnosis in a PWR nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Iona Maghali S. de; Schirru, Roberto; Medeiros, Jose A.C.C., E-mail: maghali@lmp.ufrj.b, E-mail: schirru@lmp.ufrj.b, E-mail: canedo@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2009-07-01

    The swarm-based algorithm described in this paper is a new search algorithm capable of locating good solutions efficiently and within a reasonable running time. The work presents a population-based search algorithm that mimics the food foraging behavior of honey bee swarms and can be regarded as belonging to the category of intelligent optimization tools. In its basic version, the algorithm performs a kind of random search combined with neighborhood search and can be used for solving multi-dimensional numeric problems. Following a description of the algorithm, this paper presents a new event classification system based exclusively on the ability of the algorithm to find the best centroid positions that correctly identifies an accident in a PWR nuclear power plant, thus maximizing the number of correct classification of transients. The simulation results show that the performance of the proposed algorithm is comparable to other population-based algorithms when applied to the same problem, with the advantage of employing fewer control parameters. (author)

  15. Resolving the H 2 effect on radiation induced dissolution of UO 2-based spent nuclear fuel

    Science.gov (United States)

    Trummer, Martin; Jonsson, Mats

    2010-01-01

    In recent years, the impact of H2 on α-radiation induced dissolution of UO2-based spent nuclear fuel has been studied and debated extensively. Experimental results on the effect of H2 on the concentration of H2O2 during α-radiolysis have been shown to disagree with numerical simulations. For this reason, the reaction scheme used in simulations of aqueous radiation chemistry has sometimes been questioned. In this work, we have studied the impact of H2 on the H2O2 concentration in α-irradiated aqueous solution using numerical simulations. The effects of H2 pressure, α-dose rate and HCO3- concentration were investigated by performing systematic variations in these parameters. The simulations show that the discrepancy between the previously published experimental result and numerical simulations is due to the use of a homogeneous dose rate (the energy is assumed to be equally distributed in the whole volume). Taking the actual dose rate of the α-irradiated volume into account, the simulation is in perfect agreement with the experimental results. This shows that the H2 effect is strongly α-dose rate dependent, and proves the reliability of the reaction scheme used in the simulations. The simulations also show that H2 influences the H2O2 concentration under α-radiolysis. The magnitude of the effect depends on the dose rate and the H2 pressure as well as on the concentration of HCO 3-. The impact of the radiolytic H2 effect on the rate of α-radiation induced dissolution of spent nuclear fuel is discussed along with other (α- and γ-) radiation induced processes capable of reducing the concentration of uranium in solution. The radiolytic H2 effect is quantitatively compared to the previously presented noble metal catalyzed H2 effect. This comparison shows that the noble metal catalyzed H2 effect is far more efficient than the radiolytic H2 effect. Reduction of U(VI) in solution due to low dose rate γ-radiolysis in the presence of H2 is proposed to be the cause of

  16. Design and Retrofit of Radiation Monitoring System for the PWR Nuclear Power Plant%压水堆核电厂辐射监测系统的设计与改造

    Institute of Scientific and Technical Information of China (English)

    张涛; 熊国华; 郎玉凯; 郭伟

    2011-01-01

    辐射监测系统是压水堆核电厂安全运行的重要保障,研究压水堆核电厂辐射监测系统的设计方法和原则,对于提高压水堆核电厂辐射监测系统的设计水平,减少改造风险至关重要.根据核电厂的法规和设计规范,结合大亚湾核电厂辐射监测系统的设计与改造经验,提出了压水堆核电厂辐射监测系统的一般设计原则和要求,并简要介绍了大亚湾核电厂辐射监测系统的改造措施及方法.%Radiation monitoring system is important for the PWR nuclear power plant, and the research of design methods and principles for the radiation monitoring system can greatly improve the design ability of the system for PWR nuclear power plant, and reduce the risk of system retrofit. According to the Nuclear power plant regulations and design specifications, and taking the design and retrofit experience of the radiation monitoring system in Daya Bay Nuclear Power Plant into account, the general design principles and requirements of the radiation monitoring system in the PWR nuclear power plant is proposed, and the retrofit method of the radiation monitoring system in Daya Bay Nuclear Power Plant is briefly introduced.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-03-31

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

  18. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Paul; Neall, Fiona; Snellman, Margit; Pastina, Barbara; Nordman, Henrik; Johnson, Lawrence; Hjerpe, Thomas

    2008-03-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. This safety assessment is summarised in the present report. The scientific basis of the safety assessment includes around 30 years of scientific RandD and technical development in the Swedish and Finnish KBS-3V programmes. Much of this scientific basis is directly applicable to KBS-3H. This has allowed the KBS-3H safety studies to focus on those issues that are unique to this design alternative, identified in a systematic 'difference analysis' of KBS-3H and KBS-3V. This difference analysis has shown that the key differences in the evolution and performance of KBS-3H and KBS-3V relate mainly to the engineered barrier system and to the impact of local variations in the rate of groundwater inflow on buffer saturation along the KBS-3H deposition drifts. No features or processes specific to KBS-3H have been identified that could lead to a loss or substantial degradation of the safety functions of the engineered barriers over a million year time frame. Radionuclide release from the repository near field in the

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

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.; Deible, R.

    2011-04-27

    The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report by the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The

  1. Estimation of the Waste Mass from a Pyro-Process of Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo (and others)

    2008-04-15

    Pyro-Process is now developing to retrieve reusable uranium and TRU, and to reduce the volume of high level waste from a nuclear power plant. In this situation, it is strongly required for the estimation of expected masses and their physical properties of the wastes. In this report, the amount of wastes and their physical properties are presupposed through some assumptions in regard to 10MTHM of Oxide Fuel with 4.5wt% U-235, 45,000 MWD/MTU, and 5yrs cooling. The produced wastes can be divided into three categories such as metal, CWF(Ceramic Waste Form), and VWF(Vitrified Waste Form). The 42 nuclrides in a spent nuclear fuel are distributed into the waste categories on the their physical and thermodynamic properties when they exist in metal, oxide, or chloride forms. The treated atomic groups are Uranium, TRU, Noble metal, Rare earth, Alkali metal, Halogens, and others. The mass of each waste is estimated by the distribution results. The off-gas waste is included into a CWF. The heat generations by the wastes in this Pyro-Process are calculated using a ORIGEN-ARP program. It is possible to estimate the amounts of wastes and their heat generation rates in this Pyro-Process analysis. These information are very helpful to design a waste container and its quantity also can be determined. The number of container and its heat generation rate will be key factor for the construction of interim storage facilities including a underground disposal site.

  2. Water corrosion of spent nuclear fuel: radiolysis driven dissolution at the UO2/water interface.

    Science.gov (United States)

    Springell, Ross; Rennie, Sophie; Costelle, Leila; Darnbrough, James; Stitt, Camilla; Cocklin, Elizabeth; Lucas, Chris; Burrows, Robert; Sims, Howard; Wermeille, Didier; Rawle, Jonathan; Nicklin, Chris; Nuttall, William; Scott, Thomas; Lander, Gerard

    2015-01-01

    X-ray diffraction has been used to probe the radiolytic corrosion of uranium dioxide. Single crystal thin films of UO(2) were exposed to an intense X-ray beam at a synchrotron source in the presence of water, in order to simultaneously provide radiation fields required to split the water into highly oxidising radiolytic products, and to probe the crystal structure and composition of the UO(2) layer, and the morphology of the UO(2)/water interface. By modeling the electron density, surface roughness and layer thickness, we have been able to reproduce the observed reflectivity and diffraction profiles and detect changes in oxide composition and rate of dissolution at the Ångström level, over a timescale of several minutes. A finite element calculation of the highly oxidising hydrogen peroxide product suggests that a more complex surface interaction than simple reaction with H(2)O(2) is responsible for an enhancement in the corrosion rate directly at the interface of water and UO(2), and this may impact on models of long-term storage of spent nuclear fuel.

  3. The influence of hydrogen peroxide and hydrogen on the corrosion of simulated spent nuclear fuel.

    Science.gov (United States)

    Razdan, Mayuri; Shoesmith, David W

    2015-01-01

    The synergistic influence between H(2)O(2) and H(2) on the corrosion of SIMFUEL (simulated spent nuclear fuel) has been studied in solutions with and without added HCO(3)(-)/CO(3)(2-). The response of the surface to increasing concentrations of added H(2)O(2) was monitored by measuring the corrosion potential in either Ar or Ar/H(2)-purged solutions. Using X-ray photoelectron spectroscopy it was shown that the extent of surface oxidation (U(V) + U(VI) content) was directly related to the corrosion potential. Variations in corrosion potential with time, redox conditions, HCO(3)(-)/CO(3)(2-) concentration, and convective conditions showed that surface oxidation induced by H(2)O(2) could be reversed by reaction with H(2), the latter reaction occurring dominantly on the noble metal particles in the SIMFUEL. For sufficiently large H(2)O(2) concentrations, the influence of H(2) was overwhelmed and irreversible oxidation of the surface to U(VI) occurred. Subsequently, corrosion was controlled by the chemical dissolution rate of this U(VI) layer.

  4. Uranium Oxide Rate Summary for the Spent Nuclear Fuel (SNF) Project (OCRWM)

    Energy Technology Data Exchange (ETDEWEB)

    PAJUNEN, A.L.

    2000-09-20

    The purpose of this document is to summarize the uranium oxidation reaction rate information developed by the Hanford Spent Nuclear Fuel (SNF) Project and describe the basis for selecting reaction rate correlations used in system design. The selection basis considers the conditions of practical interest to the fuel removal processes and the reaction rate application during design studies. Since the reaction rate correlations are potentially used over a range of conditions, depending of the type of evaluation being performed, a method for transitioning between oxidation reactions is also documented. The document scope is limited to uranium oxidation reactions of primary interest to the SNF Project processes. The reactions influencing fuel removal processes, and supporting accident analyses, are: uranium-water vapor, uranium-liquid water, uranium-moist air, and uranium-dry air. The correlation selection basis will consider input from all available sources that indicate the oxidation rate of uranium fuel, including the literature data, confirmatory experimental studies, and fuel element observations. Trimble (2000) summarizes literature data and the results of laboratory scale experimental studies. This document combines the information in Trimble (2000) with larger scale reaction observations to describe uranium oxidation rate correlations applicable to conditions of interest to the SNF Project.

  5. Development of flaw acceptance criteria for aging management of spent nuclear fuel multi-purpose canisters

    Energy Technology Data Exchange (ETDEWEB)

    Lam, Poh -Sang [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology; Sindelar, Robert L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic in-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  6. Development of flaw acceptance criteria for aging management of spent nuclear fuel multiple-purpose canisters

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology; Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Materials Science and Technology

    2015-03-09

    A typical multipurpose canister (MPC) is made of austenitic stainless steel and is loaded with spent nuclear fuel assemblies. The canister may be subject to service-induced degradation when it is exposed to aggressive atmospheric environments during a possibly long-term storage period if the permanent repository is yet to be identified and readied. Because heat treatment for stress relief is not required for the construction of an MPC, stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic In-service Inspection. The first-order instability flaw sizes has been determined with bounding flaw configurations, that is, through-wall axial or circumferential cracks, and part-through-wall long axial flaw or 360° circumferential crack. The procedure recommended by the American Petroleum Institute (API) 579 Fitness-for-Service code (Second Edition) is used to estimate the instability crack length or depth by implementing the failure assessment diagram (FAD) methodology. The welding residual stresses are mostly unknown and are therefore estimated with the API 579 procedure. It is demonstrated in this paper that the residual stress has significant impact on the instability length or depth of the crack. The findings will limit the applicability of the flaw tolerance obtained from limit load approach where residual stress is ignored and only ligament yielding is considered.

  7. Systematic impact of spent nuclear fuel on θ13 sensitivity at reactor neutrino experiment

    Institute of Scientific and Technical Information of China (English)

    AN Feng-Peng; TIAN Xin-Chun; ZHAN Liang; CAO Jun

    2009-01-01

    Reactor neutrino oscillation experiments, such as Daya Bay, Double Chooz and RENO are designed to determine the neutrino mixing angle θ13 with a sensitivity of 0.01--0.03 in sin2 2θ13 at 90% confidence level, an improvement over the current limit by more than one order of magnitude. The control of systematic uncertainties is critical to achieving the sin2 2θ13 sensitivity goal of these experiments. Antineutrinos emitted from spent nuclear fuel (SNF) would distort the soft part of energy spectrum and may introduce a non-negligible systematic uncertainty. In this article, a detailed calculation of SNF neutrinos is performed taking account of the operation of a typical reactor and the event rate in the detector is obtained. A further estimation shows that the event rate contribution of SNF neutrinos is less than 0.2% relative to the reactor neutrino signals. A global χ2 analysis shows that this uncertainty will degrade the θ13 sensitivity at a negligible level.

  8. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  9. Radioecological consequences of a potential accident during transport of spent nuclear fuel along an Arctic coastline.

    Science.gov (United States)

    Iosjpe, M; Reistad, O; Amundsen, I B

    2009-02-01

    This article presents results pertaining to a risk assessment of the potential consequences of a hypothetical accident occurring during the transportation by ship of spent nuclear fuel (SNF) along an Arctic coastline. The findings are based on modelling of potential releases of radionuclides, radionuclide transport and uptake in the marine environment. Modelling work has been done using a revised box model developed at the Norwegian Radiation Protection Authority. Evaluation of the radioecological consequences of a potential accident in the southern part of the Norwegian Current has been made on the basis of calculated collective dose to man, individual doses for the critical group, concentrations of radionuclides in seafood and doses to marine organisms. The results of the calculations indicate a large variability in the investigated parameters above mentioned. On the basis of the calculated parameters the maximum total activity ("accepted accident activity") in the ship, when the parameters that describe the consequences after the examined potential accident are still in agreement with the recommendations and criterions for protection of the human population and the environment, has been evaluated.

  10. DOE Spent Nuclear Fuel Information in Support of TSPA-SR

    Energy Technology Data Exchange (ETDEWEB)

    H. H. Loo

    1999-08-01

    The Department of Energy (DOE) Office of Civilian Radioactive Waste Management (RW) has started the recommendation (SR) effort to show that Yucca Mountain could be selected as the first geologic repository for spent nuclear fuel (SNF) and high-level waste. One component of the site recommendation will be a total system performance assessment (TSPA), based on the design concept and the scientific data and analysis available, describing the repository's probable behavior relative to the overall system performance standards. Thus, all the data collected from the Exploratory Studies Facilities to-date have been incorporated into the latest TSPA model. To ensure that the DOE-owned SNF continues to be acceptable for disposal in the repository, it will be included in the TSPA-SR evaluation. A number of parameters are needed in the TSPA-SR models to predict the performance of the DOE-owned SNF materials placed into the potential repository. This report documents all of the basis and/or derivation for each of these parameters. A number of properties were not readily available at the time the TSPA-SR data were requested. Thus, expert judgement and opinion were used to determine a best property value. The performance of the DOE-owned SNF will be published as part of the TSPA-SR report.

  11. Modeling evaporation from spent nuclear fuel storage pools: A diffusion approach

    Science.gov (United States)

    Hugo, Bruce Robert

    Accurate prediction of evaporative losses from light water reactor nuclear power plant (NPP) spent fuel storage pools (SFPs) is important for activities ranging from sizing of water makeup systems during NPP design to predicting the time available to supply emergency makeup water following severe accidents. Existing correlations for predicting evaporation from water surfaces are only optimized for conditions typical of swimming pools. This new approach modeling evaporation as a diffusion process has yielded an evaporation rate model that provided a better fit of published high temperature evaporation data and measurements from two SFPs than other published evaporation correlations. Insights from treating evaporation as a diffusion process include correcting for the effects of air flow and solutes on evaporation rate. An accurate modeling of the effects of air flow on evaporation rate is required to explain the observed temperature data from the Fukushima Daiichi Unit 4 SFP during the 2011 loss of cooling event; the diffusion model of evaporation provides a significantly better fit to this data than existing evaporation models.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-10-20

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

  13. A Stochastic Method for Estimating the Effect of Isotopic Uncertainties in Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    2001-08-24

    This report describes a novel approach developed at the Oak Ridge National Laboratory (ORNL) for the estimation of the uncertainty in the prediction of the neutron multiplication factor for spent nuclear fuel. This technique focuses on burnup credit, where credit is taken in criticality safety analysis for the reduced reactivity of fuel irradiated in and discharged from a reactor. Validation methods for burnup credit have attempted to separate the uncertainty associated with isotopic prediction methods from that of criticality eigenvalue calculations. Biases and uncertainties obtained in each step are combined additively. This approach, while conservative, can be excessive because of a physical assumptions employed. This report describes a statistical approach based on Monte Carlo sampling to directly estimate the total uncertainty in eigenvalue calculations resulting from uncertainties in isotopic predictions. The results can also be used to demonstrate the relative conservatism and statistical confidence associated with the method of additively combining uncertainties. This report does not make definitive conclusions on the magnitude of biases and uncertainties associated with isotopic predictions in a burnup credit analysis. These terms will vary depending on system design and the set of isotopic measurements used as a basis for estimating isotopic variances. Instead, the report describes a method that can be applied with a given design and set of isotopic data for estimating design-specific biases and uncertainties.

  14. Inspection of copper canisters for spent nuclear fuel by means of ultrasound. Algorithms for ultrasonic imaging

    Energy Technology Data Exchange (ETDEWEB)

    Stepinski, Tadeusz (ed.); Engholm, Marcus; Olofsson, Tomas (Uppsala Univ., Signals and Systems, Dept. of Technical Sciences (Sweden))

    2011-07-15

    This report contains research results concerning the use of advanced ultrasound for the inspection of copper canisters for spent nuclear fuel obtained at Signals and Systems, Uppsala Univ. in 2009 and 2010. The first part of the report deals with ultrasonic imaging of damage in planar structures using Lamb waves. We present results of the first successful attempt to apply an adaptive beamformer for Lamb waves. Our algorithm is an extension of the adaptive beamformer based on minimum variance distortion less response (MVDR) approach to dispersive, multimodal Lamb waves. We present simulation and experimental results illustrating the performance of the MVDR applied to imaging artificial damage in an aluminum plate. In the second part of the report we present two extensions of the previously proposed 2D phase shift migration algorithms for enhancing resolution in ultrasonic imaging of solid objects. The first extension enables processing 3D data in order to fully utilize the resolution enhancement potential of the technique. The second extension, consists in generalizing the technique to allow for the processing of data acquired using an array instead of a previously concerned single transducer. Robustness issue related to objects having front surfaces that are slightly tilted relative to the scanning axis is also considered

  15. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  16. Progress and prospects of nuclear fuel development in Japan, (2). Progress and future plan of research and development on PWR fuel in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Yoshiaki; Abeta, Sadaaki; Aisu, Hideo; Teranishi, Tomoyuki

    1982-06-01

    13 years have elapsed since the first PWR plant started the operation in Japan, and at present, 11 PWR plants are in operation. During this period, much results of use and experience have been accumulated for the PWR fuel. The improvement and development of the fuel have been performed to meet the supply of the fuel sufficiently adaptable to the severe environment in Japan. In this paper, the evaluation of soundness and the improvement of reliability of PWR fuel made so far are reported, and the response of fuel side to long cycle operation and load following-up operation, which will be required in near future, is explained. The inspection of fuel has been performed at reactor sites for the purpose of sufficiently observing the irradiation behavior of fuel and detecting the points out of order. Effort has been exerted to perform various inspections thoroughly on total number of fuel and reflect the results to the improved design. Fuel leak scarcely occurred from the beginning, accordingly, improvement has been made to reduce the bending of fuel rods. The change of PWR fuel design, the evaluation of soundness and the improvement of reliability of PWR fuel, and the improvement for the future are reported.

  17. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  18. Toward a Greenish Nuclear Fuel Cycle: Ionic Liquids as Solvents for Spent Nuclear Fuel Reprocessing and Other Decontamination Processes for Contaminated Metal Waste

    Science.gov (United States)

    Straka, Martin

    2016-12-01

    The final disposition of spent nuclear fuel (SNF) is an area that requires innovative solutions. The use of ionic liquids (ILs) has been examined as one means to remediate SNF in a variety of different chemical environments and with different chemical starting materials. The effectiveness of various ILs for SNF reprocessing, as well as the reaction chemistry that occurs in them, is discussed.

  19. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  20. Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Alexis C. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87544 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 500 S State St., Ann Arbor, MI 48109 (United States); Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87544 (United States); Flaska, Marek; Pozzi, Sara A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 500 S State St., Ann Arbor, MI 48109 (United States)

    2014-11-11

    As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

  1. Determination of spent nuclear fuel assembly multiplication with the differential die-away self-interrogation instrument

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Alexis C. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Henzl, Vladimir; Menlove, Howard O.; Swinhoe, Martyn T.; Belian, Anthony P. [Los Alamos National Laboratory, Los Alamos, NM 87544 (United States); Flaska, Marek; Pozzi, Sara A. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States)

    2014-09-01

    We present a novel method for determining the multiplication of a spent nuclear fuel assembly with a Differential Die-Away Self-Interrogation (DDSI) instrument. The signal, which is primarily created by thermal neutrons, is measured with four {sup 3}He detector banks surrounding a spent fuel assembly. The Rossi-alpha distribution (RAD) at early times reflects coincident events from single fissions as well as fission chains. Because of this fact, the early time domain contains information about both the fissile material and spontaneous fission material in the assembly being measured. A single exponential function fit to the early time domain of the RAD has a die-away time proportional to the spent fuel assembly (SFA) multiplication. This correlation was tested by simulating assay of 44 different SFAs with the DDSI instrument. The SFA multiplication was determined with a variance of 0.7%.

  2. Engineering development of a digital replacement protection system at an operating US PWR nuclear power plant: Installation and operational experiences

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.H. [Duke Power Co., Seneca, SC (United States)

    1995-04-01

    The existing Reactor Protection Systems (RPSs) at most US PWRs are systems which reflect 25 to 30 year-old designs, components and manufacturing techniques. Technological improvements, especially in relation to modern digital systems, offer improvements in functionality, performance, and reliability, as well as reductions in maintenance and operational burden. The Nuclear power industry and the US nuclear regulators are poised to move forward with the issues that have slowed the transition to modern digital replacements for nuclear power plant safety systems. The electric utility industry is now more than ever being driven by cost versus benefit decisions. Properly designed, engineered, and installed digital systems can provide adequate cost-benefit and allow continued nuclear generated electricity. This paper describes various issues and areas related to an ongoing RPS replacement demonstration project which are pertinant for a typical US nuclear plant to consider cost-effective replacement of an aging analog RPS with a modern digital RPS. The following subject areas relative to the Oconee Nuclear Station ISAT{trademark} Demonstrator project are discussed: Operator Interface Development; Equipment Qualification; Validation and Verification of Software; Factory Testing; Field Changes and Verification Testing; Utility Operational, Engineering and Maintenance; Experiences with Demonstration System; and Ability to operate in parallel with the existing Analog RPS.

  3. Comparative analysis between measured and calculated concentrations of major actinides using destructive assay data from Ohi-2 PWR

    Directory of Open Access Journals (Sweden)

    Oettingen Mikołaj

    2015-09-01

    Full Text Available In the paper, we assess the accuracy of the Monte Carlo continuous energy burnup code (MCB in predicting final concentrations of major actinides in the spent nuclear fuel from commercial PWR. The Ohi-2 PWR irradiation experiment was chosen for the numerical reconstruction due to the availability of the final concentrations for eleven major actinides including five uranium isotopes (U-232, U-234, U-235, U-236, U-238 and six plutonium isotopes (Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242. The main results were presented as a calculated-to-experimental ratio (C/E for measured and calculated final actinide concentrations. The good agreement in the range of ±5% was obtained for 78% C/E factors (43 out of 55. The MCB modeling shows significant improvement compared with the results of previous studies conducted on the Ohi-2 experiment, which proves the reliability and accuracy of the developed methodology.

  4. Spent nuclear fuel system dynamic stability under normal conditions of transportation

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Hao; Wang, Jy-An John, E-mail: wangja@ornl.gov

    2016-12-15

    Highlights: • A conformational potential effect of fuel assembly contact interaction induced transient shock. • Complex vibration modes and vibration load intensity were observed from fuel assembly system. • The project was able to link the periodic transient shock to spent fuel fatigue strength reduction. - Abstract: In a horizontal layout of a spent nuclear fuel (SNF) assembly under normal conditions of transportation (NCT), the fuel assembly’s skeleton formed by guide tubes and spacer grids is the primary load bearing structure for carrying and transferring the vibration loads within an SNF assembly. Therefore, the integrity of guide tubes and spacer grids will dictate the vibration amplitude/intensity of the fuel assembly during transport, and must be considered when designing multipurpose purpose canister (MPC) for safe SNF transport. This paper investigates the SNF assembly deformation dynamics during normal vibration mode, as well as the transient shock mode inside the cask during NCT. Dynamic analyses were performed in the frequency domain to study frequency characteristic of the fuel assembly system and in the time domain to simulate the transient dynamic response of the fuel assembly. To further evaluate the intensity of contact interaction induced by the local contacts’ impact loading at the spacer grid, detailed models of the actual spring and dimples of the spacer grids were created. The impacts between the fuel rod and springs and dimples were simulated with a 20 g transient shock load. The associated contact interaction intensities, in terms of reaction forces, were estimated from the finite element analyses (FEA) results. The bending moment estimated from the resultant stress on the clad under 20 g transient shock can be used to define the loading in cyclic integrated reversible-bending fatigue tester (CIRFT) vibration testing for the equivalent condition. To estimate the damage potential of the transient shock to the SNF vibration

  5. Final disposal of spent nuclear fuel in Finnish bedrock - Romuvaara site report

    Energy Technology Data Exchange (ETDEWEB)

    Anttila, P. [Fortum Engineering Oy (Finland); Ahokas, H. [Fintact Oy (Finland); Front, K. [VTT Communities and Infrastructure, Espoo (Finland)] [and others

    1999-06-01

    Posiva Oy is studying the Finnish bedrock for the geological disposal of spent nuclear fuel. The study is based on the site selection research programme started originally in 1983. The programme is in accordance with the decision in principle by the Council of State in 1983 and aims at the selection of one site in 2000. Four sites, Haestholmen in Loviisa, Kivetty in Aeaenekoski, Olkiluoto in Eurajoki and Romuvaara in Kuhmo, have been studied in detail. This report summarises the results of the site investigations carried out at Romuvaara. The bedrock of Romuvaara belongs to the Archean basement complex, whose oldest parts date back over 2800 million years. The bedrock consists mainly of migmatitic banded gneisses (tonalite, leucotonalite and mica gneiss), which are cut by granodiorite and metadiabase dykes. The rocks, excluding the metadiabase, have undergone a polyphase Archaean deformation. Altogether 31 bedrock structures (R-structures) have been modelled at the investigation site, most of them representing steeply dipping fracture zones. The rock mass between the fracture zones represents what is termed `intact rock`, which is typically hard, unweathered and sparsely fractured. The R-structures are generally hydraulically more conductive than the intact rock and their mean transmissivity is 1.6 x 10{sup -7} m{sup 2}/s. The corresponding mean of the hydraulic conductivity values for the intact rock measured using a 2 m packer interval is 8 x 10{sup -12} m/s, if a lognormal distribution for all measured values is assumed. A clear decrease in hydraulic conductivity with depth has been found, for both the R-structures and the intact rock. In addition, the hydraulically conductive fractures seem to be more frequent and their transmissivities higher in the uppermost 100 - 200 m of the bedrock than at greater depths. The groundwater of Romuvaara is classified as fresh water and the Total Dissolved Solids (TDS) and chloride contents increase with depth. The chemically

  6. FY 2016 Status Report: CIRFT Testing on Spent Nuclear Fuels and Hydride Reorientation Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Bevard, Bruce B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-08-04

    This report provides a detailed description of the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) testing conducted on spent nuclear fuel (SNF) rods in FY 2016, including hydride reorientation test results. Contact-based measurement, or three-LVDT-based curvature measurement, of SNF rods has proven to be quite reliable in CIRFT testing. However, how the linear variable differential transformer (LVDT) head contacts the SNF rod may have a significant effect on the curvature measurement, depending on the magnitude and direction of rod curvature. To correct such contact/curvature issues, sensor spacing, defined as the amount of separation between the three LVDT probes, is a critical measurement that can be used to calculate rod curvature once the deflections are obtained. Recently developed CIRFT data analyses procedures were integrated into FY 2016 CIRFT testing results for the curvature measurements. The variations in fatigue life are provided in terms of moment, equivalent stress, curvature, and equivalent strain for the tested SNFs. The equivalent stress plot collapsed the data points from all of the SNFs into a single zone. A detailed examination revealed that, at same stress level, fatigue lives display a descending order as follows: H. B. Robinson Nuclear Power Station (HBR), Limerick Nuclear Power Station (LMK), mixed uranium-plutonium oxide (MOX). If looking at the strain, then LMK fuel has a slightly longer fatigue life than HBR fuel, but the difference is subtle. The knee point of endurance limit in the curve of moment and curvature or equivalent quantities is more clearly defined for LMK and HBR fuels. The treatment affects the fatigue life of specimens. Both a drop of 12 in. and radial hydride treatment (RHT) have a negative impact on fatigue life. The effect of thermal annealing on MOX fuel rods was relatively small at higher amplitude but became significant at low amplitude of moment. Thermal annealing tended to extend the fatigue life of

  7. Policies, strategies and systems for the disposal of spent nuclear fuel; Principer, strategier och system foer slutligt omhaendertagande av anvaent kaernbraensle

    Energy Technology Data Exchange (ETDEWEB)

    Grundfelt, Bertil (Kemakta Konsult AB (Sweden))

    2010-12-15

    In this report an up-to-date account is made of the status of different principles, strategies and systems for the management and disposal of spent nuclear fuel. As large scale use of nuclear power for the production of electricity began in the 1960s and 1970s, studies of various principles and strategies for the management of the spent nuclear fuel were initialised. In particular in the USA, comprehensive studies were conducted of all strategies described in this report

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

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

    Science.gov (United States)

    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. STATE OF THE ART OF DRILLING LARGE DIAMETER BOREHOLES FOR DEPOSITION OF HIGH LEVEL WASTE AND SPENT NUCLEAR FUEL

    Directory of Open Access Journals (Sweden)

    Trpimir Kujundžić

    2012-07-01

    Full Text Available Deep geological disposal is internationally recognized as the safest and most sustainable option for the long-term management of high-level radioactive waste. Mainly, clay rock, salt rock and crystalline rock are being considered as possible host rocks. Different geological environment in different countries led to the various repository concepts. Main feature of the most matured repository concept is that canisters with spent nuclear fuel are emplaced in vertical or horizontal large diameter deposition holes. Drilling technology of the deposition holes depends on repository concept and geological and geomechanical characteristics of the rock. The deposition holes are mechanically excavated since drill & blast is not a possible method due to requirements on final geometry like surface roughness etc. Different methods of drilling large diameter boreholes for deposition of high-level waste and spent nuclear fuel are described. Comparison of methods is made considering performance and particularities in technology.

  14. Information system of corrosion and mechanical properties for steels used in nuclear power plants with PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lahodova, M.; Novotny, R.; Sajdl, P. [Inst. of Chemical Technology, Prague (Czech Republic). Dept. of Power Engineering

    1998-11-01

    This paper gives information about a new developed database system which contains information about chemical constitution of steels used in nuclear power plants. It enables to hold data from corrosion tests and allows to insert graphs and pictures into the form. This system is an application of MS Access. (orig.)

  15. Hanford spent nuclear fuel project recommended path forward, volume III: Alternatives and path forward evaluation supporting documentation

    Energy Technology Data Exchange (ETDEWEB)

    Fulton, J.C.

    1994-10-01

    Volume I of the Hanford Spent Nuclear Fuel Project - Recommended Path Forward constitutes an aggressive series of projects to construct and operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. Volume II provided a comparative evaluation of four Alternatives for the Path Forward and an evaluation for the Recommended Path Forward. Although Volume II contained extensive appendices, six supporting documents have been compiled in Volume III to provide additional background for Volume II.

  16. Technical Approach and Plan for Transitioning Spent Nuclear Fuel (SNF) Project Facilities to the Environmental Restoration Program

    Energy Technology Data Exchange (ETDEWEB)

    SKELLY, W.A.

    1999-10-06

    This document describes the approach and process in which the 100-K Area Facilities are to be deactivated and transitioned over to the Environmental Restoration Program after spent nuclear fuel has been removed from the K Basins. It describes the Transition Project's scope and objectives, work breakdown structure, activity planning, estimated cost, and schedule. This report will be utilized as a planning document for project management and control and to communicate details of project content and integration.

  17. Deployment Evaluation Methodology for the Electrometallurgical Treatment of DOE-EM Spent Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ramer, Ronald James; Adams, James Paul; Rynearson, Michael Ardel; Dahl, Christian Adam

    1999-03-01

    The Department of Energy - Environmental Management (DOE-EM) National Spent Nuclear Fuel Program (NSNFP) is charged with the disposition of legacy Spent Nuclear Fuel (SNF). The NSNFP, conducted by Lockheed Martin Idaho Technology Co. (LMITCO) at the Idaho National Engineering and Environmental Laboratory (INEEL), is evaluating final disposition of SNF in the DOE complex. While direct repository disposal of the SNF is the preferred disposition option, some DOE SNF may need treatment to meet acceptance criteria at various disposition sites. Evaluations of treatment needs and options have been previously prepared, and further evaluations are ongoing activities in the DOE-EM NSNFP. The treatments may range from electrometallurgical treatment (EMT) and chemical dissolution to engineering controls. As a planning basis, a need is assumed for a treatment process, either as a primary or backup technology, that is compatible with, and cost-effective for, this portion of the DOE-EM inventory. The current planning option for treating this SNF, pending completion of development work and National Environmental Policy Act (NEPA) analysis, is the EMT process under development by Argonne National Laboratory - West (ANL-W). A decision on the deployment of the EMT is pending completion of an engineering scale demonstration currently in progress at ANL-W. Treatment options and treatment locations will depend on fuel type and location of the fuel. One of the first steps associated with selecting one or more sites for treating 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. One of the major issues associated with SNF treatment is final disposition of treatment products and associated waste streams. During conventional SNF treatment, various chemicals are added that may increase the product

  18. Tests for manufacturing technology of disposal canisters for nuclear spent fuel; Loppusijoituskapselin valmistustekniset kokeet

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy (Finland); Salonen, T. [Outokumpu Poricopper Oy (Finland); Meuronen, I. [Suomen Teknohaus Oy (Finland); Lehto, K. [Valmet Oyj Rautpohja Foundry (Finland)

    1999-06-01

    The summary and status of the results of the manufacturing technology programmes concerning the disposal canister for spent nuclear fuel conducted by Posiva Oy are given in this report. Posiva has maintained a draft plan for a disposal canister design and an assessment of potential manufacturing technologies for about ten years in Finland. Now, during the year 1999, the first full scale demonstration canister is manufactured in Finland. The technology used for manufacturing of this prototype is developed by Posiva Oy mainly in co-operation with domestic industry. The main partner in developing the manufacturing technology for the copper shell has been Outokumpu Poricopper Oy, Pori, Finland, and the main partner in developing the technology for the iron insert of the canister has been Valmet Oyj Rautpohja Foundry, Jyvaeskylae, Finland. In both areas many subcontractors have been used, predominantly domestic engineering workshops, but also some foreign subcontractors, e.g. for EB-welding, who have had large enough welding equipment. This report describes the developing programmes for canister manufacturing, evaluates the results and presents some alternative methods, and tries to evaluate the pros and contras of them. In addition, the adequacy of the achieved technological know-how is assessed in respect of the required quality of the disposal canister. The following manufacturing technologies have been the concrete topics of the development programme: Electron beam welding technology development for thick-walled copper, Casting of massive copper billets, Hot rolling of thick-walled copper plates, Hot pressing and forging in lid manufacture, Extrusion and drawing of copper tubes, Bending of copper plates by roller or press, Machining of copper, Residual stress removal by heat treatment, Non-destructive testing, Long-term strength of EB-welds, Casting and machining of the iron insert of the canister The specialists from all the main developing partner companies have

  19. Final report spent nuclear fuel retrieval system primary cleaning development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.

    1997-09-01

    Developmental testing of the primary cleaning station for spent nuclear fuel (SNF) and canisters is reported. A primary clean machine will be used to remove the gross sludge from canisters and fuel while maintaining water quality in the downstream process area. To facilitate SNF separation from canisters and minimize the impact to water quality, all canisters will be subjected to mechanical agitation and flushing with the Primary Clean Station. The Primary Clean Station consists of an outer containment box with an internally mounted, perforated wash basket. A single canister containing up to 14 fuel assemblies will be loaded into the wash basket, the confinement box lid closed, and the wash basket rotated for a fixed cycle time. During this cycle, basin water will be flushed through the wash basket and containment box to remove and entrain the sludge and carry it out of the box. Primary cleaning tests were performed to provide information concerning the removal of sludge from the fuel assemblies while in the basin canisters. The testing was also used to determine if additional fuel cleaning is required outside of the fuel canisters. Hydraulic performance and water demand requirements of the cleaning station were also evaluated. Thirty tests are reported in this document. Tests demonstrated that sludge can be dislodged and suspended sufficiently to remove it from the canister. Examination of fuel elements after cleaning suggested that more than 95% of the exposed fuel surfaces were cleaned so that no visual evidence of remained. As a result of testing, recommendations are made for the cleaning cycle. 3 refs., 16 figs., 4 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Expected environments in high-level nuclear waste and spent fuel repositories in salt

    Energy Technology Data Exchange (ETDEWEB)

    Claiborne, H.C.; Rickertsen, L.D., Graham, R.F.

    1980-08-01

    The purpose of this report is to describe the expected environments associated with high-level waste (HLW) and spent fuel (SF) repositories in salt formations. These environments include the thermal, fluid, pressure, brine chemistry, and radiation fields predicted for the repository conceptual designs. In this study, it is assumed that the repository will be a room and pillar mine in a rock-salt formation, with the disposal horizon located approx. 2000 ft (610 m) below the surface of the earth. Canistered waste packages containing HLW in a solid matrix or SF elements are emplaced in vertical holes in the floor of the rooms. The emplacement holes are backfilled with crushed salt or other material and sealed at some later time. Sensitivity studies are presented to show the effect of changing the areal heat load, the canister heat load, the barrier material and thickness, ventilation of the storage room, and adding a second row to the emplacement configuration. The calculated thermal environment is used as input for brine migration calculations. The vapor and gas pressure will gradually attain the lithostatic pressure in a sealed repository. In the unlikely event that an emplacement hole will become sealed in relatively early years, the vapor space pressure was calculated for three scenarios (i.e., no hole closure - no backfill, no hole closure - backfill, and hole closure - no backfill). It was assumed that the gas in the system consisted of air and water vapor in equilibrium with brine. A computer code (REPRESS) was developed assuming that these changes occur slowly (equilibrium conditions). The brine chemical environment is outlined in terms of brine chemistry, corrosion, and compositions. The nuclear radiation environment emphasized in this report is the stored energy that can be released as a result of radiation damage or crystal dislocations within crystal lattices.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Application of Diagnostic/Prognostic Methods to Critical Equipment for the Spent Nuclear Fuel Cleanup Program

    Energy Technology Data Exchange (ETDEWEB)

    Casazza, Lawrence O.; Jarrell, Donald B.; Koehler, Theresa M.; Meador, Richard J.; Wallace, Dale E.

    2002-02-28

    The management of the Spent Nuclear Fuel (SNF) project at the Hanford K-Basin in the 100 N Area has successfully restructured the preventive maintenance, spare parts inventory requirements, and the operator rounds data requirements. In this investigation, they continue to examine the different facets of the operations and maintenance (O&M) of the K-Basin cleanup project in search of additional reliability and cost savings. This report focuses on the initial findings of a team of PNNL engineers engaged to identify potential opportunities for reducing the cost of O&M through the application of advanced diagnostics (fault determination) and prognostics (residual life/reliability determination). The objective is to introduce predictive technologies to eliminate or reduce high impact equipment failures. The PNNL team in conjunction with the SNF engineers found the following major opportunities for cost reduction and/or enhancing reliability: (1) Provide data routing and automated analysis from existing detection systems to a display center that will engage the operations and engineering team. This display will be operator intuitive with system alarms and integrated diagnostic capability. (2) Change operating methods to reduce major transients induced in critical equipment. This would reduce stress levels on critical equipment. (3) Install a limited sensor set on failure prone critical equipment to allow degradation or stressor levels to be monitored and alarmed. This would provide operators and engineers with advance guidance and warning of failure events. Specific methods for implementation of the above improvement opportunities are provided in the recommendations. They include an Integrated Water Treatment System (IWTS) decision support system, introduction of variable frequency drives on certain pump motors, and the addition of limited diagnostic instrumentation on specified critical equipment.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Simulation of radiation induced dissolution of spent nuclear fuel using the steady-state approach. A comparison to experimental data

    Science.gov (United States)

    Nielsen, Fredrik; Ekeroth, Ella; Eriksen, Trygve E.; Jonsson, Mats

    2008-02-01

    Using the recently developed steady-state model for simulation of radiation induced dissolution of spent nuclear fuel in water we have estimated the rate of dissolution for relatively fresh fuel to 1.64 × 10 -9 mol m -2 s -1. A series of experiments have been performed on fuel fragments in deoxygenated water containing 10 mM HCO 3-. The dissolution rates obtained from these experiments range from 2.6 × 10 -10 to 1.6 × 10 -9 mol m -2 s -1. The leaching time in the experiments is 40 days or less and during this time the amount of released uranium increases linearly with time which indicates that the system has reached steady-state. The excellent agreement between the estimated dissolution rate and the dissolution rates obtained from the spent nuclear fuel leaching experiments indicates that the steady-state approach can indeed be used to predict the rate of spent nuclear fuel dissolution.

  6. Spent nuclear fuels project: FY 1995 multi-year program plan, WBS {number_sign}1.4

    Energy Technology Data Exchange (ETDEWEB)

    Denning, J.L.

    1994-09-01

    The mission of the Spent Nuclear Fuel (SNF) program is to safely, reliably, and efficiently manage, condition, transport, and store Department of Energy (DOE)-owned SNF, so that it meets acceptance criteria for disposal in a permanent repository. The Hanford Site Spent Nuclear Fuel strategic plan for accomplishing the project mission is: Establish near-term safe storage in the 105-K Basins; Complete national Environmental Policy Act (NEPA) process to obtain a decision on how and where spent nuclear fuel will be managed on the site; Define and establish alternative interim storage on site or transport off site to support implementation of the NEPA decision; and Define and establish a waste package qualified for final disposition. This report contains descriptions of the following: Work Breakdown Structure; WBS Dictionary; Responsibility Assignment Matrix; Program Logic Diagrams; Program Master Baseline Schedule; Program Performance Baseline Schedule; Milestone List; Milestone Description Sheets; Cost Baseline Summary by Year; Basis of Estimate; Waste Type Data; Planned Staffing; and Fiscal Year Work Plan.

  7. Preparation for the Recovery of Spent Nuclear Fuel (SNF) at Andreeva Bay, North West Russia - 13309

    Energy Technology Data Exchange (ETDEWEB)

    Field, D.; McAtamney, N. [Nuvia Limited (United Kingdom)

    2013-07-01

    Andreeva Bay is located near Murmansk in the Russian Federation close to the Norwegian border. The ex-naval site was used to de-fuel nuclear-powered submarines and icebreakers during the Cold War. Approximately 22,000 fuel assemblies remain in three Dry Storage Units (DSUs) which means that Andreeva Bay has one of the largest stockpiles of highly enriched spent nuclear fuel (SNF) in the world. The high contamination and deteriorating condition of the SNF canisters has made improvements to the management of the SNF a high priority for the international community for safety, security and environmental reasons. International Donors have, since 2002, provided support to projects at Andreeva concerned with improving the management of the SNF. This long-term programme of work has been coordinated between the International Donors and responsible bodies within the Russian Federation. Options for the safe and secure management of SNF at Andreeva Bay were considered in 2004 and developed by a number of Russian Institutes with international participation. This consisted of site investigations, surveys and studies to understand the technical challenges. A principal agreement was reached that the SNF would be removed from the site altogether and transported to Russia's reprocessing facility at Mayak in the Urals. The analytical studies provided the information necessary to develop the construction plan for the site. Following design and regulatory processes, stakeholders endorsed the technical solution in April 2007. This detailed the processes, facilities and equipment required to safely remove the SNF and identified other site services and support facilities required on the site. Implementation of this strategy is now well underway with the facilities in various states of construction. Physical works have been performed to address the most urgent tasks including weather protection over one of the DSUs, installation of shielding over the cells, provision of radiation

  8. Characteristics and fabrication of cermet spent nuclear fuel casks: ceramic particles embedded in steel

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Swaney, P.M.; Tiegs, T.N. [Oak Ridge National Lab., Oak Ridge, TN (United States)

    2004-07-01

    Cermets are being investigated as an advanced material of construction for casks that can be used for storage, transport, or disposal of spent nuclear fuel (SNF). Cermets, which consist of ceramic particles embedded in steel, are a method to incorporate brittle ceramics with highly desirable properties into a strong ductile metal matrix with a high thermal conductivity, thus combining the best properties of both materials. Traditional applications of cermets include tank armor, vault armor, drill bits, and nuclear test-reactor fuel. Cermets with different ceramics (DUO{sub 2}, Al{sub 2}O{sub 3}, Gd{sub 2}O{sub 3}, etc.) are being investigated for the manufacture of SNF casks. Cermet casks offer four potential benefits: greater capacity (more SNF assemblies) for the same gross weight cask, greater capacity (more SNF assemblies) for the same external dimensions, improved resistance to assault, and superior repository performance. These benefits are achieved by varying the composition, volume fraction, and particulate size of the ceramic particles in the cermet with position in the cask body. Addition of depleted uranium dioxide (DUO{sub 2}) to the cermet increases shielding density, improves shielding effectiveness, and increases cask capacity for a given cask weight or size. Addition of low-density aluminium oxide (Al{sub 2}O{sub 3}) to the outer top and bottom sections of the cermet cask, where the radiation levels are lower, can lower cask weight without compromising shielding. The use of Al2O3 and other oxides, in appropriate locations, can increase resistance to assault. Repository performance may be improved by compositional control of the cask body to (1) create a local geochemical environment that slows the long-term degradation of the SNF and (2) enables the use of DUO{sub 2} for longterm criticality control. While the benefits of using cermets follow directly from their known properties, the primary challenge is to develop low-cost methods to fabricate

  9. Hazard Evaluation for Storage of Spent Nuclear Fuel (SNF) Sludge at the Solid Waste Treatment Facility

    Energy Technology Data Exchange (ETDEWEB)

    SCHULTZ, M.V.

    2000-08-22

    As part of the Spent Nuclear Fuel (SNF) storage basin clean-up project, sludge that has accumulated in the K Basins due to corrosion of damaged irradiated N Reactor will be loaded into containers and placed in interim storage. The Hanford Site Treatment Complex (T Plant) has been identified as the location where the sludge will be stored until final disposition of the material occurs. Long term storage of sludge from the K Basin fuel storage facilities requires identification and analysis of potential accidents involving sludge storage in T Plant. This report is prepared as the initial step in the safety assurance process described in DOE Order 5480.23, Nuclear Safety Analysis Reports and HNF-PRO-704, Hazards and Accident Analysis Process. This report documents the evaluation of potential hazards and off-normal events associated with sludge storage activities. This information will be used in subsequent safety analyses, design, and operations procedure development to ensure safe storage. The hazards evaluation for the storage of SNF sludge in T-Plant used the Hazards and Operability Analysis (HazOp) method. The hazard evaluation identified 42 potential hazardous conditions. No hazardous conditions involving hazardous/toxic chemical concerns were identified. Of the 42 items identified in the HazOp study, eight were determined to have potential for onsite worker consequences. No items with potential offsite consequences were identified in the HazOp study. Hazardous conditions with potential onsite worker or offsite consequences are candidates for quantitative consequence analysis. The hazardous conditions with potential onsite worker consequences were grouped into two event categories, Container failure due to overpressure - internal to T Plant, and Spill of multiple containers. The two event categories will be developed into accident scenarios that will be quantitatively analyzed to determine release consequences. A third category, Container failure due to

  10. National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR {section}261.4(a)(4), ``Identification and Listing of Hazardous Waste,`` as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Tsutomu; Takahashi, Akira; Ishikawa, Niroh [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

    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.

  12. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Process report

    Energy Technology Data Exchange (ETDEWEB)

    Gribi, Peter; Johnson, Lawrence; Suter, Daniel; Smith, Paul; Pastina, Barbara; Snellman, Margit

    2008-01-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is placed in a perforated steel cylinder prior to emplacement; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the main processes potentially affecting the long-term safety of the system, covering radiation-related, thermal, hydraulic, mechanical, chemical (including microbiological) and radionuclide transport-related processes. The process descriptions deal sequentially with the main sub-systems: fuel/cavity in canister, cast iron insert and copper canister, buffer and other bentonite components, supercontainer

  13. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Process report

    Energy Technology Data Exchange (ETDEWEB)

    Gribi, Peter; Johnson, Lawrence; Suter, Daniel; Smith, Paul; Pastina, Barbara; Snellman, Margit

    2008-01-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is placed in a perforated steel cylinder prior to emplacement; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the main processes potentially affecting the long-term safety of the system, covering radiation-related, thermal, hydraulic, mechanical, chemical (including microbiological) and radionuclide transport-related processes. The process descriptions deal sequentially with the main sub-systems: fuel/cavity in canister, cast iron insert and copper canister, buffer and other bentonite components, supercontainer

  14. INTRODUCTION TO NON-DESTRUCTIVE TESTING METHODS COMMONLY USED FOR PWR NUCLEAR POWER PLANT IN-SERVICE INSPECTION%压水堆核电站在役检查常用无损检测方法简介

    Institute of Scientific and Technical Information of China (English)

    彭志珍; 李玉龙; 尹芹

    2012-01-01

    This paper briefly introduce the definition, scope, type, etc. of non--destructive testing methods commonly used for PWR nuclear power plant in-service inspection, and list some applications of each non--destructive testing method.%简要介绍了压水堆核电站在役检查常用无损检测方法的定义、适用范围、类型等,并列举了每种检测方法在核电站在役检查中的一些应用。

  15. Aplicación de la teoría de perturbación método diferencial al estudio de sensibilidad en generadores de vapor de centrales nucleares tipo PWR

    OpenAIRE

    Giol, Roberto

    1989-01-01

    En este trabajo se presenta una ampliación del formalismo diferencial de la teoría de perturbación a un modelo homogéneo de simulación del comportamiento estacionario de generadores de vapor de centrales nucleares PWR. El programa PERGEVAP, desarrollado a partir del modelo del código GEVAP de Souza, permite realizar cálculos de la sensibilidad de funciones lineales (como la temperatura media del primario) y no-lineales (como el flujo de calor medio) a variaciones en los parámetros termohidráu...

  16. Temporary storage in dry of the spent nuclear fuel in the Nuclear Power Plant of Laguna Verde; Almacenamiento temporal en seco del combustible nuclear gastado en la Central Nuclear Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, N.; Vargas A, A., E-mail: natividad.hernandez@cfe.gob.mx [Comision Federal de Electricidad, Gerencia de Centrales Nucleoelectricas, Carretera Veracruz-Medellin Km. 7.5, 94270 Dos Bocas, Veracruz (Mexico)

    2013-10-15

    To guarantee the continuity in the operation of the two nuclear reactors of the nuclear power plant of Laguna Verde (NPP-L V) is an activity of high priority of the Comision Federal de Electricidad (CFE) in Mexico. At the present time, the CFE is working in the storage project in dry of the spent fuel with the purpose of to liberate space of the pools and to have the enlarged capacity of storage of the spent fuel that is discharged of the reactors. This work presents the storage option in dry of the spent fuel, considering that the original capacity of the spent fuel pools of the NPP-L V was of 1242 spaces each one and that in 1991, through a modification of the original design, the storage capacity was increased to 3177 spaces by pool. At present, the cells occupied by unit are of 2165 (68%) for the Unit-I and 1839 (58%) for the Unit-2, however, in 2017 and 2022 the capacity to discharge the complete core will be limited by what is required of a retirement option of spent fuel assemblies to liberate spaces. (author)

  17. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y Consultants, Orchard Park, NY (US); Chen, S.Y.; Biwer, B.M.; LePoire, D.J. [Argonne National Lab., IL (US)

    1995-11-01

    This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows{trademark} environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident.

  18. Instant release fraction and matrix release of high burn-up UO2 spent nuclear fuel: Effect of high burn-up structure and leaching solution composition

    Science.gov (United States)

    Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.

    2012-08-01

    Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and

  19. Final disposal of spent nuclear fuel in Finnish bedrock - Kivetty site report

    Energy Technology Data Exchange (ETDEWEB)

    Anttila, P. [Fortum Engineering Oy, Vantaa (Finland); Ahokas, H.; Front, K. [Fintact Oy (Finland)] [and others

    1999-06-01

    Posiva Oy is studying the Finnish bedrock for the geological disposal of spent nuclear fuel. The study is based on the site selection research programme started originally in 1983. The programme is in accordance with the decision in principle by the Council of State in 1983 and aims at the selection of one site in 2000. Four sites, Haestholmen in Loviisa, Kivetty in Aeaenekoski, Olkiluoto in Eurajoki and Romuvaara in Kuhmo, have been studied in detail. This report summarises the results of the site investigations carried out at Kivetty. The bedrock of Kivetty belongs to the large Svecofennian granitoid complex of central Finland, about 1880 million years in age. The most common rock type is porphyritic granodiorite, which is cut by younger medium-grained granodiorite and porphyritic or even-grained granite. Minor bodies of gabbro, older than the porphyritic granodiorite, are also present. The granitoids show evidence of two deformation phases. Altogether 29 bedrock 'structures' (R-structures) have been modelled at the investigation site, most of them representing steeply dipping fracture zones. The rock mass between the fracture zones represents what is termed 'intact rock', which is typically hard, unweathered and sparsely fractured. The R-structures are generally hydraulically more conductive than the intact rock and their mean transmissivity is 1.3-10{sup -6} m{sup 2}/s. The corresponding mean of the hydraulic conductivity values for the intact rock, measured using a 2 m packer interval is 4*10{sup -11} m{sup 2}/s, if a lognormal distribution for all measured values is assumed. A clear decrease in hydraulic conductivity with depth has been found for the intact rock, and there seems to be a parallel decrease in the transmissivity of structures. In addition, the hydraulically conductive fractures seem to be more frequent and their transmissivities higher in the uppermost 100 - 200 m of the bedrock than at greater depths. The groundwater of

  20. Final disposal of spent nuclear fuel in Finnish bedrock. Olkiluoto site report

    Energy Technology Data Exchange (ETDEWEB)

    Anttila, P. [Fortum Engineering Oy, Vantaa (Finland); Ahokas, H. [Fintact Oy, Helsinki (Finland); Front, K. [VTT Communication and Infrastructure, Espoo (Finland)] [and others

    1999-06-01

    Posiva Oy is studying the Finnish bedrock for the geological disposal of spent nuclear fuel. The study is based on the site selection research programme started originally in 1983. The programme is in accordance with the decision in principle by the Council of State in 1983 and aims at the selection of one site in 2000. Four sites, Haestholmen in Loviisa, Kivetty in Aeaenekoski, Olkiluoto in Eurajoki and Romuvaara in Kuhmo, have been studied in detail. This report summarises the results of the site investigations carried out at Olkiluoto. The bedrock of the Olkiluoto site consists of Svecofennian metasediments and platonic rocks, 1800-1900 million years in age. Migmatitic mica gneiss is the most abundant rock type, and is intruded by foliated tonalites and granodiorites and massive coarse-grained granites and pegmatites. Five successive plastic deformation phases have been defined. In total, 30 bedrock structures (R-structures) have been modelled at the site. Most of these represent steeply dipping fracture zones, but several sub-horizontal zones, gently dipping to the SE, have also been identified. The rock mass between the fracture zones represents what is termed `intact rock`, which is typically hard, unweathered and sparsely fractured. The R-structures are generally hydraulically more conductive than the intact rock and their mean transmissivity is 3 x 10{sup -7} m{sup 2}/s. The corresponding mean of the hydraulic conductivity values for the intact rock measured using a 2 m packer interval, is 8 x 10{sup -13} m/s, if a lognormal distribution for all measured values is assumed. A clear decrease in hydraulic conductivity with depth has been found for the intact rock, and there seems to be a parallel decrease in the transmissivity of structures. In addition, the hydraulically conductive fractures seem to be more frequent and their transmissivities higher in the uppermost 100 - 200 m of the bedrock than at greater depths. The groundwater chemistry reflects the

  1. Spent fuel, plutonium and nuclear waste: long-term management; Le combustible use et le plutonium en tant que dechets nucleaires: gestion a long terme

    Energy Technology Data Exchange (ETDEWEB)

    Collard, G

    1998-11-01

    Different options for the management of nuclear waste arising from the nuclear fuel cycle are discussed. Special emphasis is on reprocessing followed by geological disposal, geological disposal of reprocessing waste, direct geological disposal of spent nuclear fuel, long term storage. Particular emphasis is on the management of plutonium including recycling, immobilisation and disposal, partitioning and transmutation.

  2. FY15 Status Report: CIRFT Testing of Spent Nuclear Fuel Rods from Boiler Water Reactor Limerick

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wang, Hong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-01

    The objective of this project is to perform a systematic study of used nuclear fuel (UNF, also known as spent nuclear fuel [SNF]) 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. The additional CIRFT was conducted on three HBR rods (R3, R4, and R5) in which two specimens failed and one specimen was tested to over 2.23 10⁷ cycles without failing. The data analysis on all the HBR UNF rods demonstrated that it is necessary to characterize the fatigue life of the UNF 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 of 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, ten SNF rod segments from BWR Limerick were tested using ORNL CIRFT, with one under static and nine dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at maximum curvature 4.0 m⁻¹. The specimen did not show any sign of failure in three repeated loading cycles to almost same maximum curvature. Ten cyclic tests were conducted with amplitude varying from 15.2 to 7.1 N·m. Failure was observed in nine of the tested rod specimens. The cycles to failure were

  3. Licensing procedures for a dedicated ship for carrying spent nuclear fuel and radioactive waste. Report from workshop held at GOSAOMNADZOR, Moscow 2 -3 July 2001

    Energy Technology Data Exchange (ETDEWEB)

    Sneve, Margorzata K.; Bergman, Curt; Markarov, Valentin

    2001-07-01

    The report describes information exchange and discussion about the licensing principles and procedures for spent nuclear fuel and radioactive waste transportation at sea. Russian health, environment and safety requirements for transportation of waste by ships. (Author)

  4. Procedures for using expert judgment to estimate human-error probabilities in nuclear power plant operations. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Seaver, D.A.; Stillwell, W.G.

    1983-03-01

    This report describes and evaluates several procedures for using expert judgment to estimate human-error probabilities (HEPs) in nuclear power plant operations. These HEPs are currently needed for several purposes, particularly for probabilistic risk assessments. Data do not exist for estimating these HEPs, so expert judgment can provide these estimates in a timely manner. Five judgmental procedures are described here: paired comparisons, ranking and rating, direct numerical estimation, indirect numerical estimation and multiattribute utility measurement. These procedures are evaluated in terms of several criteria: quality of judgments, difficulty of data collection, empirical support, acceptability, theoretical justification, and data processing. Situational constraints such as the number of experts available, the number of HEPs to be estimated, the time available, the location of the experts, and the resources available are discussed in regard to their implications for selecting a procedure for use.

  5. Experience with the transport and storage casks CASTOR (registered) MTR 2 for spent nuclear fuel assemblies from research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jack, Allen; Rettenbacher, Katharina; Skrzyppek, Juergen [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2011-07-01

    The CASTOR (registered) MTR 2 cask was designed and manufactured by the company GNS during the 1990's for the transport and interim storage of spent nuclear fuel assemblies from various types of research reactors. Casks of this type have been used at the VKTA Research Centre in Rossendorf near Dresden, Germany as well as at the European Commission's Joint Research Centre at Petten and at the HOR reactor at Delft in the Netherlands. A total of 24 units have been used for the functions of transport and storage with various spent fuel types (VVER, HFR-HEU, and HOR-HEU) for more than ten years now. This type of packaging for radioactive material is a member of the CASTOR (registered) family of spent nuclear fuel casks used worldwide. Over 1000 units are loaded and in storage in Europe, Asia, Africa and North America. This paper presents the experience from the use of the casks for transport and storage in the past, as well as the prospects for the future. (author)

  6. Validation study for crediting chlorine in criticality analyses for spent nuclear fuel disposition

    Energy Technology Data Exchange (ETDEWEB)

    Sobes, Vladimir [ORNL; Scaglione, John M [ORNL; Wagner, John C [ORNL; Dunn, Michael E [ORNL

    2015-01-01

    Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10-5 keff). This study investigated the design of a series of

  7. Validation Study for Crediting Chlorine in Criticality Analyses for US Spent Nuclear Fuel Disposition

    Energy Technology Data Exchange (ETDEWEB)

    Sobes, Vladimir [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wagner, John C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dunn, Michael E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10-5 keff). This study investigated the design of a series of

  8. Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual

    Energy Technology Data Exchange (ETDEWEB)

    IRWIN, J.J.

    2000-11-18

    The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed

  9. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Paul; Johnson, Lawrence; Snellman, Margit; Pastina, Barbara; Gribi, Peter

    2008-01-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the

  10. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Paul; Johnson, Lawrence; Snellman, Margit; Pastina, Barbara; Gribi, Peter

    2008-01-15

    The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the

  11. Thermal characteristics of spent activated carbon generated from air cleaning units in korean nuclear power plants

    Directory of Open Access Journals (Sweden)

    Ji-Yang So

    2017-06-01

    Full Text Available To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

  12. Thermal characteristics of spent activated carbon generated from air cleaning units in Korean nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hang Rae; So, Ji Yang [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    To identify the feasibility of disposing of spent activated carbon as a clearance level waste, we performed characterization of radioactive pollution for spent activated carbon through radioisotope analysis; results showed that the C-14 concentrations of about half of the spent activated carbon samples taken from Korean NPPs exceeded the clearance level limit. In this situation, we selected thermal treatment technology to remove C-14 and analyzed the moisture content and thermal characteristics. The results of the moisture content analysis showed that the moisture content of the spent activated carbon is in the range of 1.2–23.9 wt% depending on the operation and storage conditions. The results of TGA indicated that most of the spent activated carbon lost weight in 3 temperature ranges. Through py-GC/MS analysis based on the result of TGA, we found that activated carbon loses weight rapidly with moisture desorption reaching to 100°C and desorbs various organic and inorganic carbon compounds reaching to 200°C. The result of pyrolysis analysis showed that the experiment of C-14 desorption using thermal treatment technology requires at least 3 steps of heat treatment, including a heat treatment at high temperature over 850°C, in order to reduce the C-14 concentration below the clearance level.

  13. Standard guide for pyrophoricity/combustibility testing in support of pyrophoricity analyses of metallic uranium spent nuclear fuel

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (SNF). The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI). 1.2 This guide describes testing of metallic uranium and metallic uranium-based SNF in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60/63). The testing described ...

  14. Deep repository for spent nuclear fuel. Facility description - Layout E. Spiral ramp with one operational area

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Stig [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Forsgren, Ebbe [SwedPower AB, Stockholm (Sweden); Lange, Fritz [Lange Art AB, Stockholm (Sweden)

    2002-04-01

    This report documents a proposal for the design of the deep repository for spent nuclear fuel. The proposal is based on the principles that were formulated in the original KBS-3 study, but has been supplemented by investigations and experience to reflect current knowledge. The purpose of the report is to provide an integrated picture of the deep repository, as a basis for SKB's other work, e.g. environmental impact assessments, transport systems, safety issues and alternative locations, and to provide a co-ordinated account of the conditions and requirements concerning all of the necessary functions in the deep repository in order to have a well functioning facility. In addition, it should be possible to use the report as: a tool in the task of achieving a co-ordinated, safe and accepted design for the facility, a basis for further planning and costing, a basis for adaptation to geographic and other conditions for the particular location, a basis for information material, both within SKB and for interested parties outside, such as public authorities, municipalities and the general public. The capacity of the deep repository has been chosen on the basis of 40 years of operation of the Swedish nuclear power reactors, which will produce approximately 9,000 tons of uranium, equivalent to approximately 4,500 canisters. The design outlined is based on theoretical analyses of functions, safety requirements, procedures etc. that can be identified during the various phases of the construction and operation of the repository. In addition, preliminary organisation and staffing plans have been drawn up, for use as the basis for planning the necessary buildings. The report gives a vision of the overall layout and function of the facility, and a proposal for the design of all individual parts of the repository. The relationships between the various parts of the repository are described, both above and below ground, as is the interplay between the part above ground and part

  15. Responsibility for safe management of spent nuclear fuel - a legal perspective; Ansvarstagande i kaernbraenslecykelns slutsteg - ett raettsligt perspektiv

    Energy Technology Data Exchange (ETDEWEB)

    Cramer, Per; Stendahl, Sara; Erhag, Thomas (Dept. of Law, Univ. of Goeteborg, Goeteborg (Sweden))

    2010-10-15

    This study analyzes, based on Section 10 of the Swedish Nuclear Activities Act, the legal structures surrounding the issue of responsibility for safe management and final disposal of spent nuclear fuel. The purpose is to shed light on the legal aspects that must be considered in the future licensing process and thereby contribute to a better understanding of the importance of the legal structures for the decisions about final disposal that lie ahead of us. The overall question is thus future-oriented: What interpretation is it reasonable to assume will be given to the requirements of the Nuclear Activities Act on 'safe management and final disposal' of the spent nuclear fuel in the coming licensing process? The approach we take to this question is in part traditionally jurisprudential and based on a study of the travaux preparatoires (drafting history) of the Act and other legal sources. In addition, a study of legal practice is included where previous licensing processes are studied. One conclusion that can be drawn from this study is that the Swedish regulation of nuclear activities creates a legal basis for exacting far-reaching industrial responsibility from the reactor owners, but also for an extensive and interventionist state influence over the activities. Of central importance in the model for division of responsibility that was established via the Nuclear Activities Act in 1984 is the RDandD programme (Research, Development and Demonstration). The RDandD programme reflects the political will that the requirement of 'safe management' should be met through research. The statutory forms for how the programme is to be organized reflect an ambition to place great responsibility for execution and financing on the industry, but also, and not least, an ambition to retain instruments of control and influence in the hands of the state. It is difficult to judge whether the hopes of the 1980s regarding the influence of the public over this process

  16. Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10

    Energy Technology Data Exchange (ETDEWEB)

    1994-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

  17. A National Tracking Center for Monitoring Shipments of HEU, MOX, and Spent Nuclear Fuel: How do we implement?

    Energy Technology Data Exchange (ETDEWEB)

    Mark Schanfein

    2009-07-01

    Nuclear material safeguards specialists and instrument developers at US Department of Energy (USDOE) National Laboratories in the United States, sponsored by the National Nuclear Security Administration (NNSA) Office of NA-24, have been developing devices to monitor shipments of UF6 cylinders and other radioactive materials , . Tracking devices are being developed that are capable of monitoring shipments of valuable radioactive materials in real time, using the Global Positioning System (GPS). We envision that such devices will be extremely useful, if not essential, for monitoring the shipment of these important cargoes of nuclear material, including highly-enriched uranium (HEU), mixed plutonium/uranium oxide (MOX), spent nuclear fuel, and, potentially, other large radioactive sources. To ensure nuclear material security and safeguards, it is extremely important to track these materials because they contain so-called “direct-use material” which is material that if diverted and processed could potentially be used to develop clandestine nuclear weapons . Large sources could be used for a dirty bomb also known as a radioactive dispersal device (RDD). For that matter, any interdiction by an adversary regardless of intent demands a rapid response. To make the fullest use of such tracking devices, we propose a National Tracking Center. This paper describes what the attributes of such a center would be and how it could ultimately be the prototype for an International Tracking Center, possibly to be based in Vienna, at the International Atomic Energy Agency (IAEA).

  18. Safety case for the disposal of spent nuclear fuel at Olkiluoto. Description of the disposal system 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-12-15

    Description of the Disposal System sits within Posiva Oy's Safety Case 'TURVA-2012' report portfolio and has the objective presenting the initial state of the disposal system for the safety case for the disposal of spent nuclear fuel at Olkiluoto, Finland. Disposal system is an entity composed of a repository system and surface environment. The repository system includes the spent nuclear fuel, canister, buffer, backfill, and closure components as well as the host rock. The repository system components have assigned safety functions (except for the spent nuclear fuel) and are subject to requirements. The initial state is presented for each component, and references to the main supporting reports are given to guide the reader for more details. Conditions for each component vary in time and space, due to the time of emplacement and due to the tolerances set for the compositions, geometries and other properties depending on the component. The disposal operation is foreseen to commence {approx} 2020. At the beginning of the postclosure period, around 2120, all the engineered components have been installed and the operation is finalised. The system evolution during the operational phase is discussed in detail in Performance Assessment. The initial state for the host rock is defined to be essentially equal to the baseline conditions prior to starting the construction of the underground characterisation facility ONKALO. For the surface environment, the initial state is the present conditions prevailing. For any other component of the disposal system, the initial state is defined as the state it has when the direct control over that specific part of the system ceases and only limited information can be made available on the subsequent development of conditions in that part of the system or its near field. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-12

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

  20. Investigation on the use digital controls instead of PID analog controls in the level control of steam generators of nuclear power PWR; Investigacao sobre o uso de controladores digitais em substituicao aos controladores analogicos PID para o controle de nivel de geradores de vapor de centrais nucleares PWR

    Energy Technology Data Exchange (ETDEWEB)

    Alvarenga, Marco Antonio Bayout

    2012-07-01

    The aim of this study is to identify current alternatives for the implementation of digital controllers in the level control of steam generators of nuclear power PWR (Pressurized Water Reaetor). It is intended to identify the types of digital controls that are available from the theoretical and conceptual viewpoints for this purpose. We investigate the advantages and disadvantages of each controller model. From this assessment are pointed the most suitable models in hierarchical scale. This evaluation also serves to suggest possible types of control installation as a whole, where the level control of the steam generators becomes just one of many controls that are part of the plant. In this case, the use of digital controls allows the non-linear and multivariable treatment which is characteristic of complex systems, such as the nuclear power generation. The treatment of nonlinearities and multivariable aspects allows a more detailed study of the stability of these plants when they are subject to transients or several accidents, such as the case of losing external power of transients. In the specific case of steam generators, the instabilities result from the emergence of the shrink and swell phenomenas, depending on the load variations of thermonuclear plant. The application of several types and digital controllers, considering these inherent characteristics of the level control of steam generators, allows to infer which types of controllers are more appropriate to treat instabilities of this type and to make conjectures in its use for the cases of more complex instabilities, considering the integration of all nucleus-plant controls.

  1. Dynamic leaching studies of 48 MWd/kgU UO2 commercial spent nuclear fuel under oxic conditions

    OpenAIRE

    Serrano Purroy, D.; Casas Pons, Ignasi; Gonzalez Robles, E.; Glatz, Jean Paul; Wegen, D.H.; Clarens Blanco, Frederic; Giménez Izquierdo, Francisco Javier; Pablo Ribas, Joan de; Martínez Esparza, A.

    2013-01-01

    The leaching of a high-burn-up spent nuclear fuel (48 MWd/KgU) has been studied in a carbonate-containing solution and under oxic conditions using a Continuously Stirred Tank Flow-Through Reactor (CSTR). Two samples of the fuel, one prepared from the centre of the pellet (labelled CORE) and another one from the fuel pellet periphery, enriched with the so-called High Burn-Up Structure (HBS, labelled OUT) have been used. For uranium and actinides, the results showed that U, Np, Am and Cm ga...

  2. Review study 1995. Localization of the repository for spent nuclear fuel; Oeversiktsstudie 95. Lokalisering av djupfoervar foer anvaent kaernbraensle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-01

    This report gives an overview of the studies performed by SKB pertinent to selection of a site for the Swedish repository for spent nuclear fuels, and is written for both experts in the various fields involved, decision-makers and the interested general public. The review can not comprise all detailed factors necessary for deciding the localization, but deals mainly with conditions on the land surface and can point out areas which are not well suited or less interesting as a site. It also treats several scientific, technical and social bases in different parts of the country. 120 refs, 53 figs.

  3. Final disposal of spent nuclear fuel in Finnish bedrock. Haestholmen site report

    Energy Technology Data Exchange (ETDEWEB)

    Anttila, P. [Fortum Engineering Oy, Vantaa (Finland); Ahokas, H. [Fintact Oy, Helsinki (Finland); Front, K. [VTT Communities and Infrastructure, Espoo (Finland)

    1999-06-01

    Posiva Oy is studying the Finnish bedrock for the geological disposal of spent nuclear fuel. The study is based on the site selection research programme started originally in 1983. The programme is in accordance with the decision in principle by the Council of State in 1983 and aims at the selection of one site in 2000. Four sites, Haestholmen in Loviisa, Kivetty in Aeaenekoski, Olkiluoto in Eurajoki and Romuvaara in Kuhmo, have been studied in detail. This report summarises the results of the site investigations carried out at Haestholmen. The Haestholmen area is located within the anorogenic Wiborg rapakivi granite batholith, about 1630 million years in age, representing one of the youngest rock formations in Finland. Wiborgite, pyterlite, porphyritic rapakivi granite and even-grained rapakivi granite are the rock types present. 25 bedrock structures have been modelled at the site. Most of them are steeply-dipping fracture zones trending NW-SE and NE-SW, but several sub-horizontal zones, mainly dipping to the N-NE and the SW, are also present. The rock mass between the fracture zones represents what is termed `intact rock`, which is typically hard, unweathered and sparsely fractured. The bedrock structures are generally hydraulically more conductive than the intact rock and their mean transmissivity is 8 x 10{sup -6} m{sup 2}/s or 1.3 x 10{sup -6} m{sup 2}/s, depending on how structures are defined. The corresponding mean of the hydraulic conductivity values measured for the intact rock using a 2 m packer interval is 1 x 10{sup -12} m/s, if a lognormal distribution for all measured values is assumed. A clear decrease in hydraulic conductivity with depth has been found in the intact rock. In addition, the hydraulically conductive fractures seem to be more frequent and their transmissivities higher in the uppermost 100-200 m of the bedrock than at greater depths. The groundwater chemistry reflects the post-glacial history of the island of Haestholmen, which rose

  4. Impact of long-term climate change on a deep geological repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Boulton, G.S. [Edinburgh Univ. (United Kingdom); Kautsky, U.; Moren, L. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Wallroth, T. [Bergab Consulting Geologists (Sweden)

    2001-05-01

    The radioactivity of spent nuclear fuel will decay over a period of time (100,000 years or longer) in which we expect major environmental change. Climatically driven changes such as glaciation, permafrost and changes in sea level will affect the subsurface environment and must therefore be considered in performance and safety assessments. We regard the state of climate to be determined by the climate system, comprising the atmosphere, the biosphere, the oceans, the ice sheets and the surface of the lithosphere. Climate can change as a consequence of external forcing or as a consequence of the internal dynamics. The changes in the past show a complex cyclical pattern with repetitive periodicities and magnitudes of change. On the relatively long time scale which is of concern in this report climate changes are believed to be triggered by changes in insolation due to cyclical changes in the earth's orbit around the sun. This is referred to as the Milankovitch theory or the astronomical climate theory. A brief review of our knowledge of the climate system, proxy records and observations of past climate change is presented in the report. The most extreme departures from modern environmental conditions in Sweden have occurred during the cold, glacial cycles with ice sheets covering the whole of Sweden. As part of SKB's palaeohydrogeological research programme a time-dependent, thermo-mechanically coupled model of ice sheet behaviour has been developed in order to simulate past fluctuations of the Scandinavian ice sheet and forecast its future. The ice sheet model can calculate the temperature field and isostatic response of the underlying bedrock, subglacial and proglacial permafrost and the subglacial melt rates. By connecting the glaciation model with hydrogeological and rock mechanical models, the response of the subsurface to climate change can be investigated. In this report the climate-driven environmental changes are represented as a series of successive

  5. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository--Volume 1: Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, L.L.; Wilson, J.R.; Sanchez, L.Z.; Aguilar, R.; Trellue, H.R.; Cochrane, K.; Rath, J.S.

    1998-10-01

    The US Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

  6. Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository--Volume 2: Methodology and Results

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, L.L.; Wilson, J.R.; Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K.; Rath, J.S.

    1998-10-01

    The US Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

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

  8. Decay heat power of spent nuclear fuel of power reactors with high burnup at long-term storage

    Science.gov (United States)

    Ternovykh, Mikhail; Tikhomirov, Georgy; Saldikov, Ivan; Gerasimov, Alexander

    2017-09-01

    Decay heat power of actinides and fission products from spent nuclear fuel of power VVER-1000 type reactors at long-term storage is calculated. Two modes of storage are considered: mode in which single portion of actinides or fission products is loaded in storage facility, and mode in which actinides or fission products from spent fuel of one VVER reactor are added every year in storage facility during 30 years and then accumulated nuclides are stored without addition new nuclides. Two values of fuel burnup 40 and 70 MW·d/kg are considered for the mode of storage of single fuel unloading. For the mode of accumulation of spent fuel with subsequent storage, one value of burnup of 70 MW·d/kg is considered. Very long time of storage 105 years accepted in calculations allows to simulate final geological disposal of radioactive wastes. Heat power of fission products decreases quickly after 50-100 years of storage. The power of actinides decreases very slow. In passing from 40 to 70 MW·d/kg, power of actinides increases due to accumulation of higher fraction of 244Cm. These data are important in the back end of fuel cycle when improved cooling system of the storage facility will be required along with stronger radiation protection during storage, transportation and processing.

  9. Public acceptability of the use of gamma rays from spent nuclear fuel as a hazardous waste treatment process

    Energy Technology Data Exchange (ETDEWEB)

    Mincher, B.J.; Wells, R.P.; Reilly, H.J.

    1992-01-01

    Three methods were used to estimate public reaction to the use of gamma irradiation of hazardous wastes as a hazardous waste treatment process. The gamma source of interest is spent nuclear fuel. The first method is Benefit-Risk Decision Making, where the benefits of the proposed technology are compared to its risks. The second analysis compares the proposed technology to the other, currently used nuclear technologies and estimates public reaction based on that comparison. The third analysis is called Analysis of Public Consent, and is based on the professional methods of the Institute for Participatory Management and Planning. The conclusion of all three methods is that the proposed technology should not result in negative public reaction sufficient to prevent implementation.

  10. Spent nuclear fuel. A review of properties of possible relevance to corrosion processes

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, R. [Caledon Consult AB, Nykoeping (Sweden)

    1995-04-01

    The report reviews the properties of spent fuel which are considered to be of most importance in determining the corrosion behaviour in groundwaters. Pellet cracking and fragment size distribution are discussed, together with the available results of specific surface area measurements on spent fuel. With respect to the importance of fuel microstructure, emphasis is placed on recent work on the so called structural rim effect, which consists of the formation of a zone of high porosity, and the polygonization of fuel grains to form many sub-grains, at the pellet rim, and appears to be initiated when the average pellet burnup exceeds a threshold of about 40 MWd/kgU. Due to neutron spectrum effects, the pellet rim is also associated with the buildup of plutonium and other actinides, which results in an enhanced local burnup and specific activity of both beta-gamma and alpha radiation, thus representing a greater potential for radiolysis effects in ingressed groundwater. The report presents and discusses the results of quantitative determination of the radial profiles of burnup and alpha activity on spent fuel with average burnups from 21.2 to 49 MWd/kgU. In addition to the radial variation of fission product and actinide inventories due to the effects mentioned above, migration, redistribution and release of some fission products can occur during reactor irradiation and the report concludes with a short review of these processes.

  11. The feasibility of modelling coupled processes in safety analysis of spent nuclear fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Rasilainen, K. [VTT Energy, Espoo (Finland); Luukkonen, A.; Niemi, A.; Poellae, J. [VTT Communities and Infrastructure, Espoo (Finland); Olin, M. [VTT Chemical Technology, Espoo (Finland)

    1999-07-01

    The potential of applying coupled modelling in the Finnish safety analysis programme has been reviewed. The study focused on the migration of radionuclides escaping from a spent fuel repository planned to be excavated in fractured bedrock. Two effects that can trigger various couplings in and around a spent fuel repository in Finland were studied in detail; namely heat generation in the spent fuel and the presence of deep, saline groundwaters. The latter have been observed in coastal areas. A systematic survey of the requirements of coupled modelling identified features that render such migration calculations a challenging task. In groundwater flow modelling there appears to be wide ranging uncertainty related to conceptualisation of flow systems and to the corresponding input data. In terms of migration related chemistry there appear to be large gaps in the underlying thermodynamic database for geochemical systems. Rock mechanical predictions are heavily dependent on knowing the location, structure and properties of dominant fractures; information which is extremely difficult to obtain. Conduction and convection of heat is understood well in principle. On the basis of this review, it appears that coupled migration modelling may not yet be at the stage of development that would allow its use as a standard modelling tool in performance assessments. However, a firmer basis for the conclusions reached can only be obtained after a systematic modelling exercise on a relevant and real migration problem has been carried out. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Scott, Mark Robert [Texas A & M Univ., College Station, TX (United States)

    2005-05-01

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

  13. CANDU堆应用RU的PWR/CANDU联合核燃料循环的研究%Study of RU Utilization in CANDU Reactor-an Advanced Nuclear Fuel Cycle of PWR/CANDU Synergism

    Institute of Scientific and Technical Information of China (English)

    霍小东; 谢仲生

    2003-01-01

    对压水堆乏燃料后处理回收铀(RU)在秦山三期CANDU堆中应用的可行性和经济性进行分析.使用ORIGEN2程序,对后处理回收铀在生产后放置不同时间后核素的成份和放射性活度进行了计算.证明RU燃料元件生产的放射性水平是可以接受的.使用DRAGON/DONJON程序对应用RU的秦山三期CANDU堆的时均堆芯和瞬时堆芯校验分析表明:采用简单的2燃耗区,2、4棒束的换料方案能满足最大通道功率、最大棒束功率限制.通过放射性分析和堆芯物理分析可以看出,秦山三期CANDU堆在不改变堆芯结构及运行模式的条件下,从天然铀(NU)燃料过渡到RU燃料是可行的.通过对秦山三期CANDU堆应用RU的经济性分析,可以看出PWR/CANDU联合核燃料循环的策略既可节约铀资源(23%),提高燃料的能量输出(41%),又减少了废燃料的处置量(66%),可大大降低核电成本.

  14. 105-K Basin Material Design Basis Feed Description for Spent Nuclear Fuel (SNF) Project Facilities VOL 1 Fuel

    Energy Technology Data Exchange (ETDEWEB)

    PACKER, M.J.

    1999-11-04

    Metallic uranium Spent Nuclear Fuel (SNF) is currently stored within two water filled pools, 105-KE Basin (KE Basin) and 105-KW Basin (KW Basin), at the United States Department of Energy (U.S. DOE) Hanford Site, in southeastern Washington State. The Spent Nuclear Fuel Project (SNF Project) is responsible to DOE for operation of these fuel storage pools and for the 2100 metric tons of SNF materials that they contain. The SNF Project mission includes safe removal and transportation of all SNF from these storage basins to a new storage facility in the 200 East Area. To accomplish this mission, the SNF Project modifies the existing KE Basin and KW Basin facilities and constructs two new facilities: the 100 K Area Cold Vacuum Drying Facility (CVDF), which drains and dries the SNF; and the 200 East Area Canister Storage Building (CSB), which stores the SNF. The purpose of this document is to describe the design basis feed compositions for materials stored or processed by SNF Project facilities and activities. This document is not intended to replace the Hanford Spent Fuel Inventory Baseline (WHC 1994b), but only to supplement it by providing more detail on the chemical and radiological inventories in the fuel (this volume) and sludge. A variety of feed definitions is required to support evaluation of specific facility and process considerations during the development of these new facilities. Six separate feed types have been identified for development of new storage or processing facilities. The approach for using each feed during design evaluations is to calculate the proposed facility flowsheet assuming each feed. The process flowsheet would then provide a basis for material compositions and quantities which are used in follow-on calculations.

  15. Standard guide for evaluation of materials used in extended service of interim spent nuclear fuel dry storage systems

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of t...

  16. Gamma spectrometric characterization of short cooling time nuclear spent fuels using hemispheric CdZnTe detectors

    CERN Document Server

    Lebrun, A; Szabó, J L; Arenas-Carrasco, J; Arlt, R; Dubreuil, A; Esmailpur-Kazerouni, K

    2000-01-01

    After years of cooling, nuclear spent fuel gamma emissions are mainly due to caesium isotopes which are emitters at 605, 662 and 796-801 keV. Extensive work has been done on such fuels using various CdTe or CdZnTe probes. When fuels have to be measured after short cooling time (during NPP outage) the spectrum is much more complex due to the important contributions of niobium and zirconium in the 700 keV range. For the first time in a nuclear power plant, four spent fuels of the Kozloduy VVER reactor no 4 were measured during outage, 37 days after shutdown of the reactor. In such conditions, good resolution is of particular interest, so a 20 mm sup 3 hemispheric crystal was used with a resolution better than 7 keV at 662 keV. This paper presents the experimental device and analyzes the results which show that CdZnTe commercially available detectors enabled us to perform a semi-quantitative determination of the burn-up after a short cooling time. In addition, it is discussed how a burn-up evolution code (CESAR)...

  17. Estimation of 85Kr dispersion from the spent nuclear fuel reprocessing plant in Rokkasho, Japan, using an atmospheric dispersion model.

    Science.gov (United States)

    Abe, K; Iyogi, T; Kawabata, H; Chiang, J H; Suwa, H; Hisamatsu, S

    2015-11-01

    The spent nuclear fuel reprocessing plant of Japan Nuclear Fuel Limited (JNFL) located in Rokkasho, Japan, discharged small amounts of (85)Kr into the atmosphere during final tests of the plant with actual spent fuel from 31 March 2006 to October 2008. During this period, the gamma-ray dose rates due to discharged (85)Kr were higher than the background rates measured at the Institute for Environmental Sciences and at seven monitoring stations of the Aomori prefectural government and JNFL. The dispersion of (85)Kr was simulated by means of the fifth-generation Penn State/NCAR Mesoscale Model and the CG-MATHEW/ADPIC models (ver. 5.0) with a vertical terrain-following height coordinate. Although the simulated gamma-ray dose rates due to discharged (85)Kr agreed fairly well with measured rates, the agreement between the estimated monthly mean (85)Kr concentrations and the observed concentrations was poor. Improvement of the vertical flow of air may lead to better estimation of (85)Kr dispersion.

  18. Basic Considerations for Dry Storage of Spent Nuclear Fuels and Revisited CFD Thermal Analysis on the Concrete Cask

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Jae Soo [ACT Co. Ltd., Daejeon (Korea, Republic of); Park, Younwon; Song, Sub Lee [BEES Inc., Daejeon (Korea, Republic of); Kim, Hyeun Min [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    The integrity of storage facility and also of the spent nuclear fuel itself is considered very important. Storage casks can be located in a designated area on a site or in a designated storage building. A number of different designs for dry storage have been developed and used in different countries. Dry storage system was classified into two categories by IAEA. One is container including cask and silo, the other one is vault. However, there is various way of categorization for dry storage system. Dry silo and cask are usually classified separately, so the dry storage system can be classified into three different types. Furthermore, dry cask storage can be categorized into two types based on the type of the materials, concrete cask and metal cask. In this paper, the design characteristics of dry storage cask are introduced and computational fluid dynamics (CFD) based thermal analysis for concrete cask is revisited. Basic principles for dry storage cask design were described. Based on that, thermal analysis of concrete dry cask was introduced from the study of H. M. Kim et al. From the CFD calculation, the temperature of concrete wall was maintained under the safety criteria. From this fundamental analysis, further investigations are expected. For example, thermal analysis on the metal cask, thermal analysis on horizontally laid spent nuclear fuel assemblies for transportation concerns, and investigations on better performance of natural air circulation in dry cask can be promising candidates.

  19. SR-CAN - a safety assessment of a repository of spent nuclear fuel: canister performance and effects on the biosphere

    Energy Technology Data Exchange (ETDEWEB)

    Kautsky, U.; Kumblad, L. [Swedish Nuclear Fuel and Waste Management Co. (SKB), Stockholm (Sweden)

    2004-07-01

    During the next few years the Swedish Nuclear Fuel and Waste Management Co. (SKB) performs site investigations at two sites in Sweden for a future repository of spent nuclear fuel. Parallel an encapsulation plant is planned to encapsulate the spent fuel in copper canisters according to the KBS-3 method. The purpose of the SR-CAN safety assessment is to show the performance of the canister isolations at different sites for a repository at 500 meters depth in crystalline rock. Moreover, SR-CAN provides an example how the site specific safety assessment of a deep repository will be made in year 2006-2008. To be able to calculate dose and risk for humans and the environment, new assessment methods were developed for the biosphere. These methods were based on a system ecological approach and used knowledge from landscape ecology to provide an integrated approach with hydrology and geology considering the discharges in a watershed and calculating consequences in terrestrial and aquatic (freshwater and marine) ecosystems. A range of methods and tools were developed in GIS and Matlab/Simulink to be able to model and understand the important processes in the landscape today and during the next few thousands of years. In this paper, an overview of the program and the novel methods are presented, as well as some examples from performance calculations from a watershed in the Forsmark area considering effects on humans and ecosystems. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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).

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

    Energy Technology Data Exchange (ETDEWEB)

    Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hardin, Ernest [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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).

  2. The PWR cores management; La gestion des coeurs REP

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Rippert, D. [CEA Cadarache, Departement d' Etudes des Reacteurs, DER, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    2000-01-25

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  3. An information management system for a spent nuclear fuel interim storage facility.

    Energy Technology Data Exchange (ETDEWEB)

    Finch, Robert J.; Chiu, Hsien-Lang (Taiwan Power Co., Taipei, 10016 Taiwan); Giles, Todd; Horak, Karl Emanuel; Jow, Hong-Nian (Jow International, Kirkland, WA)

    2010-12-01

    We describe an integrated information management system for an independent spent fuel dry-storage installation (ISFSI) that can provide for (1) secure and authenticated data collection, (2) data analysis, (3) dissemination of information to appropriate stakeholders via a secure network, and (4) increased public confidence and support of the facility licensing and operation through increased transparency. This information management system is part of a collaborative project between Sandia National Laboratories, Taiwan Power Co., and the Fuel Cycle Materials Administration of Taiwan's Atomic Energy Council, which is investigating how to implement this concept.

  4. Comparison of radiation hazard of HLW in several spent nuclear fuel reprocessing scenarios

    Directory of Open Access Journals (Sweden)

    Gladilov D.

    2012-10-01

    Full Text Available Radiation hazard of radionuclide has been calculated as a product of Aε where A is an activity of radionuclide and ε is a dose coefficient through ingestion. The values Aε of 18 radionuclide in spent fuel of WWER-440 are calculated. Because the full division of americium and curium from HLW is very complicated a separation americium from curium is considered. It is shown that a separation of americium in a special fraction allows decreasing the radiation hazard of HLW by 97.6% after 1000 years.

  5. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

    Energy Technology Data Exchange (ETDEWEB)

    Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

    2008-12-15

    The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments

  6. Dissolution of fluorite type surfaces as analogues of spent nuclear fuel : Production of suitable analogues and study the effect of surface orientation on dissolution

    OpenAIRE

    Godinho, Jose

    2011-01-01

    It is accepted worldwide that the best final solution for spent nuclear fuel is to bury it in deep geological repositories. Despite the physical and chemical barriers that are supposed to isolate the nuclear waste for at least 100.000 years, some uncertainty factors may cause underground water to get in contact with the nuclear waste. Due to radioactivity and oxidation under air, dissolution experiments using UO2 pellets are difficult and frequently lead to incoherent results. Therefore, to e...

  7. Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

    Science.gov (United States)

    González-Robles, E.; Metz, V.; Wegen, D. H.; Herm, M.; Papaioannou, D.; Bohnert, E.; Gretter, R.; Müller, N.; Nasyrow, R.; de Weerd, W.; Wiss, T.; Kienzler, B.

    2016-10-01

    During reactor operation the fission gases Kr and Xe are formed within the UO2 matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO2 fuel itself are widely used as indicators for the release properties of 129I, 137Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H2 overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

  8. Development of Collision Accident Scenario during Nuclear Spent Fuel Maritime Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Min; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Population density of South Korea is much higher than the other countries, and it is peninsula. Therefore, it is expected that major means of transportation of the spent fuel will be maritime transportation rather than overland transportation. Korea Maritime safety Tribunal (KMST) categorized various maritime accident, see table I. Among them, collision accident is one of the most important and complicated accident from Probabilistic Safety Analysis (PSA) point of view. We will show what will happen if the transportation ship is struck by other ship, how to calculate collision energy and probability of the branches on ship-ship collision with Event Tree Analysis (ETA) method. We selected and re-categorized maritime accident that KMST categorized for ship-ship collision analysis of spent fuel transportation ship. Event tree is constructed and collision energy distribution is derived from statistics and equation. And outer and inner hull fracture probabilities are calculated. If outer hull is broken but inner hull is fine, water will be flooded into the space between outer and inner hull. It will decrease mobility of the ship. If inner hull is fractured, water will be flooded into the ship inside. The ship has compartment structure to resist from foundering. Loss of mobility and compartment damage (ultimately it ends with sink) mechanism need to be analyzed to complete transportation ship collision event tree.

  9. Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

    1996-05-01

    This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

  10. Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

    1996-05-01

    This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

  11. Cladding stress during extended storage of high burnup spent nuclear fuel

    Science.gov (United States)

    Raynaud, Patrick A. C.; Einziger, Robert E.

    2015-09-01

    In an effort to assess the potential for low temperature creep and delayed hydride cracking failures in high burnup spent fuel cladding during extended dry storage, the U.S. NRC analytical fuel performance tools were used to predict cladding stress during a 300 year dry storage period for UO2 fuel burned up to 65 GWd/MTU. Fuel swelling correlations were developed and used along with decay gas production and release fractions to produce circumferential average cladding stress predictions with the FRAPCON-3.5 fuel performance code. The resulting stresses did not result in cladding creep failures. The maximum creep strains accumulated were on the order of 0.54-1.04%, but creep failures are not expected below at least 2% strain. The potential for delayed hydride cracking was assessed by calculating the critical flaw size required to trigger this failure mechanism. The critical flaw size far exceeded any realistic flaw expected in spent fuel at end of reactor life.

  12. Corrosion Surveillance for Research Reactor Spent Nuclear Fuel in Wet Basin Storage

    Energy Technology Data Exchange (ETDEWEB)

    Howell, J.P.

    1998-10-16

    Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the "Atoms for Peace" program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on "Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water" and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the RBOF basin at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993.

  13. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Summary

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This document analyzes at a programmatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For programmatic spent nuclear fuel management, this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum treatment, storage, and disposal of US Department of Energy wastes.

  14. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 2, Part A

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This document analyzes at a programmatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For programmatic spent nuclear fuel management this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum and maximum treatment, storage, and disposal of US Department of Energy wastes.

  15. 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

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This document analyzes at a pregrammatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For pregrammatic spent nuclear fuel management, this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum treatment, storage, and disposal of US Department of Energy wastes.

  16. The advanced main control console for next japanese PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, A. [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Ito, K. [Mitsubishi Heavy Industries, Ltd., Nuclear Energy Systems Engineering Center, Yokohama (Japan); Yokoyama, M. [Mitsubishi Electric Corporation, Energy and Industrial Systems Center, Kobe (Japan)

    2001-07-01

    The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

  17. Interim storage of power reactor spent nuclear fuel (SNF) and its potential application to SNF separations and closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Levy, Salomon, E-mail: slevy112@aol.com

    2009-10-15

    Interim, centralized, engineered (dry cask) storage facilities for USA light water power reactor spent nuclear fuel (SNF) should be implemented to complement and to offer much needed flexibility while the Nuclear Regulatory Commission is funded to complete its evaluation of the Yucca Mountain License and to subject it to public hearings. The interim sites should use the credo reproduced in Table 1 [Bunn, M., 2001. Interim Storage of Spent Nuclear Fuel. Harvard University and University of Tokyo] and involve both the industry and government. The sites will help settle the 50 pending lawsuits against the government and the $11 billion of potential additional liabilities for SNF delay damages if Yucca Mountain does not being operation in 2020 [DOE, 2008a. Report to Congress on the Demonstration of the Interim Storage of Spent Nuclear Fuel from Decommissioned Nuclear Power Stations (December)]. Under the developing consensus to proceed with closed fuel cycles, it will be necessary to develop SNF separation facilities with stringent requirements upon separation processes and upon generation of only highly resistant waste forms. The location of such facilities at the interim storage sites would offer great benefits to those sites and assure their long term viability by returning them to their original status. The switch from once-through to closed fuel cycle will require extensive time and development work as illustrated in 'The Path to Sustainable Nuclear Energy' [DOE, 2005. The Path to Sustainable Nuclear Energy. Basic and Applied Research Opportunities for Advanced Fuel Cycles. DOE (September)]. A carefully crafted long term program, funded for at least 5 years, managed by a strong joint government-industry team, and subjected to regular independent reviews should be considered to assure the program stability and success. The new uncertainty about Yucca Mountain role raises two key issues: (a) what to do with the weapons and other high level government

  18. The Site Selected. The Local Decision-Making Regarding the Siting of the Spent Nuclear Fuel Repository in Olkiluoto

    Energy Technology Data Exchange (ETDEWEB)

    Kojo, Matti [Univ. of Tampere (Finland). Dept. of Political Science and International Relations

    2006-09-15

    In May 1999 Posiva, the company responsible for the final disposal of spent nuclear fuel in Finland, suggested that the Finnish Government considers only Olkiluoto in Eurajoki in its application of a decision in principle to be a final disposal site. In January 2000 the municipal council of Eurajoki made a positive statement on the decision in principle. The Government made the decision in principle in Dec 2000, and the Parliament ratified the decision in May 2001. The paper is focused on the decision making of Eurajoki municipality regarding the siting of the spent nuclear fuel repository. The paper shows how the interaction between the representatives of the candidate municipality and the nuclear energy industry was the crucial factor in the decision-making. Eurajoki serves as an example, in where the parties reached an agreement of the compensations for the final disposal repository. The negotiations between the Eurajoki municipality and the nuclear energy industry in reaching a positive decision are analysed from the beginning of the 1980s. The main emphasis is however on the years 1996-99, when the nuclear energy industry negotiated with the municipality on the compensation for the final disposal repository. The loss of income was an important reason why some of the councillors of Eurajoki were interested in having the final disposal repository in Olkiluoto. The industry's problem on the other hand was to safeguard the final disposal site. From the TVO's angle Olkiluoto was a potential final disposal site for example for its limited need for transport and for the existing infrastructure. The company used the financial benefits of the project as its trump card. The attitude of Eurajoki municipality to the final disposal of spent nuclear fuel turned positive with the Olkiluoto vision in December 1998, when still five years earlier the municipal council was prepared to act and prevent the final disposal. The future image presented by the municipality

  19. A multiphase interfacial model for the dissolution of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jerden, James L., E-mail: jerden@anl.gov [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Frey, Kurt [University of Notre Dame, Notre Dame, IN 46556 (United States); Ebert, William [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States)

    2015-07-15

    Highlights: • This model accounts for chemistry, temperature, radiolysis, U(VI) minerals, and hydrogen effect. • The hydrogen effect dominates processes determining spent fuel dissolution rate. • The hydrogen effect protects uranium oxide spent fuel from oxidative dissolution. - Abstract: The Fuel Matrix Dissolution Model (FMDM) is an electrochemical reaction/diffusion model for the dissolution of spent uranium oxide fuel. The model was developed to provide radionuclide source terms for use in performance assessment calculations for various types of geologic repositories. It is based on mixed potential theory and consists of a two-phase fuel surface made up of UO{sub 2} and a noble metal bearing fission product phase in contact with groundwater. The corrosion potential at the surface of the dissolving fuel is calculated by balancing cathodic and anodic reactions occurring at the solution interfaces with UO{sub 2} and NMP surfaces. Dissolved oxygen and hydrogen peroxide generated by radiolysis of the groundwater are the major oxidizing agents that promote fuel dissolution. Several reactions occurring on noble metal alloy surfaces are electrically coupled to the UO{sub 2} and can catalyze or inhibit oxidative dissolution of the fuel. The most important of these is the oxidation of hydrogen, which counteracts the effects of oxidants (primarily H{sub 2}O{sub 2} and O{sub 2}). Inclusion of this reaction greatly decreases the oxidation of U(IV) and slows fuel dissolution significantly. In addition to radiolytic hydrogen, large quantities of hydrogen can be produced by the anoxic corrosion of steel structures within and near the fuel waste package. The model accurately predicts key experimental trends seen in literature data, the most important being the dramatic depression of the fuel dissolution rate by the presence of dissolved hydrogen at even relatively low concentrations (e.g., less than 1 mM). This hydrogen effect counteracts oxidation reactions and can limit

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

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly R. [Univ. of New Mexico, Albuquerque, NM (United States)

    2003-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    H.R. Trellue

    2003-06-01

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

  2. The development of rock suitability classification strategies in the Finnish spent nuclear fuel disposal program

    Energy Technology Data Exchange (ETDEWEB)

    Hellae, Pirjo; Hagros, Annika [Saanio and Riekkola Oy (Finland); Aaltonen, Ismo; Kosunen, Paula; Mattila, Jussi [Posiva Oy (Finland)

    2015-07-01

    This paper describes the development of the rock suitability classification strategies applied to locate the spent fuel repository in crystalline rock in Finland. Development of the classification procedure is motivated not only by the regulatory requirements, but also by the need to more closely integrate site characterization, repository design and long-term safety assessment. The classification procedure has been developed along with the increasing level of detail of the available site data and knowledge on the performance of the engineered barrier system (EBS). The classification system has also been adapted to the changes in the regulations. The present form of the classification system and experiences from testing the system at the site are described. Demonstration activities have shown that the criteria and the stepwise research, construction and decision making protocol can be applied successfully.

  3. High-density support matrices: Key to the deep borehole disposal of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gibb, F.G.F. [Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, Sheffield S1 3JD (United Kingdom)], E-mail: f.gibb@sheffield.ac.uk; McTaggart, N.A.; Travis, K.P.; Burley, D. [Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, Sheffield S1 3JD (United Kingdom); Hesketh, K.W. [Nexia Solutions Ltd., B709 Springfields, Preston PR4 0XJ (United Kingdom)

    2008-03-15

    Deep (4-5 km) boreholes are emerging as a safe, secure, environmentally sound and potentially cost-effective option for disposal of high-level radioactive wastes, including plutonium. One reason this option has not been widely accepted for spent fuel is because stacking the containers in a borehole could create load stresses threatening their integrity with potential for releasing highly mobile radionuclides like {sup 129}I before the borehole is filled and sealed. This problem can be overcome by using novel high-density support matrices deployed as fine metal shot along with the containers. Temperature distributions in and around the disposal are modelled to show how decay heat from the fuel can melt the shot within weeks of disposal to give a dense liquid in which the containers are almost weightless. Finally, within a few decades, this liquid will cool and solidify, entombing the waste containers in a base metal sarcophagus sealed into the host rock.

  4. Morphologies of uranium deposits produced during electrorefining of EBR-II spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Totemeier, T. C.

    2000-02-15

    The morphologies of U metal samples from deposits produced by electrorefining of Experimental Breeder Reactor-II (EBR-II) spent fuel were examined using scanning electron microscopy, energy- and wavelength-dispersive X-ray spectroscopy, and metallography. The morphologies were analyzed to find correlations with the chemistry of the samples, the ER run conditions, and the deposit performance. A rough correlation was observed between morphology and Zr concentration; samples with Zr contents greater than approximately 200 ppm showed fine-grained, polycrystalline dendritic morphologies, while samples with Zr contents less than approximately 100 ppm were comprised of agglomerations or linked chains of rhomboidal single crystals. There were few correlations found between morphology, run conditions, and deposit performance.

  5. Analytical solutions of the mechanical behaviour of rock with applications to a repository for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jussila, P.

    1997-10-01

    Mechanical behaviour of rock is one of the main research areas of disposal of spent nuclear fuel. Various calculation programs designed to solve the problem are utilized by the planners of spent fuel disposal. The results of the numerical approaches can be validated by comparing them to the analytical solution of a simplified problem. As a basis for this study, we used Kirsch`s equations, which give the planar solution to the stress field developing around a horizontal tunnel of a circular cross section. Kirsch`s equations and the corresponding deformation (strain) and displacement fields were derived for a tunnel with and without a circular support. The rock mass was assumed to be a continuous, homogeneous and isotropic medium obeying Hooke`s law, and the time-dependent aspects of the fields were taken into account by applying the Poynting-Thomson model to the interaction between rock and a support. The Matlab code was used to calculate and visualize the results. As a result of this study, we state that the tangential, tensile stresses at the sides of the tunnel are of the same order of magnitude as the tensile strength of rock, which influences the fracturing. With a support, the effects can be moderately reduced. 7 refs.

  6. Calculation Research of Hydrogen Production Amount in Containment after LOCAL in PWR Nuclear Power Plants%压水堆核电厂失水事故后安全壳内产氢量计算研究

    Institute of Scientific and Technical Information of China (English)

    胡建军

    2013-01-01

    The ORIGEN2 code is adopted to calculate the amount of hydrogen production in the core and sump region after LOCAL in PWR nuclear power plants,to reduce the conservatism for the design evaluation of the combustible gas control in the containment.The calculation model of radiolytic decomposition coolant and other related calculation model are used to calculate the amount of hYdrogen production after LOCA in a 600MW PWR nuclear power plant,and the results show that over conservatism of the original evaluation,and there still exists abundant time to prepare and startup the hydrogen recombiners in the containment after LOCAL.%采用ORIGEN2程序对压水堆核电厂失水事故工况下堆芯区和地坑区氢气的产生量进行计算,以合理减少安全壳内可燃气体的控制设计评价的保守性.通过冷却剂的辐照分解产氢以及其他相关计算模型,对600MW(电功率)级压水堆核电厂失水事故工况下的氢气产生量进行计算.计算结果表明原评价结果过于保守,在核电厂失水事故后仍有充分的时间准备投入安全壳内氢气复合器.

  7. Seismic stability of the survey areas of potential sites for the deep geological repository of the spent nuclear fuel

    Directory of Open Access Journals (Sweden)

    Kaláb Zdeněk

    2017-07-01

    Full Text Available This paper deals with the seismic stability of the survey areas of potential sites for the deep geological repository of the spent nuclear fuel in the Czech Republic. The basic source of data for historical earthquakes up to 1990 was the seismic website [10]. The most intense earthquake described occurred on September 15, 1590 in the Niederroesterreich region (Austria in the historical period; its reported intensity is Io = 8-9. The source of the contemporary seismic data for the period since 1991 to the end of 2014 was the website [11]. It may be stated based on the databases and literature review that in the period from 1900, no earthquake exceeding magnitude 5.1 originated in the territory of the Czech Republic.

  8. Ecohydrological Responses to Diversion of Groundwater: Case Study of a Deep-Rock Repository for Spent Nuclear Fuel in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Werner, Kent [EmpTec, Taeby (Sweden)], E-mail: kent.werner@emptec.se; Collinder, Per [Ekologigruppen AB, Stockholm (Sweden); Berglund, Sten [HydroResearch AB, Taeby (Sweden); Maartensson, Erik [DHI Sverige AB, Lund (Sweden)

    2013-05-15

    Planning and license applications concerning groundwater diversion in areas containing water-dependent or water-favored habitats must take into account both hydrological effects and associated ecological consequences. There is at present no established methodology to assess such ecohydrological responses. Thus, this paper describes a new stepwise methodology to assess ecohydrological responses to groundwater diversion from, e.g., water-drained pits, shafts, tunnels, and caverns in rock below the groundwater table. The methodology is illustrated using the planned deep-rock repository for spent nuclear fuel at Forsmark in central Sweden as a case study, offering access to a unique hydrological and ecological dataset. The case study demonstrates that results of ecohydrological assessments can provide useful inputs to planning of monitoring programs and mitigation measures in infrastructure projects. As a result of the assessment, artificial water supply to wetlands is planned in order to preserve biological diversity, nature values, and vulnerable species.

  9. Experimental observations to the electrical field for electrorefining of spent nuclear fuel in the Mark-IV electrorefiner.

    Energy Technology Data Exchange (ETDEWEB)

    Li, S. X.

    1998-05-13

    Experimental results from the pilot scale electrorefiner (Mark-IV ER) treating spent nuclear fuel are reported in this article. The electrorefining processes were carried out in a LiCl-KCl-UCl{sub 3} electrolyte. It has been noted that spool of molten cadmium below the electrolyte plays an important role in the electrorefining operations. In addition, formations of electrical shorting path between anode baskets and the electrorefiner vessel were observed, which lessened the uranium dissolution process from anode baskets, however appeared to improve the morphology of cathode deposit. The FIDAP simulation code was used to calculate the electrical potential field distributions and the potential gradient near the cathode. The effect of the electrical shorting between anode baskets and electrorefiner vessel on the morphology of cathode products is discussed.

  10. Spent nuclear fuel project multi-year work plan WBS {number_sign}1.4.1

    Energy Technology Data Exchange (ETDEWEB)

    Wells, J.L.

    1997-03-01

    The Spent Nuclear Fuel (SNF) Project Multi-Year Work Plan (MYWP) is a controlled living document that contains the current SNF Project Technical, Schedule and Cost Baselines. These baselines reflect the current Project execution strategies and are controlled via the change control process. Other changes to the MYWP document will be controlled using the document control process. These changes will be processed as they are approved to keep the MYWP a living document. The MYWP will be maintained continuously as the project baseline through the life of the project and not revised annually. The MYWP is the one document which summarizes and links these three baselines in one place. Supporting documentation for each baseline referred to herein may be impacted by changes to the MYWP, and must also be revised through change control to maintain consistency.

  11. Sensitivity analysis of parameters important to nuclear criticality safety of Castor X/28F spent nuclear fuel cask

    Energy Technology Data Exchange (ETDEWEB)

    Leotlela, Mosebetsi J. [Witwatersrand Univ., Johannesburg (South Africa). School of Physics; Koeberg Operating Unit, Johannesburg (South Africa). Regulations and Licensing; Malgas, Isaac [Koeberg Nuclear Power Station, Duinefontein (South Africa). Nuclear Engineering Analysis; Taviv, Eugene [ASARA consultants (PTY) LTD, Johannesburg (South Africa)

    2015-11-15

    In nuclear criticality safety analysis it is essential to ascertain how various components of the nuclear system will perform under certain conditions they may be subjected to, particularly if the components of the system are likely to be affected by environmental factors such as temperature, radiation or material composition. It is therefore prudent that a sensitivity analysis is performed to determine and quantify the response of the output to variation in any of the input parameters. In a fissile system, the output parameter of importance is the k{sub eff}. Therefore, in attempting to prevent reactivity-induced accidents, it is important for the criticality safety analyst to have a quantified degree of response for the neutron multiplication factor to perturbation in a given input parameter. This article will present the results of the perturbation of the parameters that are important to nuclear criticality safety analysis and their respective correlation equations for deriving the sensitivity coefficients.

  12. 78 FR 56775 - Waste Confidence-Continued Storage of Spent Nuclear Fuel

    Science.gov (United States)

    2013-09-13

    ... generated in light-water nuclear power reactors. It also covers mixed oxide (MOX) fuel,\\4\\ since the MOX... (see Section 2.1.1.3 of the DGEIS). \\4\\ Mixed oxide fuel (often called MOX fuel) is a type of...

  13. Repository for spent nuclear fuel. Plant description layout D - Forsmark; Slutfoervarsanlaeggning foer anvaent kaernbraensle. Anlaeggningsbeskrivning layout D - Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    2010-07-15

    This document describes the final repository for spent nuclear fuel, SFK, which is located at Forsmark, in Oesthammar. The bedrock at the site is part of a so-called tectonic lens, in which the rock composition is relatively homogeneous and less deformed than outside the lens. The bedrock consists mainly of granite with high quartz content and good thermal conductivity. The central parts above ground are grouped in an operations area, located at the Soederviken on the south side of the intake duct for cooling water for nuclear power plant. Operating area is divided into an internal, secured portion, where the canisters of fuel are handled and there are links to the underground part, and a outer part, where the buffer, backfill and sealing used in the repository's barriers are produced. The above-ground part of the plant and also include storage of excavated rock, ventilation stations, and supplies of bentonite. The underground portion consists of a central area and a storage area. Caverns of the central area contain features for the underground operation. It communicates with the internal operating range above ground via a spiral ramp and several shafts. The ramp used to transport capsules of spent fuel and other heavy or bulky transport. The shafts are used to transport rock, buffer, backfill and staff, as well as for ventilation. The largest part of the space below ground is the repository where the canisters with the spent fuel are disposed. The capsules are deposited in vertical holes in the tunnels. When the deposit in a tunnel is complete, the tunnel is re-filled. The two main activities underground is rock work and disposal work, which are conducted separately from each other. Rock works covers all steps required to excavate tunnels and drill deposition holes, as well as to make temporary installations in the tunnels. To the landfill works count, besides the deposit of the capsule, the placement of the bentonite buffer in the deposition hole and

  14. Determination of Neptunium, Americium and Curium in Spent Nuclear Fuel Samples by Alpha Spectrometry Using {sup 239}Np and {sup 243}Am as a Spike and a Tracer

    Energy Technology Data Exchange (ETDEWEB)

    Jeo, Kih-Soo; Song, Byung-Chul; Kim, Young-Bok; Han, Sun-Ho; Jeon, Young-Shin; Jung, Euo-Chang; Jee, Kwang-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    Determination of actinide elements and fission products in spent nuclear fuels is of importance for a burnup determination and source term evaluation. Especially, the amounts of uranium and plutonium isotopes are used for the evaluation of a burnup credit in spent nuclear fuels. Additionally, other actinides such as Np, Am and Cm in spent nuclear fuel samples is also required for the purposes mentioned above. In this study, {sup 237}Np, {sup 241}Am and {sup 244}Cm were determined by an alpha spectrometry for the source term data for high burnup spent nuclear fuels ranging from 37 to 62.9 GWD/MtU as a burnup. Generally, mass spectrometry has been known as the most powerful method for isotope determinations such as high concentrations of uranium and plutonium. However, in the case of minor actinides such as Np, Am and Cm, alpha spectrometry would be recommended instead. Determination of the transuranic elements in spent nuclear fuel samples is different from that for environmental samples because the amount of each nuclide in the spent fuel samples is higher and the relative ratios between each nuclide are also different from those for environmental samples. So, it is important to select an appropriate tracer and an optimum sample size depending on the nuclides and analytical method. In this study {sup 237}Np was determined by an isotope dilution alpha(gamma) spectrometry using {sup 239}Np as a spike, and {sup 241}Am and curium isotopes were determined by alpha spectrometry using {sup 243}Am as a tracer. The content of each nuclide was compared with that by the Origen-2 code.

  15. Development of the GANEX process for the reprocessing of Gen IV spent nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Miguirditchian, M.; Chareyre, L.; Sorel, C.; Bisel, I.; Baron, P.; Masson, M. [CEA Marcoule - DEN/DRCP/SCPS - BP 17171, Bagnols-sur-Ceze, 30207 (France)

    2008-07-01

    The GANEX (group actinide extraction) process is composed of two extraction cycles following the dissolution of the spent fuel. In the first cycle, uranium(VI) is selectively extracted from the dissolution solution using a mono amide extractant DEHiBA (NN--di-(ethyl-2-hexyl)iso-butyr-amide) diluted in HTP (hydrogen tetra-propylene). Experimental data and modelling of uranium(VI) and nitric acid extractions are presented. A flowsheet was designed and was successfully tested in laboratory scale mixer-settlers on a surrogate uranium(VI)/HNO{sub 3} feed. For the group actinide separation in the second cycle, the DIAMEX-SANEX process was adjusted to separate neptunium and plutonium along with americium and curium. The data showed the possibility to extract all actinides together with good selectivities versus lanthanides. The flowsheets of the two GANEX cycles which will be tested on a high active feed at the end of 2008 in Atalante facility are presented. (authors)

  16. SHIELDING ANALYSIS OF DUAL PURPOSE CASKS FOR SPENT NUCLEAR FUEL UNDER NORMAL STORAGE CONDITIONS

    Directory of Open Access Journals (Sweden)

    JAE-HUN KO

    2014-08-01

    The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a 2×10 cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the 2×10 cask array, dose rates at the center point of the array and at the center of the casks’ height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the 2×10 cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

  17. Economic Analysis of the Management of the Nuclear Spent Fuel in Spain

    Directory of Open Access Journals (Sweden)

    B. Yolanda Moratilla Soria

    2014-01-01

    Full Text Available This study aims to analyze the economic and technical viability for either nuclear fuel reprocessing or permanent storage in Spain. Utilizing various international studies regarding nuclear fuel reprocessing, this study reaches an objective conclusion while taking into consideration the various variable and stable costs for the open and closed cycles. A sensitivity analysis was then introduced which identifies the most influential parameters in the final price. This analysis is essential in understanding the results obtained and emphasizes the need to specify a range of costs for both cycles and to see what factors affect these results. The sensitivity analysis describes the factors that play a large role in determining costs and will display the range of values that arise from the variability of costs for those factors. The uncertainty analysis compares the nominal values used in this study and describes how these values are likely to change with time resulting in a range of values for both cycles.

  18. Seismic analysis and design of a spent nuclear fuel dry storage system

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Kyu-Sup; Jeong, In-Su; Kim, Kap-Sun; Kim, Jong-Soo [KONES Corporation, Seoul (Korea, Republic of); Yoon, Jeong-Hyoun; Kang, Young-Gon; Cho, Chun-Hyung; Lee, Heung-Young [Nuclear Environment Technology Institute, Daejeon (Korea, Republic of)

    2005-11-15

    A comprehensive seismic analysis has been conducted in this paper by using the computer programs, SHAKE and SASSI, for a consolidated dry storage module named MACSTOR/KN-400, which will be constructed in Wolsung CANDU Nuclear Power Plant site. Es