WorldWideScience

Sample records for high-burnup nuclear fuels

  1. Evaluation of safety criteria on LOCA and RIA for high burnup nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Ki; Bang, Je Geon; Kim, Dae Ho; Yang, Yong Sik; Song, Keun Woo

    2007-12-15

    Comprehensive researches in many countries and some international research programs to investigate the applicability to high burnup nuclear fuels have been performing as the existing safety criteria of DBA such as LOCA and RIA was established several decades ago. In this report, main research programs for the safety criteria of DBA such as LOCA and RIA are introduced, and also the current status on the modification of the safety criteria are also introduced.

  2. 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.

  3. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  4. High Frequency Acoustic Microscopy for the Determination of Porosity and Young's Modulus in High Burnup Uranium Dioxide Nuclear Fuel

    Science.gov (United States)

    Marchetti, Mara; Laux, Didier; Cappia, Fabiola; Laurie, M.; Van Uffelen, P.; Rondinella, V. V.; Wiss, T.; Despaux, G.

    2016-06-01

    During irradiation UO2 nuclear fuel experiences the development of a non-uniform distribution of porosity which contributes to establish varying mechanical properties along the radius of the pellet. Radial variations of both porosity and elastic properties in high burnup UO2 pellet can be investigated via high frequency acoustic microscopy. For this purpose ultrasound waves are generated by a piezoelectric transducer and focused on the sample, after having travelled through a coupling liquid. The elastic properties of the material are related to the velocity of the generated Rayleigh surface wave (VR). A UO2 pellet with a burnup of 67 GWd/tU was characterized using the acoustic microscope installed in the hot cells of the JRC-ITU at a 90 MHz frequency, with methanol as coupling liquid. VR was measured at different radial positions. A good agreement was found, when comparing the porosity values obtained via acoustic microscopy with those determined using SEM image analysis, especially in the areas close to the centre. In addition, Young's modulus was calculated and its radial profile was correlated to the corresponding burnup profile and to the hardness radial profile data obtained by Vickers micro-indentation.

  5. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

  6. Analysis of high burnup fuel behavior under control rod ejection accident in Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-07-01

    Test results of high burnup fuel behavior under RIA(reactivity insertion accident) indicated that fuel might fail at the fuel enthalpy lower than that in the current fuel failure criteria was derived by the conservative assumptions and analysis of fuel failure mechanisms, and applied to the analysis of control rod ejection accident in the 1,000 MWe Korea standard PWR. Except that three dimensional core analysis was performed instead of conventional zero dimensional analysis, all the other conservative assumptions were kept. Analysis results showed that less than on percent of the fuel rods in the core has failed which was much less than the conventional fuel failure fraction, 9.8 %, even though a newly derived fuel failure criteria -Fuel failure occurs at the power level lower than that in the current fuel failure criteria. - was applied, since transient fuel rod power level was significantly decreased by analyzing the transient fuel rod power level was significantly decreased by analyzing the transient core three dimensionally. Therefore, it can be said that results of the radiological consequence analysis for the control rod ejection accident in the FSAR where fuel failure fraction was assumed 9.8 % is still bounding. 18 tabs., 48 figs., 39 refs. (Author).

  7. Models for fuel rod behaviour at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)

    2004-12-01

    This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be

  8. Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

    Energy Technology Data Exchange (ETDEWEB)

    Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

    2011-01-01

    Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

  9. 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.

  10. Mechanical Fatigue Testing of High Burnup Fuel for Transportation Applications

    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)

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  11. 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.

  12. Summary of high burnup fuel issues and NRC`s plan of action

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, R.O.

    1997-01-01

    For the past two years the Office of Nuclear Regulatory Research has concentrated mostly on the so-called reactivity-initiated accidents -- the RIAs -- in this session of the Water Reactor Safety Information Meeting, but this year there is a more varied agenda. RIAs are, of course, not the only events of interest for reactor safety that are affected by extended burnup operation. Their has now been enough time to consider a range of technical issues that arise at high burnup, and a list of such issues being addressed in their research program is given here. (1) High burnup capability of the steady-state code (FRAPCON) used for licensing audit calculations. (2) General capability (including high burnup) of the transient code (FRAPTRAN) used for special studies. (3) Adequacy at high burnup of fuel damage criteria used in regulation for reactivity accidents. (4) Adequacy at high burnup of models and fuel related criteria used in regulation for loss-of-coolant accidents (LOCAs). (5) Effect of high burnup on fuel system damage during normal operation, including control rod insertion problems. A distinction is made between technical issues, which may or may not have direct licensing impacts, and licensing issues. The RIAs became a licensing issue when the French test in CABRI showed that cladding failures could occur at fuel enthalpies much lower than a value currently used in licensing. Fuel assembly distortion became a licensing issue when control rod insertion was affected in some operating plants. In this presentation, these technical issues will be described and the NRC`s plan of action to address them will be discussed.

  13. Assessment of reactivity transient experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ozer, O.; Yang, R.L.; Rashid, Y.R.; Montgomery, R.O.

    1996-03-01

    A few recent experiments aimed at determining the response of high-burnup LWR fuel during a reactivity initiated accident (RIA) have raised concerns that existing failure criteria may be inappropriate for such fuel. In particular, three experiments (SPERT CDC-859, NSRR HBO-1 and CABRI REP Na-1) appear to have resulted in fuel failures at only a fraction of the anticipated enthalpy levels. In evaluating the results of such RIA simulation experiments, however, it is necessary that the following two key considerations be taken into account: (1) Are the experiments representative of conditions that LWR fuel would experience during an in-reactor RIA event? (2) Is the fuel that is being utilized in the tests representative of the present (or anticipated) population of LWR fuel? Conducting experiments under conditions that can not occur in-reactor can trigger response modes that could not take place during in-reactor operation. Similarly, using unrepresentative fuel samples for the tests will produce failure information that is of limited relevance to commercial LWR fuel. This is particularly important for high-burnup fuel since the manner under which the test samples are base-irradiated prior to the test will impact the mechanical properties of the cladding and will therefore affect the RIA response. A good example of this effect can be seen in the results of the SPERT CDC-859 test and in the NSRR JM-4 and JM-5 tests. The conditions under which the fuel used for these tests was fabricated and/or base-irradiated prior to the RIA pulse resulted in the formation of multiple cladding defects in the form of hydride blisters. When this fuel was subjected to the RIA power pulse, it failed by developing multiple cracks that were closely correlated with the locations of the pre-existing hydride blisters. In the case of the JM tests, many of the cracks formed within the blisters themselves and did not propagate beyond the heavily hydrided regions.

  14. New high burnup fuel models for NRC`s licensing audit code, FRAPCON

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L. [Pacific Northwest Laboratory, Richland, WA (United States)

    1996-03-01

    Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data.

  15. The Challenges Associated with High Burnup and High Temperature for UO2 TRISO-Coated Particle Fuel

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; John Maki

    2005-02-01

    The fuel service conditions for the DOE Next Generation Nuclear Plant (NGNP) will be challenging. All major fuel related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO2 TRISO-coated particle fuel development program in the 1980s. While TRISO-coated particle fuel has been irradiated at NGNP relevant levels for two or three of the design parameters, no data exist for TRISO-coated particle fuel for all five parameters simultaneously. Of particular concern are the high burnup and high temperatures expected in the NGNP. In this paper, where possible, we evaluate the challenges associated with high burnup and high temperature quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are also discussed.

  16. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  17. Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

    Science.gov (United States)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Ishijima, Kiyomi

    1997-09-01

    To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

  18. Model for evolution of grain size in the rim region of high burnup UO2 fuel

    Science.gov (United States)

    Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng

    2016-04-01

    The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.

  19. EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

    Energy Technology Data Exchange (ETDEWEB)

    Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hanson, Brady [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billone, Mike [Argonne National Lab. (ANL), Argonne, IL (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.

  20. Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

    This report describes testing designed to determine the ability of high burnup (HBU) (>45 GWd/MTU) spent fuel to maintain its integrity under normal conditions of transportation. An innovative system, Cyclic Integrated Reversible-bending Fatigue Tester (CIRFT), has been developed at Oak Ridge National Laboratory (ORNL) to test and evaluate the mechanical behavior of spent nuclear fuel (SNF) under conditions relevant to storage and transportation. The CIRFT system is composed of a U-frame equipped with load cells for imposing the pure bending loads on the SNF rod test specimen and measuring the in-situ curvature of the fuel rod during bending using a set up with three linear variable differential transformers (LVDTs).

  1. Review of Technical Issues Related to Predicting Isotopic Compositions and Source Terms for High-Burnup LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.; Parks, C. V.

    2000-12-11

    This report has been prepared to review the technical issues important to the prediction of isotopic compositions and source terms for high-burnup, light-water-reactor (LWR) fuel as utilized in the licensing of spent fuel transport and storage systems. The current trend towards higher initial 235U enrichments, more complex assembly designs, and more efficient fuel management schemes has resulted in higher spent fuel burnups than seen in the past. This trend has led to a situation where high-burnup assemblies from operating LWRs now extend beyond the area where available experimental data can be used to validate the computational methods employed to calculate spent fuel inventories and source terms. This report provides a brief review of currently available validation data, including isotopic assays, decay heat measurements, and shielded dose-rate measurements. Potential new sources of experimental data available in the near term are identified. A review of the background issues important to isotopic predictions and some of the perceived technical challenges that high-burnup fuel presents to the current computational methods are discussed. Based on the review, the phenomena that need to be investigated further and the technical issues that require resolution are presented. The methods and data development that may be required to address the possible shortcomings of physics and depletion methods in the high-burnup and high-enrichment regime are also discussed. Finally, a sensitivity analysis methodology is presented. This methodology is currently being investigated at the Oak Ridge National Laboratory as a computational tool to better understand the changing relative significance of the underlying nuclear data in the different enrichment and burnup regimes and to identify the processes that are dominant in the high-burnup regime. The potential application of the sensitivity analysis methodology to help establish a range of applicability for experimental

  2. 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.

  3. Progress of the RIA experiments with high burnup fuels and their evaluation in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Ishijima, Kiyomi; Fuketa, Toyoshi [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-01-01

    Recent results obtained in the NSRR power burst experiments with high burnup PWR fuel rods are described and discussed in this paper. Data concerning test condition, transient records during pulse irradiation and post irradiation examination are described. Another high burnup PWR fuel rod failed in the test HBO-5 at the slightly higher energy deposition than that in the test HBO-1. The failure mechanism of the test HBO-5 is the same as that of the test HBO-1, that is, hydride-assisted PCMI. Some influence of the thermocouples welding on the failure behavior of the HBO-5 rod was observed.

  4. Development of the CANDU high-burnup fuel design/analysis technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Sim, K. S.; Oh, D. J.; Park, J. H.; Jun, J. S.; Yoo, K. J.

    1997-08-01

    This report contains all the information related to the development of the CANDU advanced fuel, so-called CANFLEX-NU, which is composed of 43 elements with natural uranium fuel. Also, it contains the compatibility study of CANFLEX-RU which is considered as a CANDU high burnup fuel. This report describes the mechanical design, thermalhydraulic and safety evaluations of CANFLEX fuel bundle. (author). 38 refs., 24 tabs., 74 figs.

  5. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)

    2005-07-01

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.

  6. New results from the NSRR experiments with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Fuketa, Toyoshi; Ishijima, Kiyomi; Mori, Yukihide [Japan Atomic Research Institute, Toaki, Ibaraki (Japan)] [and others

    1996-03-01

    Results obtained in the NSRR power burst experiments with irradiated PWR fuel rods with fuel burnup up to 50 MWd/kgU are described and discussed in this paper. Data concerning test method, test fuel rod, pulse irradiation, transient records during the pulse and post irradiation examination are described, and interpretations and discussions on fission gas release and fuel pellet fragmentation are presented. During the pulse-irradiation experiment with 50 MWd/kgU PWR fuel rod, the fuel rod failed at considerably low energy deposition level, and large amount of fission gas release and fragmentation of fuel pellets were observed.

  7. High burnup effects on fuel behaviour under accident conditions: the tests CABRI REP-Na

    Science.gov (United States)

    Schmitz, Franz; Papin, Joelle

    A large, performance based, knowledge and experience in the field of nuclear fuel behaviour is available for nominal operation conditions. The database is continuously completed and precursor assembly irradiations are performed for testing of new materials and innovative designs. This procedure produces data and arguments to extend licencing limits in the permanent research for economic competitiveness. A similar effort must be devoted to the establishment of a database for fuel behaviour under off-normal and accident conditions. In particular, special attention must be given to the so-called design-basis-accident (DBA) conditions. Safety criteria are formulated for these situations and must be respected without consideration of the occurrence probability and the risk associated to the accident situation. The introduction of MOX fuel into the cores of light water reactors and the steadily increasing goal burnup of the fuel call for research work, both experimental and analytical, in the field of fuel response to DBA conditions. In 1992, a significant programme step, CABRI REP-Na, has been launched by the French Nuclear Safety and Protection Institute (IPSN) in the field of the reactivity initiated accident (RIA). After performing the nine experiments of the initial test matrix it can be concluded that important new findings have been evidenced. High burnup clad corrosion and the associated degradation of the mechanical properties of the ZIRCALOY4 clad is one of the key phenomena of the fuel behaviour under accident conditions. Equally important is the evidence that transient, dynamic fission gas effects resulting from the close to adiabatic heating introduces a new explosive loading mechanism which may lead to clad rupture under RIA conditions, especially in the case of heterogeneous MOX fuel.

  8. Economic incentives and recommended development for commercial use of high burnup fuels in the once-through LWR fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B.; Merckx, K.R.; Holm, J.S.

    1981-01-01

    This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.

  9. Review of Halden Reactor Project high burnup fuel data that can be used in safety analyses

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W. [OECD Halden Reactor Project (Norway)

    1996-03-01

    The fuels and materials testing programmes carried out at the OECD Halden Reactor Project are aimed at providing data in support of a mechanistic understanding of phenomena, especially as related to high burnup fuel. The investigations are focused on identifying long term property changes, and irradiation techniques and instrumentation have been developed over the years which enable to assess fuel behaviour and properties in-pile. The fuel-cladding gap has an influence on both thermal and mechanical behaviour. Improved gap conductance due to gap closure at high exposure is observed even in the case of a strong contamination with released fission gas. On the other hand, pellet-cladding mechanical interaction, which is measured with cladding elongation detectors and diameter gauges, is re-established after a phase with less interaction and is increasing. These developments are exemplified with data showing changes of fuel temperature, hydraulic diameter and cladding elongation with burnup. Fuel swelling and cladding primary and secondary creep have been successfully measured in-pile. They provide data for, e.g., the possible cladding lift-off to be accounted for at high burnup. Fuel conductivity degradation is observed as a gradual temperature increase with burnup. This affects stored heat, fission gas release and temperature dependent fuel behaviour in general. The Halden Project`s data base on fission gas release shows that the phenomenon is associated with an accumulation of gas atoms at the grain boundaries to a critical concentration before appreciable release occurs. This is accompanied by an increase of the surface-to-volume ratio measured in-pile in gas flow experiments. A typical observation at high burnup is also that a burst release of fission gas may occur during a power decrease. Gas flow and pressure equilibration experiments have shown that axial communication is severely restricted at high burnup.

  10. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  11. Investigation and basic evaluation for ultra-high burnup fuel cladding material

    Energy Technology Data Exchange (ETDEWEB)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Suga, Masataka [Kokan Keisoku Co., Kawasaki, Kanagawa (Japan)

    2001-03-01

    In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)

  12. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    Energy Technology Data Exchange (ETDEWEB)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.

  13. Analysis of bubble pressure in the rim region of high burnup PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-02-01

    Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

  14. ThO{sub 2}-UO{sub 2} annular pins for high burnup fuels

    Energy Technology Data Exchange (ETDEWEB)

    Caner, Marc; Dugan, Edward T

    2000-06-01

    The main purpose of this work is to investigate the use of annular fuel pins (particularly pins containing thorium dioxide) for high burnup fuel. The following parameters were evaluated and compared between postulated mixed thorium-uranium dioxide standard and annular (9% void fraction) type fuel assemblies, as a function of burnup: the infinite multiplication factor, the uranium and plutonium isotopic compositions, the fuel temperature coefficient of reactivity and the conversion ratio. We used the SCALE-4.3 code system. The calculation method consisted in obtaining actinide and fission product number densities as functions of assembly burnup, by means of a 1-D transport calculation combined with a 0-D burnup calculation. These number densities were then used in a 3-D Monte Carlo code for obtaining k{sub {infinity}} from two-dimensional-symmetry 'snapshots'.

  15. High burnup fuel behavior related to fission gas effects under reactivity initiated accidents (RIA) conditions

    Science.gov (United States)

    Lemoine, F.

    1997-09-01

    Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup, which, under overpower conditions, can lead to solid fuel pressurization and swelling causing severe PCMI (pellet clad mechanical interaction). In order to assess the reliability of high burnup fuel under RIAs, experimental programs have been initiated which have provided important data concerning the transient fission gas behavior and the clad loading mechanisms. The importance of the rim zone is demonstrated based on three experiments resulting in clad failure at low enthalpy, which are explained by energetic considerations. High gas release in non-failure tests with low energy deposition underlines the importance of grain boundary and porosity gas. Measured final releases are strongly correlated to the microstructure evolution, depending on energy deposition, pulse width, initial and refabricated fuel rod design. Observed helium release can also increase internal pressure and gives hints to the gas behavior understanding.

  16. Study of irradiation induced restructuring of high burnup fuel - Use of computer and accelerator for fuel science and engineering -

    Energy Technology Data Exchange (ETDEWEB)

    Sataka, M.; Ishikawa, N.; Chimn, Y.; Nakamura, J.; Amaya, M. [Japan Atomic Energy Agency, Naka Gun (Japan); Iwasawa, M.; Ohnuma, T.; Sonoda, T. [Central Research Institute of Electric Power Industry, Tokyo (Japan); Kinoshita, M.; Geng, H. Y.; Chen, Y.; Kaneta, Y. [The Univ. of Tokyo, Tokyo (Japan); Yasunaga, K.; Matsumura, S.; Yasuda, K. [Kyushu Univ., Motooka (Japan); Iwase [Osaka Prefecture Univ., Osaka (Japan); Ichinomiya, T.; Nishiuran, Y. [Hokkaido Univ., Kitaku (Japan); Matzke, HJ. [Academy of Ceramics, Karlsruhe (Germany)

    2008-10-15

    In order to develop advanced fuel for future LWR reactors, trials were made to simulate the high burnup restructuring of the ceramics fuel, using accelerator irradiation out of pile and with computer simulation. The target is to reproduce the principal complex process as a whole. The reproduction of the grain subdivision (sub grain formation) was successful at experiments with sequential combined irradiation. It was made by recovery process of the accumulated dislocations, making cells and sub-boundaries at grain boundaries and pore surfaces. Details of the grain sub division mechanism is now in front of us outside of the reactor. Extensive computer science studies, first principle and molecular dynamics gave behavior of fission gas atoms and interstitial oxygen, assisting the high burnup restructuring.

  17. Angra 1 high burnup fuel behaviour under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Silva, Antonio Teixeira e, E-mail: dsgomes@ipen.b, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The 16x16 NGF (Next Generation Fuel) fuel assembly, comprising of highly corrosive-resistant ZIRLO clad fuel rods, been replacing the current 16x16 Standard (16STD) fuel assembly in the Angra 1, a pressurized water reactor, with a net output of 626 MWe. The 16x16 NGF fuel assemblies are designed for a peak rod average burnup of up to 75 GWd/MTU, thus improving fuel utilization and reducing spent fuel storage issues. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for a further increase in burnup as the simulated RIA tests revealed a lower enthalpy threshold for fuel failure. Two fuel performance codes, FRAPCON and FRAPTRAN, were used to predict high burnup behavior of Angra 1, during an RIA. The maximum average linear fuel rating used was 17.62 KW/m. The FRAPCON 3.4 code was applied to simulate the steady-state performance of the 16 NGF fuel rods up to a burnup of 55 GWd/MTU. With FRAPTRAN-1.4 the fuel behavior was simulated for an RIA power pulse of 4.5 ms (FHWH), and enthalpy peak of 130 Cal/g. With FRAPCON-3.4, the corrosion and hydrogen pickup characteristics of the advanced ZIRLO clad fuel rods were added to the code by modifying the actual corrosion model for Zircaloy-4 through the multiplication of empirical factors, which were appropriate to each alloy, and by means of reducing the current hydrogen pickup fraction. (author)

  18. Oxygen potential in the rim region of high burnup UO 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1994-01-01

    Small specimens from the rim region (fuel surface) of a UO 2 fuel rod with an average burnup of 7.6% FIMA were analysed in a miniaturized galvanic cell to determine their oxygen potential Δ Ḡ(O 2) . These fuel pieces represented the porous rim structure with very small grains known to be formed near the periphery of high burnup UO 2 fuel pellets. The oxygen potential of the rim material was very low, corresponding to that of unirradiated stoichiometric UO 2, or to that of slightly substoichiometric UO 2 containing rare earth fission products. No indication of oxidation due to fission was found, though most fission was that of Pu. Measurements on pieces from the inner, unrestructured fuel showed somewhat higher oxygen potentials corresponding to those of very slightly substoichiometric fuel if allowance is made for the incorporation of rare earths. These results are in contrast to some generally accepted ideas of burnup effects, and the possible reasons and implications are discussed.

  19. Oxygen potential measurements in high burnup LWR U0 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1995-05-01

    A miniature solid state galvanic cell was used to measure the oxygen potential Δ overlineG( O2) of reactor irradiated U0 2 fuel at different burnups in the range of 28 to ⩾ 150 GWd d/t M. This very high burnup was achieved in the rim region of a fuel with a cross section average burnup of 75 GWd d/t M. The fuels had different enrichments and therefore different contributions of fission of 235U and 239Pu. The temperature range covered was 900 to 1350 K. None of the fuels showed a significant oxidation. Rather, if allowance is made for the dissolved rare earth fission products and the Pu formed during irradiation, some of the fuels were very slightly substoichiometric and the highest possible degree of oxidation corresponded to U0 2.001. In general, the Δ overlineG( O2) at 750°C was about -400 kJ/mol, corresponding to the Δ overlineG( O2) of the reaction Mo + O 2 → MoO 2. The implication of these results which are in contrast to commonly assumed ideas that U0 2 fuel oxidizes due to burnup, are discussed and the importance of the fission product Mo and of the zircaloy clad as oxygen buffers is outlined.

  20. Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-12-15

    The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional {sup 235}U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using {sup 233}U, {sup 242}Pu, {sup 150}Nd, and {sup 133}Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

  1. A study of fuel failure behavior in high burnup HTGR fuel. Analysis by STRESS3 and STAPLE codes

    Energy Technology Data Exchange (ETDEWEB)

    Martin, David G.; Sawa, Kazuhiro; Ueta, Shouhei; Sumita, Junya [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-05-01

    In current high temperature gas-cooled reactors (HTGRs), Tri-isotropic coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. This report attempts to model fuel behavior in irradiation tests using the U.K. codes STRESS3 and STAPLE. Test results in 91F-1A and HRB-22 capsules irradiation tests, which were carried out at the Japan Materials Testing Reactor of JAERI and at the High Flux Isotope Reactor of Oak Ridge National Laboratory, respectively, were employed in the calculation. The maximum burnup and fast neutron fluence were about 10%FIMA and 3 x 10{sup 25} m{sup -2}, respectively. The fuel for the irradiation tests was called high burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the High Temperature Engineering Test Reactor. The calculation results demonstrated that if only mean fracture stress values of PyC and SiC are used in the calculation it is not possible to predict any particle failures, by which is meant when all three load bearing layers have failed. By contrast, when statistical variations in the fracture stresses and particle specifications are taken into account, as is done in the STAPLE code, failures can be predicted. In the HRB-22 irradiation test, it was concluded that the first two particles which had failed were defective in some way, but that the third and fourth failures can be accounted for by the pressure vessel model. In the 91F-1A irradiation test, the result showed that 1 or 2 particles had failed towards the end of irradiation in the upper capsule and no particles failed in the lower capsule. (author)

  2. Thermochemical prediction of chemical form distributions of fission products in LWR oxide fuels irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-09-01

    Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. In addition, the effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that approximately 13 mole% of the total amount of Cs compounds exists as CsI and virtually remaining fraction as Cs{sub 2}MoO{sub 4} under the operation condition of LHR below 400 W/cm. On the other hand, when LHR is beyond 400 W/cm under the transient operation condition, its distribution did not change so much from the one under normal operation condition. (author)

  3. Separation of metallic residues from the dissolution of a high-burnup BWR fuel using nitrogen trifluoride

    Energy Technology Data Exchange (ETDEWEB)

    McNamara, Bruce K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soderquist, Chuck Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mausolf, Edward J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Scheele, Randall D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-03-23

    Nitrogen trifluoride (NF3) was used to fluorinate the metallic residue from the dissolution of a high burnup, boiling water reactor fuel (~70 MWd/kgU). The metallic residue included the noble metal phase (containing ruthenium, rhodium, palladium, technetium, and molybdenum), and smaller amounts of zirconium, selenium, tellurium, and silver. Exposing the noble metal phase to 10% NF3 in argon between 400 and 550°C, removed molybdenum and technetium near 400°C as their volatile fluorides, and ruthenium near 500C as its volatile fluoride. The events were thermally and temporally distinct and the conditions specified are a recipe to separate these transition metals from each other and from the noble metal phase nonvolatile residue. Depletion of the volatile fluorides resulted in substantial exothermicity. Thermal excursion behavior was recorded under non-adiabatic, isothermal conditions that typically minimize heat release. Physical characterization of the metallic noble phase and its thermal behavior are consistent with high kinetic velocity reactions encouraged by the nanoparticulate phase or perhaps catalytic influences of the mixed platinum metals with nearly pure phase structure. Post-fluorination, only two phases were present in the residual nonvolatile fraction. These were identified as a nano-crystalline, metallic palladium cubic phase and a hexagonal rhodium trifluoride (RhF3) phase. The two phases were distinct as the sub-µm crystallites of metallic palladium were in contrast to the RhF3 phase, which grew from the parent nano-crystalline noble-metal phase during fluorination, to acicular crystals exceeding 20-µm in length.

  4. Post Irradiation Examination Plan for High-Burnup Demonstration Project Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Montgomery, Rose [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    This test plan describes the experimental work to be implemented by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) to characterize high burnup (HBU) spent nuclear fuel (SNF) in conjunction with the High Burnup Dry Storage Cask Research and Development Project and serves to coordinate and integrate the multi-year experimental program to collect and develop data regarding the continued storage and eventual transport of HBU (i.e., >45 GWd/MTU) SNF. The work scope involves the development, performance, technical integration, and oversight of measurements and collection of relevant data, guided by analyses and demonstration of need.

  5. Corrosion studies with high burnup light water reactor fuel. Release of nuclides into simulated groundwater during accumulated contact time of up to two years

    Energy Technology Data Exchange (ETDEWEB)

    Zwicky, Hans-Urs (Zwicky Consulting GmbH, Remigen (Switzerland)); Low, Jeanett; Ekeroth, Ella (Studsvik Nuclear AB, Nykoeping (Sweden))

    2011-03-15

    In the framework of comprehensive research work supporting the development of a Swedish concept for the disposal of highly radioactive waste and spent fuel, Studsvik has performed a significant number of spent fuel corrosion studies under a variety of different conditions. These experiments, performed between 1990 and 2002, covered a burnup range from 27 to 49 MWd/kgU, which was typical for fuel to be disposed at that time. As part of this work, the so called Series 11 tests were performed under oxidising conditions in synthetic groundwater with fuel samples from a rod irradiated in the Ringhals 1 Boiling Water Reactor (BWR). In the meantime, Swedish utilities tend to increase the discharge burnup of fuel operated in their reactors. This means that knowledge of spent fuel corrosion performance has to be extended to higher burnup as well. Therefore, a series of experiments has been started at Studsvik, aiming at extending the data base acquired in the Series 11 corrosion tests to higher burnup fuel. Fuel burnup leads to complex and significant changes in the composition and properties of the fuel. The transformed microstructure, which is referred to as the high burnup structure or rim structure in the outer region of the fuel, consists of small grains of submicron size and a high concentration of pores of typical diameter 1 to 2 mum. This structure forms in UO{sub 2} fuel at a local burnup above 50 MWd/kgU, as long as the temperature is below 1,000-1,100 deg C. The high burnup at the pellet periphery is the consequence of plutonium build-up by neutron capture in 238U followed by fission of the formed plutonium. The amount of fission products in the fuel increases more or less linearly with burnup, in contrast to alpha emitting actinides that increase above average. As burnup across a spent fuel pellet is not uniform, but increases towards the periphery, the radiation field is also larger at the pellet surface. At the same time, it is easier for water to access the

  6. Measurement of the composition of noble-metal particles in high-burnup CANDU fuel by wavelength dispersive X-ray microanalysis

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H.; Szostak, F.J

    1999-09-01

    An investigation of the composition of the metallic inclusions in CANDU fuel, which contain Mo, Tc, Ru, Rh and Pd, has been conducted as a function of burnup by wavelength dispersive X-ray (WDX) microanalysis. Quantitative measurements were performed on micrometer sized particles embedded in thin sections of fuel using elemental standards and the ZAF method. Because the fission yields of the noble metals change with burnup, as a consequence of a shift from almost entirely {sup 235}U fission to mainly {sup 239}Pu fission, their inventories were calculated from the fuel power histories using the WIMS-Origin code for comparison with experiment. Contrary to expectations that the oxygen potential would be buffered by progressive Mo oxidation, little evidence was obtained for reduced incorporation of Mo in the noble-metal particles at high burnup. These surprising results are discussed with respect to the oxygen balance in irradiated CANDU fuels and the likely intrinsic and extrinsic sinks for excess oxygen. (author)

  7. Determination of plutonium content in high burnup pressurized water reactor fuel samples and its use for isotope correlations for isotopic composition of plutonium.

    Science.gov (United States)

    Joe, Kihsoo; Jeon, Young-Shin; Han, Sun-Ho; Lee, Chang-Heon; Ha, Yeong-Keong; Song, Kyuseok

    2012-06-01

    The content of plutonium isotopes in high burnup pressurized water reactor fuel samples was examined using both alpha spectrometry and mass spectrometry after anion exchange separation. The measured values were compared with results calculated by the ORIGEN-2 code. On average, the ratios (m/c) of the measured values (m) over the calculated values (c) were 1.22±0.16 for (238)Pu, 1.02±0.14 for (239)Pu, 1.08±0.06 for (240)Pu, 1.06±0.16 for (241)Pu, and 1.13±0.08 for (242)Pu. Using the Pu data obtained in this work, correlations were derived between the alpha activity ratios of (238)Pu/((239)Pu+(240)Pu), the alpha specific activities of Pu, and the atom % abundances of the Pu isotopes. Using these correlations, the atom % abundances of the plutonium isotopes in the target samples were calculated. These calculated results agreed within a range from 2 to 8% of the experimentally derived values according to the isotopes of plutonium.

  8. 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.

  9. Power excursion analysis for BWR`s at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D.J.; Neymoith, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    A study has been undertaken to determine the fuel enthalpy during a rod drop accident and during two thermal-hydraulic transients. The objective was to understand the consequences to high burnup fuel and the sources of uncertainty in the calculations. The analysis was done with RAMONA-4B, a computer code that models the neutron kinetics throughout the core along with the thermal-hydraulics in the core, vessel, and steamline. The results showed that the maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important parameters in each of these categories are discussed in the paper.

  10. Analysis of the effect of UO{sub 2} high burnup microstructure on fission gas release

    Energy Technology Data Exchange (ETDEWEB)

    Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2002-10-01

    This report deals with high-burnup phenomena with relevance to fission gas release from UO{sub 2} nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas.

  11. A semi-empirical model for the formation and depletion of the high burnup structure in UO2

    Science.gov (United States)

    Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.

    2017-04-01

    In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.

  12. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, T.A.

    1980-03-01

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation.

  13. Twenty-second water reactor safety information meeting. Volume 2: Severe accident research, thermal hydraulic research for advanced passive LWRs, high-burnup fuel behavior

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.

    1995-04-01

    This three-volume report contains papers presented at the Twenty-Second Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 24-26, 1994. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Finland, France, Italy, Japan, Russia, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  14. Vented nuclear fuel element

    Science.gov (United States)

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

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

  15. High Burnup Dry Storage Cask Research and Development Project, Final Test Plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-02-27

    EPRI is leading a project team to develop and implement the first five years of a Test Plan to collect data from a SNF dry storage system containing high burnup fuel.12 The Test Plan defined in this document outlines the data to be collected, and the storage system design, procedures, and licensing necessary to implement the Test Plan.13 The main goals of the proposed test are to provide confirmatory data14 for models, future SNF dry storage cask design, and to support license renewals and new licenses for ISFSIs. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must mimic real conditions that high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to the ISFSI for multi-year storage.15 Along with other optional modeling, SETs, and SSTs, the data collected in this Test Plan can be used to evaluate the integrity of dry storage systems and the high burnup fuel contained therein over many decades. It should be noted that the Test Plan described in this document discusses essential activities that go beyond the first five years of Test Plan implementation.16 The first five years of the Test Plan include activities up through loading the cask, initiating the data collection, and beginning the long-term storage period at the ISFSI. The Test Plan encompasses the overall project that includes activities that may not be completed until 15 or more years from now, including continued data collection, shipment of the Research Project Cask to a Fuel Examination Facility, opening the cask at the Fuel Examination Facility, and examining the high burnup fuel after the initial storage period.

  16. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  17. Composite nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Dollard, W.J.; Ferrari, H.M.

    1982-04-27

    An open lattice elongated nuclear fuel assembly including small diameter fuel rods disposed in an array spaced a selected distance above an array of larger diameter fuel rods for use in a nuclear reactor having liquid coolant flowing in an upward direction. Plenums are preferably provided in the upper portion of the upper smaller diameter fuel rods and in the lower portion of the lower larger diameter fuel rods. Lattice grid structures provide lateral support for the fuel rods and preferably the lowest grid about the upper rods is directly and rigidly affixed to the highest grid about the lower rods.

  18. Fuel performance annual report for 1990. Volume 8

    Energy Technology Data Exchange (ETDEWEB)

    Preble, E.A.; Painter, C.L.; Alvis, J.A.; Berting, F.M.; Beyer, C.E.; Payne, G.A. [Pacific Northwest Lab., Richland, WA (United States); Wu, S.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology

    1993-11-01

    This annual report, the thirteenth in a series, provides a brief description of fuel performance during 1990 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience and trends, fuel problems high-burnup fuel experience, and items of general significance are provided . References to additional, more detailed information, and related NRC evaluations are included where appropriate.

  19. French investigations of high burnup effect on LOCA thermomecanical behavior. Part two. Oxidation and quenching experiments under simulated LOCA conditions with high burnup clad material

    Energy Technology Data Exchange (ETDEWEB)

    GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)] [and others

    1997-01-01

    In the frame of the high burnup fuel studies to support a possible extension of the current discharge burnup limit, experimental programs have been undertaken, jointly by EDF and IPSN in order to study the thermal-shock behavior of high burnup fuel claddings under typical LOCA conditions. The TAGUS program used unirradiated cladding samples, bare or bearing a pre-corrosion state simulating the end-of-life state of high burnup fuel claddings: the TAGCIR program used actually irradiated cladding samples taken from high burnup rods irradiated over 5 cycles in a commercial EDF PWR and having reached a rod burnup close to 60 GWd/tU. The thermal-shock failure tests consisted in oxidizing the cladding samples under steam flow, on both inner and outer faces or on the outer face alone, and subjecting them to a final water quench. The heating was provided by an inductive furnace the power of which being regulated through monitoring of the sample surface temperature with use of a single-wave optical pyrometer. Analysis of the irradiated tests (TAGCIR series) evidenced an increased oxidation rate as compared to similar tests on unirradiated samples. Results of the quenching tests series on unirradiated and irradiated samples are plotted under the usual presentation of failure maps relative to the oxidation parameters ECR (equivalent cladding reacted) or e{sub {beta}} (thickness of the remaining beta phase layer) as a function of the oxidation temperature. Comparison of the failure limits for irradiated specimens to those for unirradiated specimens indicates a lower brittleness under two side oxidation and possibly the opposite under one-side oxidation. The tentative analysis of the oxidation and quenching tests results on irradiated samples reveals the important role played by the hydrogen charged during in-reactor corrosion on the oxidation kinetics and the failure bearing capability of the cladding under LOCA transient conditions.

  20. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Michael F. Simpson; Jack D. Law

    2010-02-01

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

  1. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

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

  2. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport- Demonstration of Approach and Results on Used Fuel Performance Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, Harold [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Ken [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Koeppel, Brian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bignell, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Flores, Gregg [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Jy-An [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sanborn, Scott [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Spears, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Klymyshyn, Nick [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-30

    This document addresses Oak Ridge National Laboratory milestone M2FT-13OR0822015 Demonstration of Approach and Results on Used Nuclear Fuel Performance Characterization. This report provides results of the initial demonstration of the modeling capability developed to perform preliminary deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of transport (NCT) conditions. This report also provides results from the sensitivity studies that have been performed. Finally, discussion on the long-term goals and objectives of this initiative are provided.

  3. Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

    Energy Technology Data Exchange (ETDEWEB)

    Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

    2006-10-31

    The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

  4. Rethinking nuclear fuel recycling.

    Science.gov (United States)

    von Hippel, Frank N

    2008-05-01

    Spent nuclear fuel contains plutonium which can be extracted and used in new fuel. To reduce the amount of long-lived radioactive waste, the U.S. Department of Energy has proposed reprocessing spent fuel in this way and then "burning" the plutonium in special reactors. But reprocesssing is very expensive. Also, spent fuel emits lethal radiation, whereas separated plutonium can be handled easily. So reprocessing invites the possibility that terrorists might steal plutonium and construct an atom bormb. The authors argue against reprocessing and for storing the waste in casks until an underground repository is ready.

  5. Fuel performance annual report for 1989

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.; Berting, F.M. (Pacific Northwest Lab., Richland, WA (United States)); Wu, S. (Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology)

    1992-06-01

    This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulatory Commission evaluations are included.

  6. Fuel performance annual report for 1986

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.; Wu, S.

    1988-03-01

    This annual report, the ninth in a series, provides a brief description of fuel performance during 1986 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related U.S. Nuclear Regulatory Commission evaluations are included. 550 refs., 12 figs., 31 tabs.

  7. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect I: Effects of variations of the fuel particle volume fractions

    Science.gov (United States)

    Wang, Qiming; Yan, Xiaoqing; Ding, Shurong; Huo, Yongzhong

    2010-05-01

    A new method of modeling the in-pile mechanical behaviors of dispersion nuclear fuel elements is proposed. Considering the irradiation swelling together with the thermal effect, numerical simulations of the in-pile mechanical behaviors are performed with the developed finite element models for different fuel particle volume fractions of the fuel meat. The effects of the particle volume fractions on the mechanical performances of the fuel element are studied. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the particle volume fractions at each burnup; the locations of the maximum first principal stresses shift with increasing burnup; at low burnups, the maximum first principal stresses increase with the particle volume fractions; while at high burnups, the 20% volume fraction case holds the lowest value; (2) at the cladding, the maximum equivalent plastic strains and the tensile principal stresses increase with the particle volume fractions; while the maximum Mises stresses do not follow this order at high burnups; (3) the maximum Mises stresses at the fuel particles increase with the particle volume fractions, and the particles will engender plastic strains until the particle volume fraction reaches high enough.

  8. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Harold F. McFarlane; Terry Todd

    2013-11-01

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

  9. 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.

  10. Fuel performance annual report for 1991. Volume 9

    Energy Technology Data Exchange (ETDEWEB)

    Painter, C.L.; Alvis, J.M.; Beyer, C.E. [Pacific Northwest Lab., Richland, WA (United States); Marion, A.L. [Oregon State Univ., Corvallis, OR (United States). Dept. of Nuclear Engineering; Payne, G.A. [Northwest Coll. and Univ. Association for Science, Richland, WA (United States); Kendrick, E.D. [Nuclear Regulatory Commission, Washington, DC (United States)

    1994-08-01

    This report is the fourteenth in a series that provides a compilation of information regarding commercial nuclear fuel performance. The series of annual reports were developed as a result of interest expressed by the public, advising bodies, and the US Nuclear Regulatory Commission (NRC) for public availability of information pertaining to commercial nuclear fuel performance. During 1991, the nuclear industry`s focus regarding fuel continued to be on extending burnup while maintaining fuel rod reliability. Utilities realize that high-burnup fuel reduces the amount of generated spent fuel, reduces fuel costs, reduces operational and maintenance costs, and improves plant capacity factors by extending operating cycles. Brief summaries of fuel operating experience, fuel design changes, fuel surveillance programs, high-burnup experience, problem areas, and items of general significance are provided.

  11. Effects of Lower Drying-Storage Temperature on the Ductility of High-Burnup PWR Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Burtseva, T. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-08-30

    The purpose of this research effort is to determine the effects of canister and/or cask drying and storage on radial hydride precipitation in, and potential embrittlement of, high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures (PCT) and hoop stresses. Extensive precipitation of radial hydrides could lower the failure hoop stresses and strains, relative to limits established for as-irradiated cladding from discharged fuel rods stored in pools, at temperatures below the ductile-to-brittle transition temperature (DBTT).

  12. Using Finite Model Analysis and Out of Hot Cell Surrogate Rod Testing to Analyze High Burnup Used Nuclear Fuel Mechanical Properties

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

    Based on a series of FEA simulations, the discussions and the conclusions concerning the impact of the interface bonding efficiency to SNF vibration integrity are provided in this report; this includes the moment carrying capacity distribution between pellets and clad, and the impact of cohesion bonding on the flexural rigidity of the surrogate rod system. As progressive de-bonding occurs at the pellet-pellet interfaces and at the pellet-clad interface, the load ratio of the bending moment carrying capacity gradually shifts from the pellets to the clad; the clad starts to carry a significant portion of the bending moment resistance until reaching the full de-bonding state at the pellet-pellet interface regions. This results in localized plastic deformation of the clad at the pellet-pellet-clad interface region; the associated plastic deformations of SS clad leads to a significant degradation in the stiffness of the surrogate rod. For instance, the flexural rigidity was reduced by 39% from the perfect bond state to the de-bonded state at the pellet-pellet interfaces.

  13. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  14. 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.

  15. Nuclear reactors and fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

  16. Nuclear Fuel Cycle & Vulnerabilities

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, Brian D. [Los Alamos National Laboratory

    2012-06-18

    The objective of safeguards is the timely detection of diversion of significant quantities of nuclear material from peaceful nuclear activities to the manufacture of nuclear weapons or of other nuclear explosive devices or for purposes unknown, and deterrence of such diversion by the risk of early detection. The safeguards system should be designed to provide credible assurances that there has been no diversion of declared nuclear material and no undeclared nuclear material and activities.

  17. Modeling the Nuclear Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  18. Alternatives for nuclear fuel disposal

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Badillo A, V.; Palacios H, J.; Celis del Angel, L., E-mail: ramon.ramirez@inin.gob.m [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2010-10-15

    The spent fuel is one of the most important issues in the nuclear industry, currently spent fuel management is been cause of great amount of research, investments in the construction of repositories or constructing the necessary facilities to reprocess the fuel, and later to recycle the plutonium recovered in thermal reactors. What is the best solution? or, What is the best technology for a specific solution? Many countries have deferred the decision on selecting an option, while other works actively constructing repositories and others implementing the reprocessing facilities to recycle the plutonium obtained from nuclear spent fuel. In Mexico the nuclear power is limited to two reactors BWR type and medium size. So the nuclear spent fuel discharged has been accommodated at reactor's spent fuel pools. Originally these pools have enough capacity to accommodate spent fuel for the 40 years of designed plant operation. However, currently is under process an extended power up rate to 20% of their original power and also there are plans to extend operational life for 20 more years. Under these conditions there will not be enough room for spent fuel in the pools. So this work describes some different alternatives that have been studied in Mexico to define which will be the best alternative to follow. (Author)

  19. Nuclear Fuel Cycle Introductory Concepts

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-02

    The nuclear fuel cycle is a complex entity, with many stages and possibilities, encompassing natural resources, energy, science, commerce, and security, involving a host of nations around the world. This overview describes the process for generating nuclear power using fissionable nuclei.

  20. Nuclear fuels - Present and future

    Science.gov (United States)

    Olander, D.

    2009-06-01

    The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

  1. Used Fuel Testing Transportation Model

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); LeDuc, Dan [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-25

    This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

  2. Used Fuel Testing Transportation Model

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B.; Best, Ralph E.; Maheras, Steven J.; Jensen, Philip J.; England, Jeffery L.; LeDuc, Dan

    2014-09-24

    This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

  3. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  4. Research on high-burn fuel: the Cabri water loop reactor project; La investigacion sobre combustible de alto quemado: El proyecto del reactor Cabri con lazo de agua

    Energy Technology Data Exchange (ETDEWEB)

    Conde Lopez, J. M.; Recio Santamaria, M. [Consejo de Seguridad Nuclear. Madrid (Spain)

    2000-07-01

    This paper describes the present status of the nuclear fuel utilization both in the national and international arenas. Details on the fuel-related research programs actually ongoing are given. The interests and strategic lines stablished by the CSN regarding high burnup fuel research are described. Specifically, an analysis is made of the reasons underlying the CSN's decision to participate in the IPSN's CABRI Water Loop high burnup fuel research program. The core of the paper is devoted to the description of the CABRI project contents and of the technical aspects in the behaviour of high burnup fuel that will be studied through the foreseen integral and separate effects tests. Finally, a summary of the project organization, schedule and technical program is included. (Author)

  5. 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.

  6. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  7. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  8. Nuclear reactor composite fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

  9. Protected Nuclear Fuel Element

    Science.gov (United States)

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

    1962-12-01

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

  10. 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.

  11. Fully ceramic nuclear fuel and related methods

    Science.gov (United States)

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

    2016-03-29

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

  12. An empirical formulation to describe the evolution of the high burnup structure

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-15

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  13. An empirical formulation to describe the evolution of the high burnup structure

    Science.gov (United States)

    Lemes, Martín; Soba, Alejandro; Denis, Alicia

    2015-01-01

    In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in

  14. Compositions and methods for treating nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-28

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

  15. Compositions and methods for treating nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-13

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

  16. 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

    hardness and thermal conductivity. In order to reproduce cumulative damage effects expected after very long storage time within acceptable laboratory timescales, accelerated damage build-up conditions are applied, e.g. by using unirradiated (U,Pu) oxide with high specific alpha-activity (alpha-doped UO{sub 2}). The measurements performed so far show that saturation of macroscopic hardness and thermal conductivity occurs for a simulated timescale corresponding to spent fuel after decades or centuries of storage (the exact time scale depends on parameters like fuel composition and burnup). The outcome of the studies also show a non-negligible extent of lattice swelling, peaking at an accumulated damage level of ∝1 dpa. The behavior observed in tailor-made samples is compared to actual high burnup fuel (UO{sub 2}, MOX), to verify extent and significance of eventual macroscopic alterations of spent fuel during long-term storage. This paper will present the main outcome of the extended storage studies performed on alpha-doped UO{sub 2} and on spent fuel. The experimental approach to investigate the possible impact of spent fuel rods alterations on their retrieval after extended storage, and the expected behavior under accident conditions will be described. Extrapolations to very long timeframes corresponding to spent fuel evolution after emplacement in the geologic repository will be illustrated.

  17. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-18

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

  18. Nuclear Fuels: Present and Future

    Directory of Open Access Journals (Sweden)

    Donald R. Olander

    2009-02-01

    Full Text Available The important new developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of these fuels and the reactors they power are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel-rod designs, the hydride fuel with liquid metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the Very High Temperature Reactor and the Sodium Fast Reactor, and the accompanying reprocessing technologies, aqueous-based UREX and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the material's behavior under irradiation and in the reprocessing schemes are emphasized.

  19. Model biases in high-burnup fast reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Touran, N.; Cheatham, J.; Petroski, R. [TerraPower LLC, 11235 S.E. 6th St, Bellevue, WA 98004 (United States)

    2012-07-01

    A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

  20. FUEL COMPOSITION FOR NUCLEAR REACTORS

    Science.gov (United States)

    Andersen, J.C.

    1963-08-01

    A process for making refractory nuclear fuel elements involves heating uranium and silicon powders in an inert atmosphere to 1600 to 1800 deg C to form USi/sub 3/; adding silicon carbide, carbon, 15% by weight of nickel and aluminum, and possibly also molybdenum and silicon powders; shaping the mixture; and heating to 1700 to 2050 deg C again in an inert atmosphere. Information on obtaining specific compositions is included. (AEC)

  1. 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.

  2. Thermal property change of MOX and UO2 irradiated up to high burnup of 74 GWd/t

    Science.gov (United States)

    Nakae, Nobuo; Akiyama, Hidetoshi; Miura, Hiromichi; Baba, Toshikazu; Kamimura, Katsuichiro; Kurematsu, Shigeru; Kosaka, Yuji; Yoshino, Aya; Kitagawa, Takaaki

    2013-09-01

    Thermal property is important because it controls fuel behavior under irradiation. The thermal property change at high burnup of more than 70 GWd/t is examined. Two kinds of MOX fuel rods, which were fabricated by MIMAS and SBR methods, and one referenced UO2 fuel rod were used in the experiment. These rods were taken from the pre-irradiated rods (IFA 609/626, of which irradiation test were carried out by Japanese PWR group) and re-fabricated and re-irradiated in HBWR as IFA 702 by JNES. The specification of fuel corresponds to that of 17 × 17 PWR type fuel and the axially averaged linear heat rates (LHR) of MOX rods are 25 kW/m (BOL of IFA 702) and 20 kW/m (EOL of IFA 702). The axial peak burnups achieved are about 74 GWd/t for both of MOX and UO2. Centerline temperature and plenum gas pressure were measured in situ during irradiation. The measured centerline temperature is plotted against LHR at the position where thermocouples are fixed. The slopes of MOX are corresponded to each other, but that of UO2 is higher than those of MOX. This implies that the thermal conductivity of MOX is higher than that of UO2 at high burnup under the condition that the pellet-cladding gap is closed during irradiation. Gap closure is confirmed by the metallography of the postirradiation examinations. It is understood that thermal conductivity of MOX is lower than that of UO2 before irradiation since phonon scattering with plutonium in MOX becomes remarkable. A phonon scattering with plutonium decreases in MOX when burnup proceeds. Thus, thermal conductivity of MOX becomes close to that of UO2. A reverse phenomenon is observed at high burnup region. The phonon scattering with fission products such as Nd and Zr causes a degradation of thermal conductivity of burnt fuel. It might be speculated that this scattering effect causes the phenomenon and the mechanism is discussed here.

  3. ALD coating of nuclear fuel actinides materials

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M.; Pellin, Michael J.; Yun, Di; Billone, Mike

    2017-09-05

    The invention provides a method of forming a nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, with the steps of obtaining a fuel form in a powdered state; coating the fuel form in a powdered state with at least one layer of a material; and sintering the powdered fuel form into a fuel pellet. Also provided is a sintered nuclear fuel pellet of a uranium containing fuel alternative to UO.sub.2, wherein the pellet is made from particles of fuel, wherein the particles of fuel are particles of a uranium containing moiety, and wherein the fuel particles are coated with at least one layer between about 1 nm to about 4 nm thick of a material using atomic layer deposition, and wherein the at least one layer of the material substantially surrounds each interfacial grain barrier after the powdered fuel form has been sintered.

  4. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-07-11

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  5. Concepts for Small-Scale Testing of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven Craig [Idaho National Laboratory; Winston, Philip Lon [Idaho National Laboratory

    2015-09-01

    This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

  6. Storage, transportation and disposal system for used nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Scaglione, John M.; Wagner, John C.

    2017-01-10

    An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

  7. Additional Studies of the Criticality Safety of Failed Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, William BJ J [ORNL; Wagner, John C [ORNL

    2013-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories and specific configurations were evaluated to understand trends and quantify the consequences of worst-case potential reconfiguration progressions. These results will be summarized here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g., >20% keff). It can be concluded that the consequences of credible fuel failure configurations from ES or transportation following ES are manageable (e.g., <5% keff). The current work expands on these efforts and examines some modified scenarios and modified approaches to investigate the effectiveness of some techniques for reducing the calculated increase in keff. The areas included here are more realistic modeling of some assembly types and the effect of reconfiguration of some assemblies in the storage and transportation canister.

  8. Nuclear fuel grid outer strap

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, R.; Craver, J.E.

    1989-10-10

    This patent describes a nuclear reactor fuel assembly grid. It comprises a first outer grip strap segment end. The first end having a first tab arranged in substantially the same plane as the plane defined by the first end; a second outer grip strap end. The second end having a second slot arranged in substantially the same plane as the plane defined by the second end, with the tab being substantially disposed in the slot, defining a socket therebetween; and a fort tine interposed substantially perpendicularly in the socket.

  9. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Massaro, Lawrence M. [Federal Railroad Administration (FRA) (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-10-01

    visits. Every site was found to have at least one off-site transportation mode option for removing its UNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 12 shutdown sites use designs from 4 different suppliers involving 9 different (horizontal and vertical) dry storage systems that would require the use of 8 different transportation cask designs to remove the UNF and GTCC waste from the shutdown sites; Although there are common aspects, each site has some unique features and/or conditions; Although some regulatory actions will be required, all UNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in licensed systems that can be transported, including a small amount of high-burnup fuel; Each site indicated that 2-3 years of advance time would be required for its preparations before shipments could begin; Most sites have more than one transportation option, e.g., rail, barge, or heavy haul truck, as well as constraints and preferences. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

  10. Studies of nuclear fuel by means of nuclear spectroscopy methods

    Energy Technology Data Exchange (ETDEWEB)

    Jansson, Peter

    2000-02-01

    This paper is a summary text of several works performed by the author regarding spectroscopic measurements on spent nuclear fuel. Methods for determining the decay heat of spent nuclear fuel by means of gamma-ray spectroscopy and for verifying the integrity of nuclear fuel by means of tomography is presented. A summary of work performed regarding gamma-ray detector technology for studies of fission gas release is presented.

  11. OECD - HRP Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  12. International Summer School on Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on nuclear fuel in the period August 28 September 1, 2000. The summer school was primarily intended for people who wanted to become acquainted with fuel-related subjects and issues without being experts. It was especially hoped that the summer school would serve to transfer knowledge to the ''young generation'' in the field of nuclear fuel. Experts from Halden Project member organisations gave the following presentations: (1) Overview of the nuclear community, (2) Criteria for safe operation and design of nuclear fuel, (3) Fuel design and fabrication, (4) Cladding Manufacturing, (5) Overview of the Halden Reactor Project, (6) Fuel performance evaluation and modelling, (7) Fission gas release, and (8) Cladding issues. Except for the Overview, which is a written paper, the other contributions are overhead figures from spoken lectures.

  13. Nuclear Fusion Fuel Cycle Research Perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Hongsuk; Koo, Daeseo; Park, Jongcheol; Kim, Yeanjin [KAERI, Daejeon (Korea, Republic of); Yun, Sei-Hun [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    As a part of the International Thermonuclear Experimental Reactor (ITER) Project, we at the Korea Atomic Energy Research Institute (KAERI) and our National Fusion Research Institute (NFRI) colleagues are investigating nuclear fusion fuel cycle hardware including a nuclear fusion fuel Storage and Delivery System (SDS). To have a better knowledge of the nuclear fusion fuel cycle, we present our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). To have better knowledge of the nuclear fusion fuel cycle, we presented our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). Our efforts to enhance the tritium confinement will be continued for the development of cleaner nuclear fusion power plants.

  14. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Massaro, Lawrence M. [Fermi Research Alliance (FRA), Batavia, IL (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-30

    important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. Additional conclusions from this evaluation include: The 13 shutdown sites use designs from 4 different suppliers involving 11 different (horizontal and vertical) dry storage systems that would require the use of 9 different transportation cask designs to remove the SNF and GTCC waste from the shutdown sites. Although some changes to transportation certificates of compliance will be required, the SNF at the initial 9 shutdown sites (Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion) is in dual purpose dry storage canisters that can be transported, including a small amount of high-burnup fuel. Most sites indicated that 2-3 years of advance time would be required for its preparations before shipments could begin. Some sites could be ready in less time. As additional sites such as Fort Calhoun, Clinton, Quad Cities, Pilgrim, Oyster Creek, and Diablo Canyon shut down, these sites will be included in updates to the evaluation.

  15. Variants of closing the nuclear fuel cycle

    Science.gov (United States)

    Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.; Tsibulskiy, S. V.

    2015-12-01

    Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.

  16. Sustainability Features of Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Stefano Passerini

    2012-09-01

    Full Text Available The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC is the current fuel cycle implemented in the United States; in which an appropriate form of the fuel is irradiated through a nuclear reactor only once before it is disposed of as waste. The discharged fuel contains materials that can be suitable for use as fuel. Thus, different types of fuel recycling technologies may be introduced in order to more fully utilize the energy potential of the fuel, or reduce the environmental impacts and proliferation concerns about the discarded fuel materials. Nuclear fuel cycle systems analysis is applied in this paper to attain a better understanding of the strengths and weaknesses of fuel cycle alternatives. Through the use of the nuclear fuel cycle analysis code CAFCA (Code for Advanced Fuel Cycle Analysis, the impact of a number of recycling technologies and the associated fuel cycle options is explored in the context of the U.S. energy scenario over 100 years. Particular focus is given to the quantification of Uranium utilization, the amount of Transuranic Material (TRU generated and the economics of the different options compared to the base-line case, the OTC option. It is concluded that LWRs and the OTC are likely to dominate the nuclear energy supply system for the period considered due to limitations on availability of TRU to initiate recycling technologies. While the introduction of U-235 initiated fast reactors can accelerate their penetration of the nuclear energy system, their higher capital cost may lead to continued preference for the LWR-OTC cycle.

  17. The IFR modern nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hannum, W.H.

    1991-01-01

    Nuclear power is an essential component of the world's energy supply. The IFR program, by returning to fundamentals, offers a fresh approach to closing the nuclear fuel cycle. This closed fuel cycle represents the ultimate in efficient resource utilization and environmental accountability. 35 refs., 2 tabs.

  18. Multiphase Nanocrystalline Ceramic Concept for Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mecartnery, Martha [Univ. of California, Irvine, CA (United States); Graeve, Olivia [Univ. of California, San Diego, CA (United States); Patel, Maulik [Univ. of Liverpool (United Kingdom)

    2017-05-25

    The goal of this research is to help develop new fuels for higher efficiency, longer lifetimes (higher burn-up) and increased accident tolerance in future nuclear reactors. Multiphase nanocrystalline ceramics will be used in the design of simulated advanced inert matrix nuclear fuel to provide for enhanced plasticity, better radiation tolerance, and improved thermal conductivity

  19. 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.

  20. 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

  1. Modeling and Simulation of Nuclear Fuel Materials

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Van Brutzel, Laurent; Chartier, Alan; Gueneau, Christine; Mattsson, Ann E.; Tikare, Veena; Bartel, Timothy; Besmann, T. M.; Stan, Marius; Van Uffelen, Paul

    2010-10-01

    We review the state of modeling and simulation of nuclear fuels with emphasis on the most widely used nuclear fuel, UO2. The hierarchical scheme presented represents a science-based approach to modeling nuclear fuels by progressively passing information in several stages from ab initio to continuum levels. Such an approach is essential to overcome the challenges posed by radioactive materials handling, experimental limitations in modeling extreme conditions and accident scenarios, and the small time and distance scales of fundamental defect processes. When used in conjunction with experimental validation, this multiscale modeling scheme can provide valuable guidance to development of fuel for advanced reactors to meet rising global energy demand.

  2. Strategy for Used Fuel Acquisition

    Energy Technology Data Exchange (ETDEWEB)

    Steven C. Marschman; Chris Rusch

    2013-09-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The Storage and Transportation staffs within the UFDC are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While both wet and dry storage have been shown to be safe options for storing UNF, the focus of the program is on dry storage at reactor or centralized locations. Because limited information is available on the properties of high burnup fuel (exceeding 45 gigawatt-days per metric tonne of uranium [GWd/MTU]), and because much of the fuel currently discharged from today’s reactors exceeds this burnup threshold, a particular emphasis of this program is on high burnup fuels. Since high burnup used fuels have only been loaded into dry storage systems in the past decade or so, these materials are available to the UFDC for testing in only very limited quantities. Much of what is available has come via NRC testing programs. Some of these fuels may have achieved "high burnup," but that does not mean they were designed for high burnup use (e.g. lower enrichments, smaller plenum spaces, extra reactor cycles). The handling and transfer of these materials from utility to laboratory has not always been

  3. Establishment of China Nuclear Fuel Assembly Database

    Institute of Scientific and Technical Information of China (English)

    CHENPeng; ZHANGYing-chao; LIUTing-jin; JINYong-li

    2003-01-01

    During researching, designing, manufacturing and post irradiation, a large amount of data on fuel assembly of China nuclear power plants has been accumulated. It is necessary to collect the data together,so that the researchers, designers, manufactures and managers could use the data conveniently. It was proposed to establish a China Nuclear Fuel Assembly Database through the Internet on workstations during the year of 2003 to 2006, so the data would be shared in China nuclear industry.

  4. Nuclear fuel elements having a composite cladding

    Science.gov (United States)

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

    1983-09-20

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

  5. Annotated Bibliography for Drying Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rebecca E. Smith

    2011-09-01

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

  6. Globalisation of the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Rougeau, J.-P.; Durret, L.-F.

    1995-12-31

    Three main features of the globalisation of the nuclear fuel cycle are identified and discussed. The first is an increase in the scale of the nuclear fuel cycle materials and services markets in the past 20 years. This has been accompanied by a growth in the sophistication of the fuel cycle. Secondly, the nuclear industry is now more vulnerable to outside pressures; it is no longer possible to make strategic decisions on the industry within a country solely on national considerations. Thirdly, there are changes in the decision-making process at the political, regulatory, operational and industrial level which are the consequence of global factors. (UK).

  7. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-05-01

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

  8. A high burnup model developed for the DIONISIO code

    Energy Technology Data Exchange (ETDEWEB)

    Soba, A. [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Denis, A., E-mail: denis@cnea.gov.ar [U.A. Combustibles Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Romero, L. [U.A. Reactores Nucleares, Comisión Nacional de Energía Atómica, Avenida del Libertador 8250, 1429 Buenos Aires (Argentina); Villarino, E.; Sardella, F. [Departamento Ingeniería Nuclear, INVAP SE, Comandante Luis Piedra Buena 4950, 8430 San Carlos de Bariloche, Río Negro (Argentina)

    2013-02-15

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO{sub 2} fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in {sup 235}U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  9. A high burnup model developed for the DIONISIO code

    Science.gov (United States)

    Soba, A.; Denis, A.; Romero, L.; Villarino, E.; Sardella, F.

    2013-02-01

    A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in 235U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.

  10. Development of Fabrication Technology for Ceramic Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H. S.; Lee, Y. W.; Na, S. H.; Kim, Y. G.; Jung, C. Y.; Kim, S. H.; Lee, S. C.; Son, D. S

    2006-04-15

    annular (U,Ce)O{sub 2} pellet - Fabrication technology of IMF pellet - Derivation of the improvement methods of the pellet characteristics from SimMOX analysis. The great performance of KAERI MOX of which has been irradiating in HALDEN reactor is a fact in support of the capability of our unique technology. It is a fundamental technology which can be applied to improve fuel performance and safety and to fabricate a new type of fuel for next generation as well. The database constructed with the unique technology supplies how to fabricate a fuel pellet which has a given density, grain size and pore distribution. Man power, time schedule and budget can be saved by using this technology in a workshop or a research group as reducing the repetition of trial and error. The superiority of the mill developed by the unique technology was proved by an on the spot test in a fuel production workshop. The workshop plans to apply this mill to fabricate a burnable poison fuel pellet or to recover scrap powder. The glove box technology can be used in a nuclear fuel company or in a relative workshop in order to enhance the work safety and the efficiency. To achieve both mixing homogeneity and sinterability of a powder mixture is a key technology to fabricate high burnup MOX, IMF or SimMOX pellet. This project developed a milling machine and a powder treatment technology, obtained a patent for the technology. This technology can be used in a general ceramic plant as well as a nuclear fuel field to improve quality and productivity.

  11. 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.

  12. 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.

  13. Nuclear Fuels & Materials Spotlight Volume 4

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-04-01

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

  14. Nuclear Fuel Cycle Evaluation and Real Options

    Directory of Open Access Journals (Sweden)

    L. Havlíček

    2008-01-01

    Full Text Available The first part of this paper describes the nuclear fuel cycle. It is divided into three parts. The first part, called Front-End, covers all activities connected with fuel procurement and fabrication. The middle part of the cycle includes fuel reload design activities and the operation of the fuel in the reactor. Back-End comprises all activities ensuring safe separation of spent fuel and radioactive waste from the environment. The individual stages of the fuel cycle are strongly interrelated. Overall economic optimization is very difficult. Generally, NPV is used for an economic evaluation in the nuclear fuel cycle. However the high volatility of uranium prices in the Front-End, and the large uncertainty of both economic and technical parameters in the Back-End, make the use of NPV difficult. The real option method is able to evaluate the value added by flexibility of decision making by a company under conditions of uncertainty. The possibility of applying this method to the nuclear fuel cycle evaluation is studied. 

  15. Fundamental aspects of nuclear reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Olander, D.R.

    1976-01-01

    The book presented is designed to function both as a text for first-year graduate courses in nuclear materials and as a reference for workers involved in the materials design and performance aspects of nuclear power plants. The contents are arranged under the following chapter headings: statistical thermodynamics, thermal properties of solids, crystal structures, cohesive energy of solids, chemical equilibrium, point defects in solids, diffusion in solids, dislocations and grain boundaries, equation of state of UO/sub 2/, fuel element thermal performance, fuel chemistry, behavior of solid fission products in oxide fuel elements, swelling due to fission gases, pore migration and fuel restructuring kinetics, fission gas release, mechanical properties of UO/sub 2/, radiation damage, radiation effects in metals, interaction of sodium and stainless steel, modeling of the structural behavior of fuel elements and assemblies. (DG)

  16. Dry Transfer Systems for Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Brett W. Carlsen; Michaele BradyRaap

    2012-05-01

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

  17. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH.

  18. Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Weber, William [Univ. of Tennessee, Knoxville, TN (United States)

    2016-09-20

    years for high-burnup used fuels, and in 10,000 years for current commercial used nuclear fuel. In collaboration with Peking University, investigations on helium damage accumulation in yttria-stabilized zirconia (YSZ) found that the critical helium concentration to form bubbles in YSZ (3.2 at%) is much higher than in nanocrystalline ZrO2 and CeO2; the difference is speculated to be due to trapping of helium in the structural oxygen vacancy sites present in YSZ. At a peak helium concentration of 25 at%, which is not expected in any used nuclear fuel for at least 1 million years, the volumetric swelling due to bubble formation in YSZ is 27%. Comparable swelling might be expected at shorter time scales in actual spent nuclear fuel that does not have structural vacancies. While beta decay is not expected to affect the performance of used nuclear fuel, alpha decay will result in the accumulation of helium that will form helium bubbles once the helium concentration exceeds about 0.15 at% helium, which will occur within 100, 1000 or 10,000 years for MOX, highburnup or current commercial used fuels, respectively. Due to the rim-effect in high-burnup nuclear fuels, bubble formation may be greatly accelerated and substantially higher in the rim region of nuclear fuel pellets because the local burnup is much higher (factor of two or more) than the average burnup reported for used nuclear fuel, which could lead to possible fragmentation of the outer rim on geologic time scales. Two model waste form materials based on the pyrochlore structure, Gd2Ti2O7 (radiation sensitive) and Gd2Zr2O7 (radiation tolerant), have been investigated. The Gd2Ti2O7 and Gd2Zr2O7 samples were irradiated at room temperature to peak doses corresponding to 6 dpa in order to create a fully amorphous state in Gd2Ti2O7 and the

  19. Waste Stream Analyses for Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    N. R. Soelberg

    2010-08-01

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

  20. Nuclear fuel supply view in Argentina

    Energy Technology Data Exchange (ETDEWEB)

    Cirimello, R.O. [Comision Nacional de Energia Atomica, Conuar SA (Argentina)

    1997-07-01

    The Argentine Atomic Energy Commission promoted and participated in a unique achievement in the R and D system in Argentina: the integration of science technology and production based on a central core of knowledge for the control and management of the nuclear fuel cycle technology. CONUAR SA, as a fuel manufacturer, FAE SA, the manufacturer of Zircaloy tubes, CNEA and now DIOXITEC SA producer of Uranium Dioxide, have been supply, in the last ten years, the amount of products required for about 1300 Tn of equivalent U content in fuels. The most promising changes for the fuel cycle economy is the Slight Enriched Uranium project which begun in Atucha I reactor. In 1997 seventy five fuel assemblies, equivalent to 900 Candu fuel bundles, will complete its irradiation. (author)

  1. Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

    Science.gov (United States)

    Settle, Frank A.

    2009-01-01

    The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

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

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg; Steve Piet

    2010-11-01

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

  3. Evaluation of the Frequency for Gas Sampling for the High Burnup Confirmatory Data Project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-01

    This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations.

  4. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1992-05-01

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

  5. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I W; Mitchell, S J

    1990-01-01

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

  6. International nuclear fuel cycle fact book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1988-01-01

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

  7. Nuclear fuel supply: challenges and opportunities

    Energy Technology Data Exchange (ETDEWEB)

    Lowen, S. [Cameco Corp., Saskatoon, Saskatchewan (Canada)

    2006-07-01

    Prices of uranium, conversion services and enrichment services have all significantly increased in the last few years. These price increases have generally been driven by a tightening in the supply of these products and services, mostly due to long lead times required to bring these products and services to the market. This paper will describe the various steps in the nuclear fuel cycle for natural and enriched uranium fuel, will discuss the development of the front-end fuel cycle for low void reactivity fuel, and will address the challenges faced in the long-term supply of each component, particularly in the light of potential demand increases as a result of a nuclear renaissance. The opportunities for new capacity and uranium production will be outlined and the process required to achieve sufficient new supply will be discussed. (author)

  8. Double-clad nuclear fuel safety rod

    Science.gov (United States)

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

    1984-01-01

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

  9. 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.

  10. 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.

  11. Computational Design of Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Savrasov, Sergey [Univ. of California, Davis, CA (United States); Kotliar, Gabriel [Rutgers Univ., Piscataway, NJ (United States); Haule, Kristjan [Rutgers Univ., Piscataway, NJ (United States)

    2014-06-03

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

  12. Reference Neutron Radiographs of Nuclear Reactor Fuel

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

  13. Seismic response of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Hlaváč Z.

    2014-06-01

    Full Text Available The paper deals with mathematical modelling and computer simulation of the seismic response of fuel assembly components. The seismic response is investigated by numerical integration method in time domain. The seismic excitation is given by two horizontal and one vertical synthetic accelerograms at the level of the pressure vessel seating. Dynamic response of the hexagonal type nuclear fuel assembly is caused by spatial motion of the support plates in the reactor core investigated on the reactor global model. The modal synthesis method with condensation is used for calculation of the fuel assembly component displacements and speeds on the level of the spacer grid cells.

  14. SOLID GAS SUSPENSION NUCLEAR FUEL ASSEMBLY

    Science.gov (United States)

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

    1962-05-01

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

  15. Nuclear reactor composite fuel assembly. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, D.M.; Cappiello, M.W.; Marr, D.R.; Omberg, R.P.

    1980-11-25

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

  16. Nuclear Fuels & Materials Spotlight Volume 5

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-10-01

    As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • Evaluation and modeling of light water reactor accident tolerant fuel concepts • Status and results of recent TRISO-coated particle fuel irradiations, post-irradiation examinations, high-temperature safety testing to demonstrate the accident performance of this fuel system, and advanced microscopy to improve the understanding of fission product transport in this fuel system. • Improvements in and applications of meso and engineering scale modeling of light water reactor fuel behavior under a range of operating conditions and postulated accidents (e.g., power ramping, loss of coolant accident, and reactivity initiated accidents) using the MARMOT and BISON codes. • Novel measurements of the properties of nuclear (actinide) materials under extreme conditions, (e.g. high pressure, low/high temperatures, high magnetic field) to improve the scientific understanding of these materials. • Modeling reactor pressure vessel behavior using the GRIZZLY code. • New methods using sound to sense temperature inside a reactor core. • Improved experimental capabilities to study the response of fusion reactor materials to a tritium plasma. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at Idaho National Laboratory, and hope that you find this issue informative.

  17. Thorium nuclear fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Eom, Tae Yoon; Do, Jae Bum; Choi, Yoon Dong; Park, Kyoung Kyum; Choi, In Kyu; Lee, Jae Won; Song, Woong Sup; Kim, Heong Woo

    1998-03-01

    Since thorium produces relatively small amount of TRU elements after irradiation in the reactor, it is considered one of possible media to mix with the elements to be transmuted. Both solid and molten-salt thorium fuel cycles were investigated. Transmutation concepts being studied involved fast breeder reactor, accelerator-driven subcritical reactor, and energy amplifier with thorium. Long-lived radionuclides, especially TRU elements, could be separated from spent fuel by a pyrochemical process which is evaluated to be proliferation resistance. Pyrochemical processes of IFR, MSRE and ATW were reviewed and evaluated in detail, regarding technological feasibility, compatibility of thorium with TRU, proliferation resistance, their economy and safety. (author). 26 refs., 22 figs

  18. Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-01

    The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bending Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.

  19. Nuclear Energy and Synthetic Liquid Transportation Fuels

    Science.gov (United States)

    McDonald, Richard

    2012-10-01

    This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

  20. 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.

  1. 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.

  2. 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.

  3. Multidimensional multiphysics simulation of nuclear fuel behavior

    Science.gov (United States)

    Williamson, R. L.; Hales, J. D.; Novascone, S. R.; Tonks, M. R.; Gaston, D. R.; Permann, C. J.; Andrs, D.; Martineau, R. C.

    2012-04-01

    Nuclear fuel operates in an environment that induces complex multiphysics phenomena, occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. This multiphysics behavior is often tightly coupled and many important aspects are inherently multidimensional. Most current fuel modeling codes employ loose multiphysics coupling and are restricted to 2D axisymmetric or 1.5D approximations. This paper describes a new modeling tool able to simulate coupled multiphysics and multiscale fuel behavior, for either 2D axisymmetric or 3D geometries. Specific fuel analysis capabilities currently implemented in this tool are described, followed by a set of demonstration problems which include a 10-pellet light water reactor fuel rodlet, three-dimensional analysis of pellet clad mechanical interaction in the vicinity of a defective fuel pellet, coupled heat transfer and fission product diffusion in a TRISO-coated fuel particle, a demonstration of the ability to couple to lower-length scale models to account for material property variation with microstructural evolution, and a demonstration of the tool's ability to efficiently solve very large and complex problems using massively-parallel computing. A final section describes an early validation exercise, comparing simulation results to a light water reactor fuel rod experiment.

  4. 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.

  5. Nuclear Fuel Cycle Options Catalog FY15 Improvements and Additions.

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Barela, Amanda Crystal [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Schetnan, Richard Reed [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Walkow, Walter M. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The United States Department of Energy, Office of Nuclear Energy, Fuel Cycle Technology Program sponsors nuclear fuel cycle research and development. As part of its Fuel Cycle Options campaign, the DOE has established the Nuclear Fuel Cycle Options Catalog. The catalog is intended for use by the Fuel Cycle Technologies Program in planning its research and development activities and disseminating information regarding nuclear energy to interested parties. The purpose of this report is to document the improvements and additions that have been made to the Nuclear Fuel Cycle Options Catalog in the 2015 fiscal year.

  6. Nuclear reactor fuel element. Kernreaktorbrennelement

    Energy Technology Data Exchange (ETDEWEB)

    Lippert, H.J.

    1985-03-28

    The fuel element box for a BWR is situated with a corner bolt on the inside in one corner of its top on the top side of the top plate. This corner bolt is screwed down with a bolt with a corner part which is provided with leaf springs outside on two sides, where the bolt has a smaller diameter and an expansion shank. The bolt is held captive to the bolt head on the top and the holder on the bottom of the corner part. The holder is a locknut. If the expansion forces are too great, the bolt can only break at the expansion shank.

  7. 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.

  8. 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.

  9. 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.

  10. Supply Security in Future Nuclear Fuel Markets

    Energy Technology Data Exchange (ETDEWEB)

    Seward, Amy M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wood, Thomas W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gitau, Ernest T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ford, Benjamin E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-11-18

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

  11. 77 FR 19278 - Informational Meeting on Nuclear Fuel Cycle Options

    Science.gov (United States)

    2012-03-30

    ... criteria or the pros and cons of any particular fuel cycle option. Opportunity for providing input on the... Informational Meeting on Nuclear Fuel Cycle Options AGENCY: Office of Fuel Cycle Technologies, Office of Nuclear Energy, Department of Energy. ACTION: Notice of meeting. SUMMARY: The Office of Fuel Cycle...

  12. 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 ...

  13. Nuclear power generation and fuel cycle report 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

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

  14. Development and Experimental Benchmark of Simulations to Predict Used Nuclear Fuel Cladding Temperatures during Drying and Transfer Operations

    Energy Technology Data Exchange (ETDEWEB)

    Greiner, Miles [Univ. of Nevada, Reno, NV (United States)

    2017-03-31

    Radial hydride formation in high-burnup used fuel cladding has the potential to radically reduce its ductility and suitability for long-term storage and eventual transport. To avoid this formation, the maximum post-reactor temperature must remain sufficiently low to limit the cladding hoop stress, and so that hydrogen from the existing circumferential hydrides will not dissolve and become available to re-precipitate into radial hydrides under the slow cooling conditions during drying, transfer and early dry-cask storage. The objective of this research is to develop and experimentallybenchmark computational fluid dynamics simulations of heat transfer in post-pool-storage drying operations, when high-burnup fuel cladding is likely to experience its highest temperature. These benchmarked tools can play a key role in evaluating dry cask storage systems for extended storage of high-burnup fuels and post-storage transportation, including fuel retrievability. The benchmarked tools will be used to aid the design of efficient drying processes, as well as estimate variations of surface temperatures as a means of inferring helium integrity inside the canister or cask. This work will be conducted effectively because the principal investigator has experience developing these types of simulations, and has constructed a test facility that can be used to benchmark them.

  15. Nuclear rocket using indigenous Martian fuel NIMF

    Science.gov (United States)

    Zubrin, Robert

    1991-01-01

    In the 1960's, Nuclear Thermal Rocket (NTR) engines were developed and ground tested capable of yielding isp of up to 900 s at thrusts up to 250 klb. Numerous trade studies have shown that such traditional hydrogen fueled NTR engines can reduce the inertial mass low earth orbit (IMLEO) of lunar missions by 35 percent and Mars missions by 50 to 65 percent. The same personnel and facilities used to revive the hydrogen NTR can also be used to develop NTR engines capable of using indigenous Martian volatiles as propellant. By putting this capacity of the NTR to work in a Mars descent/acent vehicle, the Nuclear rocket using Indigenous Martian Fuel (NIMF) can greatly reduce the IMLEO of a manned Mars mission, while giving the mission unlimited planetwide mobility.

  16. Survey of nuclear fuel-cycle codes

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, C.R.; de Saussure, G.; Marable, J.H.

    1981-04-01

    A two-month survey of nuclear fuel-cycle models was undertaken. This report presents the information forthcoming from the survey. Of the nearly thirty codes reviewed in the survey, fifteen of these codes have been identified as potentially useful in fulfilling the tasks of the Nuclear Energy Analysis Division (NEAD) as defined in their FY 1981-1982 Program Plan. Six of the fifteen codes are given individual reviews. The individual reviews address such items as the funding agency, the author and organization, the date of completion of the code, adequacy of documentation, computer requirements, history of use, variables that are input and forecast, type of reactors considered, part of fuel cycle modeled and scope of the code (international or domestic, long-term or short-term, regional or national). The report recommends that the Model Evaluation Team perform an evaluation of the EUREKA uranium mining and milling code.

  17. Holdup measurement for nuclear fuel manufacturing plants

    Energy Technology Data Exchange (ETDEWEB)

    Zucker, M.S.; Degen, M.; Cohen, I.; Gody, A.; Summers, R.; Bisset, P.; Shaub, E.; Holody, D.

    1981-07-13

    The assay of nuclear material holdup in fuel manufacturing plants is a laborious but often necessary part of completing the material balance. A range of instruments, standards, and a methodology for assaying holdup has been developed. The objectives of holdup measurement are ascertaining the amount, distribution, and how firmly fixed the SNM is. The purposes are reconciliation of material unbalance during or after a manufacturing campaign or plant decommissioning, to decide security requirements, or whether further recovery efforts are justified.

  18. Holdup measurement for nuclear fuel manufacturing plants

    Energy Technology Data Exchange (ETDEWEB)

    Zucker, M.S.; Degen, M.; Cohen, I.; Gody, A.; Summers, R.; Bisset, P.; Shaub, E.; Holody, D.

    1981-07-13

    The assay of nuclear material holdup in fuel manufacturing plants is a laborious but often necessary part of completing the material balance. A range of instruments, standards, and a methodology for assaying holdup has been developed. The objectives of holdup measurement are ascertaining the amount, distribution, and how firmly fixed the SNM is. The purposes are reconciliation of material unbalance during or after a manufacturing campaign or plant decommissioning, to decide security requirements, or whether further recovery efforts are justified.

  19. Current Comparison of Advanced Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-04-01

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

  20. POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

    Science.gov (United States)

    Dwyer, O.E.

    1958-12-23

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

  1. 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.

  2. 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.

  3. Rail Shock and Vibration Pre-Test Modeling of a Used Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B.; Klymyshyn, Nicholas A.; Jensen, Philip J.; Best, Ralph E.; Maheras, Steven J.; McConnell, Paul E.; Orchard, John

    2015-04-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Available information is not sufficient to determine the ability of ES UNF, including high-burnup fuel, to withstand shock and vibration forces that could occur when the UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility. There are three major gaps in the available information – 1) the forces that UNF assemblies would be subjected to when transported by rail, 2) the mechanical characteristics of fuel rod cladding, which is an essential structure for controlling the geometry of the UNF, a safety related feature, and 3) modeling methodologies to evaluate multiple possible degradation or damage mechanisms over the UNF lifetime. In order to address the first gap, options for tests to determine the physical response of surrogate UNF assemblies subjected to shock and vibration forces that are expected to be experienced during normal conditions of transportation (NCT) by rail must be identified and evaluated. The objective of the rail shock and vibration tests is to obtain data that will help researchers understand the mechanical loads that ES UNF assemblies would be subjected to under normal conditions of transportation and to fortify the computer modeling that will be necessary to evaluate the impact

  4. International trade and waste and fuel managment issue, 2007

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal (ed.)

    2007-01-15

    The focus of the January-February issue is on international trade and waste and fuel managment. Major articles/reports in this issue include: New plants with high safety and availability, by Bill Poirier, Westinghouse Electric Company; Increased reliability and competitiveness, by Russell E. Stachowski, GE Energy, Nuclear; Fuel for long-term supply of nuclear power, by Kumiaki Moriya, Hitachi, Ltd., Japan; Super high burnup fuel, By Noboru Itagaki and Tamotsu Murata, Nuclear Fuel Industries LTD., Japan; Zero fuel failures by 2010, by Tom Patten, AREVA NP Inc.; Decommissioning opportunities in the UK, by David Brown and William Thorn, US Department of Commerce; Industry's three challenges, by Dale E. Klein, US Nuclear Regulatory Commission; and, A step ahead of the current ABWR's, compiled by Claire Zurek, GE Energy.

  5. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

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

    2008-07-22

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

  6. Ultrasonic spectral analysis for nuclear fuel characterization

    Energy Technology Data Exchange (ETDEWEB)

    Baroni, Douglas B.; Bittencourt, Marcelo S.Q.; Leal, Antonio M.M., E-mail: douglasbaroni@ien.gov.b, E-mail: bittenc@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    Ceramic materials have been widely used for various purposes in many different industries due to certain characteristics, such as high melting point and high resistance to corrosion. Concerning the areas of applications, automobile, aeronautics, naval and even nuclear, the characteristics of these materials should be strictly controlled. In the nuclear area, ceramics are of great importance once they are the nuclear fuel pellets and must have, among other features, a well controlled porosity due to mechanical strength and thermal conductivity required by the application. Generally, the techniques used to characterize nuclear fuel are destructive and require costly equipment and facilities. This paper aims to present a nondestructive technique for ceramic characterization using ultrasound. This technique differs from other ultrasonic techniques because it uses ultrasonic pulse in frequency domain instead of time domain, associating the characteristics of the analyzed material with its frequency spectrum. In the present work, 40 Alumina (Al{sub 2}O{sub 3}) ceramic pellets with porosities ranging from 5% to 37%, in absolute terms measured by Archimedes technique, were tested. It can be observed that the frequency spectrum of each pellet varies according to its respective porosity and microstructure, allowing a fast and non-destructive association of the same characteristics with the same spectra pellets. (author)

  7. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels -- Final Report under the International Nuclear Energy Research Initiative (I-NERI)

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David [Idaho National Lab. (INL), Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab. (INEEL); Martin, Philippe [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Phelip, Mayeul [Commissariat a l' Energie Atomique et aux Energies Alternatives (CEA-Saclay), Gif-sur-Yvette (France); Ballinger, Ronald [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2004-12-01

    The objective of this INERI project was to develop improved fuel behavior models for gas reactor coated-particle fuels and to explore improved coated-particle fuel designs that could be used reliably at very high burnups and potentially in gas-cooled fast reactors. Project participants included the Idaho National Engineering Laboratory (INEEL), Centre Étude Atomique (CEA), and the Massachusetts Institute of Technology (MIT). To accomplish the project objectives, work was organized into five tasks.

  8. 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.

  9. Dynamic Systems Analysis Report for Nuclear Fuel Recycle

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-12-01

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

  10. 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.

  11. Nuclear safety. Technical progress journal, October 1996--December 1996

    Energy Technology Data Exchange (ETDEWEB)

    None

    1996-01-01

    The five papers in this issue address various issues associated with the behavior of high burnup fuels, especially under reactivity initiated accident (RIA) conditions. The mechanisms and parameters that have an effect on the fuel behavior are detailed, based on tests and analyses. The ultimate goal of the research reported is the development of new regulatory criteria for high burnup fuel under design basis accident conditions. Specific topics of the papers, which are abstracted individually in the database, are: (1) regulatory assessment of test data for RIAs, (2) high burnup fuel transient behavior under RIA conditions, (3) NSRR/RIA experiments with high burnup PWR fuels, (4) the Russian RIA research program, and (5) RIA simulation experiments on the intermediate and high burnup test rods. The papers are contributed from the United States, France, Japan, and Russia.

  12. 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.

  13. Composite nuclear fuel fabrication methodology for gas fast reactors

    Science.gov (United States)

    Vasudevamurthy, Gokul

    An advanced fuel form for use in Gas Fast Reactors (GFR) was investigated. Criteria for the fuel includes operation at high temperature (˜1400°C) and high burnup (˜150 MWD/MTHM) with effective retention of fission products even during transient temperatures exceeding 1600°C. The GFR fuel is expected to contain up to 20% transuranics for a closed fuel cycle. Earlier evaluations of reference fuels for the GFR have included ceramic-ceramic (cercer) dispersion type composite fuels of mixed carbide or nitride microspheres coated with SiC in a SiC matrix. Studies have indicated that ZrC is a potential replacement for SiC on account of its higher melting point, increased fission product corrosion resistance and better chemical stability. The present work investigated natural uranium carbide microspheres in a ZrC matrix instead of SiC. Known issues of minor actinide volatility during traditional fabrication procedures necessitated the investigation of still high temperature but more rapid fabrication techniques to minimize these anticipated losses. In this regard, fabrication of ZrC matrix by combustion synthesis from zirconium and graphite powders was studied. Criteria were established to obtain sufficient matrix density with UC microsphere volume fractions up to 30%. Tests involving production of microspheres by spark erosion method (similar to electrodischarge machining) showed the inability of the method to produce UC microspheres in the desired range of 300 to 1200 mum. A rotating electrode device was developed using a minimum current of 80A and rotating at speeds up to 1500 rpm to fabricate microspheres between 355 and 1200 mum. Using the ZrC process knowledge, UC electrodes were fabricated and studied for use in the rotating electrode device to produce UC microspheres. Fabrication of the cercer composite form was studied using microsphere volume fractions of 10%, 20%, and 30%. The macrostructure of the composite and individual components at various stages were

  14. Development status and research directions on the structural components of the fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Ki Nam; Jeong, Yeon Ho; Kim, Hyung Kyu; Kang, Heung Seok; Yoon, Kyung Ho; Bang, Jae Keon

    1997-06-01

    Survey on the structural components of the state-of-the art of the PWR fuel assembly developed by various nuclear fuel vendors has been performed. As a result, some developmental directions and mechanical/structural basic technology to be established for these structural components have been drawn out. The developmental directions are as follows; The top end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function for easy reconstitution of the fuel assembly. The bottom end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function for easy reconstitution of the fuel assembly. The bottom end piece shall be designed in shape to reduce its height to accommodate the fuel rod growth for high burnup and to have a function of debris protection. The spacer grid shall be designed in shape to have a function of enhancing the thermal margin and maintaining the fuel rod integrity without fuel failure due to fuel rod fretting and vibration. The mechanical/structural basic technology which must be established is as follows; The stress analysis results shall comply with the stress criteria specified in the ASME code stress limits and the shape optimization technology shall be developed for the top/bottom end pieces. For the spacer grid cell, the nonlinear analysis model of the fuel rod and the analysis model on the flow-induced fuel rod vibration, and a study of the mechanism and a quantified model on the fuel rod fretting wear shall be developed. In addition, numerical analysis model to estimate the buckling strength of the spacer grid assembly shall be developed. Besides above technology, technology related the verification test should be developed. (author). 30 figs., 54 refs.

  15. Recycling as an option of used nuclear fuel management strategy

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, Tomaz, E-mail: tomaz.zagar@gen-energija.s [GEN energija, d.o.o., Cesta 4. julija 42, 8270 Krsko (Slovenia); Institute Jozef Stefan, Jamova 39, 1000 Ljubljana (Slovenia); Bursic, Ales; Spiler, Joze [GEN energija, d.o.o., Cesta 4. julija 42, 8270 Krsko (Slovenia); Kim, Dana; Chiguer, Mustapha; David, Gilles; Gillet, Philippe [AREVA, 33 rue La Fayette, 75009 Paris (France)

    2011-04-15

    The paper presents recycling as an option of used nuclear fuel management strategy with specific focus on the Slovenia. GEN energija is an independent supplier of integral and competitive electricity for Slovenia. In response to growing energy needs, GEN has conducted several feasibility and installation studies of a new nuclear power plant in Slovenia. With sustainable development, the environment, and public acceptance in mind, GEN conducted a study with AREVA concerning the options for the management of its' new plant's used nuclear fuel. After a brief reminder of global political and economic context, solutions for used nuclear fuel management using current technologies are presented in the study as well as an economic assessment of a closed nuclear fuel cycle. The paper evaluates and proposes practical solutions for mid-term issues on used nuclear fuel management strategies. Different scenarios for used nuclear fuel management are presented, where used nuclear fuel recycling (as MOX, for mixed oxide fuel, and ERU, for enriched reprocessed uranium) are considered. The study concludes that closing the nuclear fuel cycle will allow Slovenia to have a supplementary fuel supply for its new reactor via recycling, while reducing the radiotoxicity, thermal output, and volume of its wastes for final disposal, reducing uncertainties, gaining public acceptance, and allowing time for capitalization on investments for final disposal.

  16. Evaluation of thorium based nuclear fuel. Chemical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Konings, R.J.M.; Blankenvoorde, P.J.A.M.; Cordfunke, E.H.P.; Bakker, K.

    1995-07-01

    This report describes the chemical aspects of a thorium-based fuel cycle. It is part of a series devoted to the study of thorium-based fuel as a means to achieve a considerable reduction of the radiotoxicity of the waste from nuclear power production. Therefore special emphasis is placed on fuel (re-)fabrication and fuel reprocessing in the present work. (orig.).

  17. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Jubin, R. T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Strachan, D. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ilas, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Spencer, B. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Soelberg, N. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

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

  18. Accident source terms for boiling water reactors with high burnup cores.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

    2007-11-01

    The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

  19. Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project

    Energy Technology Data Exchange (ETDEWEB)

    Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-03-26

    This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.

  20. Accident source terms for pressurized water reactors with high-burnup cores calculated using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Powers, Dana Auburn; Ashbaugh, Scott G.; Leonard, Mark Thomas; Longmire, Pamela

    2010-04-01

    In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs2MoO4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

  1. International nuclear fuel cycle fact book. Revision 6

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-01-01

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

  2. Fabrication and Characterization of UN-USix Nuclear Fuel

    OpenAIRE

    Raftery, Alicia Marie

    2015-01-01

    In this thesis, UN-U3Si2 nuclear fuel was fabricated using spark plasma sintering and characterized to analyze the microstructure and crystal structure of the resulting pellets. This work was done in collaboration with accident tolerant fuel research, an effort which aims at developing nuclear fuel with superior safety and performance compared to currently used oxide fuels. Uranium silicide was manufactured by arc melting to produce U3Si2 and uranium mononitride was synthesized by using the h...

  3. FY14 Status Report: CIRFT Testing Results on High Burnup UNF

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

    The objective of this project is to perform a systematic study of SNF/UNF (spent nuclear fuel/or used nuclear fuel) integrity under simulated transportation environments by using hot cell testing technology developed recently at Oak Ridge National Laboratory (ORNL), CIRFT (Cyclic Integrated Reversible-Bending Fatigue Tester). Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmarking tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. With support from the US Department of Energy and the NRC, CIRFT testing has been continued. The CIRFT testing was conducted on three HBR rods (R3, R4, and R5), with two specimens failed and one specimen un-failed. The total number of cycles in the test of un-failed specimens went over 2.23 107; the test was stopped as because the specimen did not show any sign of failure. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of used fuel rods in terms of both the curvature amplitude and the maximum of absolute of curvature extremes. The latter is significant because the maxima of extremes signify the maximum of tensile stress of the outer fiber of the bending rod. So far, a large variety of hydrogen contents has been covered in the CIRFT testing on HBR rods. It has been shown that the load amplitude is the dominant factor that controls the lifetime of bending rods, but the hydrogen content also has an important effect on the lifetime attained, according to the load range tested.

  4. Technology readiness levels for advanced nuclear fuels and materials development

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W.J., E-mail: jon.carmack@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Braase, L.A.; Wigeland, R.A. [Idaho National Laboratory, Idaho Falls, ID (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States)

    2017-03-15

    Highlights: • Definition of nuclear fuels system technology readiness level. • Identification of evaluation criteria for nuclear fuel system TRLs. • Application of TRLs to fuel systems. - Abstract: The Technology Readiness process quantitatively assesses the maturity of a given technology. The National Aeronautics and Space Administration (NASA) pioneered the process in the 1980s to inform the development and deployment of new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications. It was also adopted by the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is needed to improve the performance and safety of current and advanced reactors, and ultimately close the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the assessment process to advanced fuel development is useful as a management, communication, and tracking tool. This article provides definition of technology readiness levels (TRLs) for nuclear fuel technology as well as selected examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).

  5. The nuclear fuel cycle; Le cycle du combustible nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-05-01

    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  6. Characterization of Hydrogen Content in ZIRCALOY-4 Nuclear Fuel Cladding

    Science.gov (United States)

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

    2010-02-01

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

  7. Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Warmann, Stephan A. [Portage, Inc., Idaho Falls, ID (United States); Rusch, Chris [NAC International, Inc., Norcross, GA (United States)

    2014-03-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying

  8. MMSNF 2005. Materials models and simulations for nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Freyss, M.; Durinck, J.; Carlot, G.; Sabathier, C.; Martin, P.; Garcia, P.; Ripert, M.; Blanpain, P.; Lippens, M.; Schut, H.; Federov, A.V.; Bakker, K.; Osaka, M.; Miwa, S.; Sato, I.; Tanaka, K.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Govers, K.; Verwerft, M.; Hou, M.; Lemehov, S.E.; Terentyev, D.; Govers, K.; Kotomin, E.A.; Ashley, N.J.; Grimes, R.W.; Van Uffelen, P.; Mastrikov, Y.; Zhukovskii, Y.; Rondinella, V.V.; Kurosaki, K.; Uno, M.; Yamanaka, S.; Minato, K.; Phillpot, S.; Watanabe, T.; Shukla, P.; Sinnott, S.; Nino, J.; Grimes, R.; Staicu, D.; Hiernaut, J.P.; Wiss, T.; Rondinella, V.V.; Ronchi, C.; Yakub, E.; Kaye, M.H.; Morrison, C.; Higgs, J.D.; Akbari, F.; Lewis, B.J.; Thompson, W.T.; Gueneau, C.; Gosse, S.; Chatain, S.; Dumas, J.C.; Sundman, B.; Dupin, N.; Konings, R.; Noel, H.; Veshchunov, M.; Dubourg, R.; Ozrin, C.V.; Veshchunov, M.S.; Welland, M.T.; Blanc, V.; Michel, B.; Ricaud, J.M.; Calabrese, R.; Vettraino, F.; Tverberg, T.; Kissane, M.; Tulenko, J.; Stan, M.; Ramirez, J.C.; Cristea, P.; Rachid, J.; Kotomin, E.; Ciriello, A.; Rondinella, V.V.; Staicu, D.; Wiss, T.; Konings, R.; Somers, J.; Killeen, J

    2006-07-01

    The MMSNF Workshop series aims at stimulating research and discussions on models and simulations of nuclear fuels and coupling the results into fuel performance codes.This edition was focused on materials science and engineering for fuel performance codes. The presentations were grouped in three technical sessions: fundamental modelling of fuel properties; integral fuel performance codes and their validation; collaborations and integration of activities. (A.L.B.)

  9. 78 FR 67348 - Invitation for Public Comment on Draft Test Plan for the High Burnup Dry Storage Cask Research...

    Science.gov (United States)

    2013-11-12

    ...: U.S. Department of Energy, C/O Melissa Bates, 1955 Freemont Ave., MS 1235, Idaho Falls, ID 83415..., 1955 Fremont Ave., Attn: Melissa Bates, Idaho Falls, ID, between 8 a.m. and 3:30 p.m. MT, Monday.... Melissa Bates, Contracting Officers Representative, High Burnup Dry Storage Cask Research and...

  10. Survey of nuclear fuel cycle economics: 1970--1985

    Energy Technology Data Exchange (ETDEWEB)

    Prince, B. E.; Peerenboom, J. P.; Delene, J. G.

    1977-03-01

    This report is intended to provide a coherent view of the diversity of factors that may affect nuclear fuel cycle economics through about 1985. The nuclear fuel cycle was surveyed as to past trends, current problems, and future considerations. Unit costs were projected for each step in the fuel cycle. Nuclear fuel accounting procedures were reviewed; methods of calculating fuel costs were examined; and application was made to Light Water Reactors (LWR) over the next decade. A method conforming to Federal Power Commission accounting procedures and used by utilities to account for backend fuel-cycle costs was described which assigns a zero net salvage value to discharged fuel. LWR fuel cycle costs of from 4 to 6 mills/kWhr (1976 dollars) were estimated for 1985. These are expected to reach 6 to 9 mills/kWr if the effect of inflation is included.

  11. Nuclear fuels: Development, processing and disposal

    Energy Technology Data Exchange (ETDEWEB)

    Allday, C.

    1982-08-01

    The successful development of the world's energy resources has enabled industries in the more advanced countries to provide the economic basis on which improved living standards are based. As the less well-developed countries seek to improve their standards of living the pressure on existing energy resources will increase. In this context it is essential not to allow the current industrial recession in the developed countries, with its associated apparent abundancy of coal, oil and gas, to mask the longer-term energy situation. It is not here proposed to discuss the role of nuclear power in the energy scene except to say that, with the continuing need to develop energy resources, nuclear as a proven safe and economic system - will have a vital role to fulfil in meeting the world's future energy demands. This paper is concerned with the development of nuclear fuel and the industry which has grown around it during the last 30 years. It shall concentrate on its development in this country and describe the history and activities of BNFL.

  12. Fuel Cycle Services the Heart of Nuclear Energy

    Directory of Open Access Journals (Sweden)

    S. Soentono

    2007-01-01

    Full Text Available Fuel is essential for development whether for survival and or wealth creation purposes. In this century the utilization of fuels need to be improved although energy mix is still to be the most rational choice. The large amount utilization of un-renewable fossil has some disadvantages since its low energy content requires massive extraction, transport, and processing while emitting CO2 resulting degradation of the environment. In the mean time the advancement of nuclear science and technology has improved significantly the performance of nuclear power plant, management of radioactive waste, enhancement of proliferation resistance, and more economic competitiveness. Ever since the last decade of the last century the nuclear renaissance has taken place. This is also due to the fact that nuclear energy does not emit GHG. Although the nuclear fuel offers a virtually limitless source of economic energy, it is only so if the nuclear fuel is reprocessed and recycled. Consequently, the fuel cycle is to be even more of paramount important in the future. The infrastructure of the fuel cycle services worldwide has been adequately available. Various International Initiatives to access the fuel cycle services are also offered. However, it is required to put in place the International Arrangements to guaranty secured sustainable supply of services and its peaceful use. Relevant international co-operations are central for proceeding with the utilization of nuclear energy, while this advantageous nuclear energy utilization relies on the fuel cycle services. It is therefore concluded that the fuel cycle services are the heart of nuclear energy, and the international nuclear community should work together to maintain the availability of this nuclear fuel cycle services timely, sufficiently, and economically.

  13. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  14. 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.

  15. 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.

  16. Coupon Surveillance For Corrosion Monitoring In Nuclear Fuel Basin

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I.; Murphy, T. R.; Deible, R.

    2012-10-01

    Aluminum and stainless steel coupons were put into a nuclear fuel basin to monitor the effect of water chemistry on the corrosion of fuel cladding. These coupons have been monitored for over ten years. The corrosion and pitting data is being used to model the kinetics and estimate the damage that is occurring to the fuel cladding.

  17. Grooved Fuel Rings for Nuclear Thermal Rocket Engines

    Science.gov (United States)

    Emrich, William

    2009-01-01

    An alternative design concept for nuclear thermal rocket engines for interplanetary spacecraft calls for the use of grooved-ring fuel elements. Beyond spacecraft rocket engines, this concept also has potential for the design of terrestrial and spacecraft nuclear electric-power plants. The grooved ring fuel design attempts to retain the best features of the particle bed fuel element while eliminating most of its design deficiencies. In the grooved ring design, the hydrogen propellant enters the fuel element in a manner similar to that of the Particle Bed Reactor (PBR) fuel element.

  18. World nuclear capacity and fuel cycle requirements, November 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-30

    This analysis report presents the current status and projections of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. Long-term projections of US nuclear capacity, generation, fuel cycle requirements, and spent fuel discharges for three different scenarios through 2030 are provided in support of the Department of Energy`s activities pertaining to the Nuclear Waste Policy Act of 1982 (as amended in 1987). The projections of uranium requirements also support the Energy Information Administration`s annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment.

  19. 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.

  20. Development of nuclear fuel cycle technologies - bases of long-term provision of fuel and environmental safety of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Solonin, M.I.; Polyakov, A.S.; Zakharkin, B.S.; Smelov, V.S.; Nenarokomov, E.A.; Mukhin, I.V. [SSC, RF, A.A. Bochvar ALL-Russia Research Institute of Inorganic Materials, Moscow (Russian Federation)

    2000-07-01

    To-day nuclear power is one of the options, however, to-morrow it may become the main source of the energy, thus, providing for the stable economic development for the long time to come. The availability of the large-scale nuclear power in the foreseeable future is governed by not only the safe operation of nuclear power plants (NPP) but also by the environmentally safe management of spent nuclear fuel, radioactive waste conditioning and long-term storage. More emphasis is to be placed to the closing of the fuel cycle in view of substantial quantities of spent nuclear fuel arisings. The once-through fuel cycle that is cost effective at the moment cannot be considered to be environmentally safe even for the middle term since the substantial build-up of spent nuclear fuel containing thousands of tons Pu will require the resolution of the safe management problem in the nearest future and is absolutely unjustified in terms of moral ethics as a transfer of the responsibility to future generations. The minimization of radioactive waste arisings and its radioactivity is only feasible with the closed fuel cycle put into practice and some actinides and long-lived fission radionuclides burnt out. The key issues in providing the environmentally safe fuel cycle are efficient processes of producing fuel for NPP, radionuclide after-burning included, a long-term spent nuclear fuel storage and reprocessing as well as radioactive waste management. The paper deals with the problems inherent in producing fuel for NPP with a view for the closed fuel cycle. Also discussed are options of the fuel cycle, its effectiveness and environmental safety with improvements in technologies of spent nuclear fuel reprocessing and long-lived radionuclide partitioning. (authors)

  1. Logistics of nuclear fuel production for nuclear submarines; Logistica de producao de combustiveis para submarinos nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Guimaraes, Leonam dos Santos [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: leosg@uol.com.br

    2000-07-01

    The future acquisition of nuclear attack submarines by Brazilian Navy along next century will imply new requirements on Naval Logistic Support System. These needs will impact all the six logistic functions. Among them, fuel supply could be considered as the one which requires the most important capacitating effort, including not only technological development of processes but also the development of a national industrial basis for effective production of nuclear fuel. This paper presents the technical aspects of the processes involved and an annual production dimensioning for an squadron composed by four units. (author)

  2. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The purpose of this Handbook is to provide guidance on how to calculate the characteristics of releases of radioactive materials and/or hazardous chemicals from nonreactor nuclear facilities. In addition, the Handbook provides guidance on how to calculate the consequences of those releases. There are four major chapters: Hazard Evaluation and Scenario Development; Source Term Determination; Transport Within Containment/Confinement; and Atmospheric Dispersion and Consequences Modeling. These chapters are supported by Appendices, including: a summary of chemical and nuclear information that contains descriptions of various fuel cycle facilities; details on how to calculate the characteristics of source terms for releases of hazardous chemicals; a comparison of NRC, EPA, and OSHA programs that address chemical safety; a summary of the performance of HEPA and other filters; and a discussion of uncertainties. Several sample problems are presented: a free-fall spill of powder, an explosion with radioactive release; a fire with radioactive release; filter failure; hydrogen fluoride release from a tankcar; a uranium hexafluoride cylinder rupture; a liquid spill in a vitrification plant; and a criticality incident. Finally, this Handbook includes a computer model, LPF No.1B, that is intended for use in calculating Leak Path Factors. A list of contributors to the Handbook is presented in Chapter 6. 39 figs., 35 tabs.

  3. Nuclear chemistry model of borated fuel crud

    Energy Technology Data Exchange (ETDEWEB)

    Sawicki, J.A. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    2002-07-01

    Fuel crud deposits on Callaway Cycle 9 once-burnt high-axial offset anomaly (AOA {approx} -15%) feed assemblies revealed a complex 4-phase matted-layered morphology of a new type that is uncommon in pressurized water reactors [1-3]. The up to 140-{open_square}m-thick crud flakes consisted predominantly of insoluble needle-like particles of Ni-Fe oxy-borate Ni{sub 2}FeBO{sub 5} (bonaccordite) and granular precipitates of m-ZrO{sub 2} (baddeleyite), along with nickel oxide NiO (bunsenite) and minor amount of nickel ferrite NiFe{sub 2}O{sub 4} (trevorite). Furthermore, boron in crud flakes showed that the concentration of {sup 10}B had depleted to 10.2{+-}0.2%, as compared to its 20% natural isotopic abundance and its 17% end-of-cycle abundance in bulk coolant. The form and depth distribution of Ni{sub 2}FeBO{sub 5} and m-ZrO{sub 2} precipitates, as well as substantial {sup 10}B burn-up, point to a strongly alkaline environment at the clad surface of the high-duty fuel rods. This paper extends a nuclear chemistry model of heavily borated fuel crud deposits. The paper shows that the local nuclear heat and lithium buildup from {sup 10}B(n,{open_square}){sup 7}Li reactions may help to create hydrothermal and chemical conditions within the crud layer in favor of Ni{sub 2}FeBO{sub 5} formation and a ZrO{sub 2} dissolution-reprecipitation mechanism. Consistent with the model, the hydrothermal formation of Ni{sub 2}FeBO{sub 5} needles was recently proved to be possible in laboratory tests with aqueous NiO-Fe{sub 2}O{sub 3}-H{sub 3}BO{sub 3}-LiOH slurries, at temperatures only slightly exceeding 400 C. (author)

  4. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

    2013-09-30

    This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

  5. 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.

  6. Modelling and modal properties of nuclear fuel assembly

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2011-12-01

    Full Text Available The paper deals with the modelling and modal analysis of the hexagonal type nuclear fuel assembly. This very complicated mechanical system is created from the many beam type components shaped into spacer grids. The cyclic and central symmetry of the fuel rod package and load-bearing skeleton is advantageous for the fuel assembly decomposition into six identical revolved fuel rod segments, centre tube and skeleton linked by several spacer grids in horizontal planes. The derived mathematical model is used for the modal analysis of the Russian TVSA-T fuel assembly and validated in terms of experimentally determined natural frequencies, modes and static deformations caused by lateral force and torsional couple of forces. The presented model is the first necessary step for modelling of the nuclear fuel assembly vibration caused by different sources of excitation during the nuclear reactor VVER type operation.

  7. Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-01

    A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

  8. Basic data for integrated assessment of nuclear fuel cycle system

    Energy Technology Data Exchange (ETDEWEB)

    Nomura, Yasushi; Tamaki, Hitoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ito, Chihiro; Saegusa, Toshiari [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2001-03-01

    In our country, where natural energy resources such as oil and coal are scarce, it is vital to establish a nuclear fuel cycle to reprocess spent fuel and reuse valuable nuclear fuel in electric power generation reactors. However spent fuel is now being accumulated too much so that, for the time being, it is necessary to establish a system for tentatively storing spent fuel. In this report, in order to deal with these issues, evaluation methods, which were developed, prepared and discussed by Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI), are rendered together with sample results of their application. Also reported is some important information on the data and methods for the safety assessment of nuclear fuel cycle facilities, which have been surveyed by JAERI and CRIEPI. (author)

  9. Handbook on process and chemistry on nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Atsuyuki (ed.) [Tokyo Univ., Tokyo (Japan); Asakura, Toshihide; Adachi, Takeo (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2001-12-01

    'Wet-type' nuclear fuel reprocessing technology, based on PUREX technology, has wide applicability as the principal reprocessing technology of the first generation, and relating technologies, waste management for example, are highly developed, too. It is quite important to establish a database summarizing fundamental information about the process and the chemistry of 'wet-type' reprocessing, because it contributes to establish and develop fuel reprocessing process and nuclear fuel cycle treating high burn-up UO{sub 2} fuel and spent MOX fuel, and to utilize 'wet-type' reprocessing technology much widely. This handbook summarizes the fundamental data on process and chemistry, which was collected and examined by 'Editing Committee of Handbook on Process and Chemistry of Nuclear Fuel Reprocessing', from FY 1993 until FY 2000. (author)

  10. 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

  11. Basic research on cermet nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hiroshi; Sto, Seichi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering; Takano, Masahide; Minato, Kazuo; Fukuda, Kosaku

    1998-01-01

    Production of cermet nuclear fuel having fine uranium dioxide (UO{sub 2}) particles dispersed in matrix metal requires basic property data on the compatibility of matrix metal with fission product compounds. It is thermodynamically suggested that, as burnup increases, cesium in oxide fuel reacts with the fuel, other fission products or cladding pipe and produces cesium uranates, cesium molybdate, or cesium chromate in stainless steel cladding pipe. Attempt was made to measure the thermal expansion coefficient and thermal conductivity of cesium uranates (Cs{sub 2}UO{sub 4} and Cs{sub 2}U{sub 2}O{sub 7}), cesium molybdate (Cs{sub 2}MoO{sub 4}) and cesium chromate (Cs{sub 2}CrO{sub 4}). Thermal expansion was measured by X-ray diffraction and determined by Cohen`s method. Thermal conductivity was obtained by measuring thermal diffusion by laser flash method. The thermal expansion of Cs{sub 2}UO{sub 4} and Cs{sub 2}U{sub 2}O{sub 7} is as low as 1.2% for the former and 1.0% for the latter, up to 1000K. The thermal expansion of Cs{sub 2}MoO{sub 4} is as high as that of Cs{sub 2}CrO{sub 4}, 2.1% for the former and 2.5% for the latter at temperatures from room temperature to 873K. Average thermal expansion in this temperature range is 4.4 x 10{sup -5} K{sup -1} for Cs{sub 2}MoO{sub 4} and 4.2 x 10{sup -5} K{sup -1}. The thermal expansion of Cs{sub 2}CrO{sub 4} is four times higher than that of UO{sub 2} and five times higher than that of Cr{sub 2}O{sub 3}. The thermal conductivity of Cs{sub 2}UO{sub 4} is nearly equal to that of Cs{sub 2}U{sub 2}O{sub 7} in absolute value and temperature dependency. Cs{sub 2}U{sub 2}O{sub 7}, having different thermal conductivity between {alpha} and {beta} phases, shows higher conductivity with {beta} than with {alpha}, about 1/4 of that of UO{sub 2} at 1000K. The thermal conductivity of Cs{sub 2}CrO{sub 4} is nearly equal to that of Cs{sub 2}MoO{sub 4} in absolute value and temperature dependency. (N.H.)

  12. Dynamic response of nuclear fuel assembly excited by pressure pulsations

    Directory of Open Access Journals (Sweden)

    Zeman V.

    2012-12-01

    Full Text Available The paper deals with dynamic load calculation of the hexagonal type nuclear fuel assembly caused by spatial motion of the support plates in the reactor core. The support plate motion is excited by pressure pulsations generated by main circulation pumps in the coolant loops of the primary circuit of the nuclear power plant. Slightly different pumps revolutions generate the beat vibrations which causes an amplification of fuel assembly component dynamic deformations and fuel rods coating abrasion. The cyclic and central symmetry of the fuel assembly makes it possible the system decomposition into six identical revolved fuel rod segments which are linked with central tube and skeleton by several spacer grids in horizontal planes.The modal synthesis method with condensation of the fuel rod segments is used for calculation of the normal and friction forces transmitted between fuel rods and spacer grids cells.

  13. Laser-based characterization of nuclear fuel plates

    Science.gov (United States)

    Smith, James A.; Cottle, Dave L.; Rabin, Barry H.

    2014-02-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  14. Laser-Based Characterization of Nuclear Fuel Plates

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; David L. Cottle; Barry H. Rabin

    2013-07-01

    Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

  15. Nuclear fuel alloys or mixtures and method of making thereof

    Science.gov (United States)

    Mariani, Robert Dominick; Porter, Douglas Lloyd

    2016-04-05

    Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

  16. Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    B.R. Westphal; G.L. Fredrickson; G.G. Galbreth; D. Vaden; M.D. Elliott; J.C. Price; E.M. Honeyfield; M.N. Patterson; L. A. Wurth

    2013-10-01

    Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.

  17. The high burn-up structure in nuclear fuel

    Directory of Open Access Journals (Sweden)

    Vincenzo V. Rondinella

    2010-12-01

    Full Text Available During its operating life in the core of a nuclear reactor nuclear fuel is subjected to significant restructuring processes determined by neutron irradiation directly through nuclear reactions and indirectly through the thermo-mechanical conditions established as a consequence of such reactions. In today's light water reactors, starting after ∼4 years of operation the cylindrical UO2 fuel pellet undergoes a transformation that affects its outermost radial region. The discovery of a newly forming structure necessitated the answering of important questions concerning the safety of extended fuel operation and still today poses the fascinating scientific challenge of fully understanding the microstructural mechanisms responsible for its formation.

  18. Migration behaviour of iodine in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hocking, W.H.; Verrall, R.A.; Muir, I.J

    2001-07-01

    A novel out-reactor method has been further developed for investigating the migration behaviour of fission products in UO{sub 2} nuclear fuel, which allows the effects of thermal diffusion. radiation damage and local segregation to be independently assessed. Tailored concentration profiles of any desired species are first created in the near-surface region of polished samples by ion implantation. The impact of either thermal annealing or simulated fission is then precisely determined by depth profiling with high-performance secondary ion mass spectrometry (SIMS). Comparison of iodine migration in U0{sub 2} wafers that had been ion-implanted to fluences spanning five orders of magnitude has revealed subtle radiation-damage effects and a pronounced concentration dependence for thermal diffusion. At concentrations above {approx}10{sup 16} atoms/cm{sup 3} much of the iodine became trapped, likely in microscopic bubbles. True thermal diffusion coefficients for iodine in polycrystalline U0{sub 2} have been derived by modelling the low-fluence data. (author)

  19. Experience of air transport of nuclear fuel material in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yamashita, T.; Toguri, D. [Transnuclear, LTD. (AREVA group), Tokyo (Japan); Kawasaki, M. [Japan Nuclear Cycle Development Inst., Muramatsu, Ibaraki (Japan)

    2004-07-01

    Certified Reference Materials (hereafter called as to CRMs), which are indispensable for Quality Assurance and Material Accountability in nuclear fuel plants, are being provided by overseas suppliers to Japanese nuclear entities as Type A package (non-fissile) through air transport. However, after the criticality accident at JCO in Japan, special law defining nuclear disaster countermeasures (hereafter called as to the LAW) has been newly enforced in June 2000. Thereafter, nuclear fuel materials must meet not only to the existing transport regulations but also to the LAW for its transport.

  20. 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.

  1. Modeling of Flow in Nuclear Reactor Fuel Cell Outlet

    Directory of Open Access Journals (Sweden)

    František URBAN

    2010-12-01

    Full Text Available Safe and effective load of nuclear reactor fuel cells demands qualitative and quantitative analysis of relations between coolant temperature in fuel cell outlet temperature measured by thermocouple and middle temperature of coolant in thermocouple plane position. In laboratory at Insitute of thermal power engineering of the Slovak University of Technology in Bratislava was installed an experimental physical fuel cell model of VVER 440 nuclear power plant with V 213 nuclear reactors. Objective of measurements on physical model was temperature and velocity profiles analysis in the fuel cell outlet. In this paper the measured temperature and velocity profiles are compared with the results of CFD simulation of fuel cell physical model coolant flow.

  2. Energy Return on Investment from Recycling Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    None

    2011-08-17

    This report presents an evaluation of the Energy Return on Investment (EROI) from recycling an initial batch of 800 t/y of used nuclear fuel (UNF) through a Recycle Center under a number of different fuel cycle scenarios. The study assumed that apart from the original 800 t of UNF only depleted uranium was available as a feed. Therefore for each subsequent scenario only fuel that was derived from the previous fuel cycle scenario was considered. The scenarios represent a good cross section of the options available and the results contained in this paper and associated appendices will allow for other fuel cycle options to be considered.

  3. Nuclear Fuel Test Rod Fabrication for Data Acquisition Test

    Energy Technology Data Exchange (ETDEWEB)

    Joung, Chang-Young; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A nuclear fuel test rod must be fabricated with precise welding and assembly technologies, and confirmed for their soundness. Recently, we have developed various kinds of processing systems such as an orbital TIG welding system, a fiber laser welding system, an automated drilling system and a helium leak analyzer, which are able to fabricate the nuclear fuel test rods and rigs, and keep inspection systems to confirm the soundness of the nuclear fuel test rods and rids. The orbital TIG welding system can be used with two kinds of welding methods. One can perform the round welding for end-caps of a nuclear fuel test rod by an orbital head mounted in a low-pressure chamber. The other can do spot welding for a pin-hole of a nuclear fuel test rod in a high-pressure chamber to fill up helium gas of high pressure. The fiber laser welding system can weld cylindrical and 3 axis samples such as parts of a nuclear fuel test rod and instrumentation sensors which is moved by an index chuck and a 3 axis (X, Y, Z) servo stage controlled by the CNC program. To measure the real-time temperature change at the center of the nuclear fuel during the irradiation test, a thermocouple should be instrumented at that position. Therefore, a hole needs to be made at the center of fuel pellet to instrument the thermocouple. An automated drilling system can drill a fine hole into a fuel pellet without changing tools or breaking the work-piece. The helium leak analyzer (ASM-380 model of DEIXEN Co.) can check the leak of the nuclear fuel test rod filled with helium gas. This paper describes not only the assembly and fabrication methods used by the process systems, but also the results of the data acquisition test for the nuclear fuel test rod. A nuclear fuel test rod for the data acquisition test was fabricated using the welding and assembling echnologies acquired from previous tests.

  4. 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)

  5. Bubble Effect in Heterogeneous Nuclear Fuel Solution System

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Xiao-ping; LUO; Huang-da; ZHANG; Wei; ZHU; Qing-fu

    2013-01-01

    Bubble effect means system reactivity changes due to the bubble induced solution volume,neutron leakage and absorption properties,neutron energy spectrum change in the nuclear fuel solution system.In the spent fuel dissolver,during uranium element shearing,the oxygen will be inlet to accelerate the

  6. Exploration for fossil and nuclear fuels from orbital altitudes

    Science.gov (United States)

    Short, N. M.

    1977-01-01

    The paper discusses the application of remotely sensed data from orbital satellites to the exploration for fossil and nuclear fuels. Geological applications of Landsat data are described including map editing, lithologic identification, structural geology, and mineral exploration. Specific results in fuel exploration are reviewed and a series of related Landsat images is included.

  7. Exploration for fossil and nuclear fuels from orbital altitudes

    Science.gov (United States)

    Short, N. M.

    1977-01-01

    The paper discusses the application of remotely sensed data from orbital satellites to the exploration for fossil and nuclear fuels. Geological applications of Landsat data are described including map editing, lithologic identification, structural geology, and mineral exploration. Specific results in fuel exploration are reviewed and a series of related Landsat images is included.

  8. 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.

  9. Dynamic Analysis of Nuclear Waste Generation Based on Nuclear Fuel Cycle Transition Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, S. R. [University of Science and Technology, Daejeon (Korea, Republic of); Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    According to the recommendations submitted by the Public Engagement Commission on Spent Nuclear Fuel Management (PECOS), the government was advised to pick the site for an underground laboratory and interim storage facilities before the end of 2020 followed by the related research for permanent and underground disposal of spent fuel after 10 years. In the middle of the main issues, the factors of environmentally friendly and safe way to handle nuclear waste are inextricable from nuclear power generating nation to ensure the sustainability of nuclear power. For this purposes, the closed nuclear fuel cycle has been developed regarding deep geological disposal, pyroprocessing, and burner type sodium-cooled fast reactors (SFRs) in Korea. Among two methods of an equilibrium model and a dynamic model generally used for screening nuclear fuel cycle system, the dynamic model is more appropriate to envisage country-specific environment with the transition phase in the long term and significant to estimate meaningful impacts based on the timedependent behavior of harmful wastes. This study aims at analyzing the spent nuclear fuel generation based on the long-term nuclear fuel cycle transition scenarios considered at up-to-date country specific conditions and comparing long term advantages of the developed nuclear fuel cycle option between once-through cycle and Pyro-SFR cycle. In this study, a dynamic analysis was carried out to estimate the long-term projection of nuclear electricity generation, installed capacity, spent nuclear fuel arising in different fuel cycle scenarios based on the up-to-date national energy plans.

  10. 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.

  11. Monitoring methods for nuclear fuel waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, R.B.; Barnard, J.W.; Bird, G.A. [and others

    1997-11-01

    This report examines a variety of monitoring activities that would likely be involved in a nuclear fuel waste disposal project, during the various stages of its implementation. These activities would include geosphere, environmental, vault performance, radiological, safeguards, security and community socioeconomic and health monitoring. Geosphere monitoring would begin in the siting stage and would continue at least until the closure stage. It would include monitoring of regional and local seismic activity, and monitoring of physical, chemical and microbiological properties of groundwater in rock and overburden around and in the vault. Environmental monitoring would also begin in the siting stage, focusing initially on baseline studies of plants, animals, soil and meteorology, and later concentrating on monitoring for changes from these benchmarks in subsequent stages. Sampling designs would be developed to detect changes in levels of contaminants in biota, water and air, soil and sediments at and around the disposal facility. Vault performance monitoring would include monitoring of stress and deformation in the rock hosting the disposal vault, with particular emphasis on fracture propagation and dilation in the zone of damaged rock surrounding excavations. A vault component test area would allow long-term observation of containers in an environment similar to the working vault, providing information on container corrosion mechanisms and rates, and the physical, chemical and thermal performance of the surrounding sealing materials and rock. During the operation stage, radiological monitoring would focus on protecting workers from radiation fields and loose contamination, which could be inhaled or ingested. Operational zones would be established to delineate specific hazards to workers, and movement of personnel and materials between zones would be monitored with radiation detectors. External exposures to radiation fields would be monitored with dosimeters worn by

  12. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

  13. MOX fuel arrangement for nuclear core

    Science.gov (United States)

    Kantrowitz, Mark L.; Rosenstein, Richard G.

    1998-01-01

    In order to use up a stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly types. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types are selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using urania fuel. The arrangement of the MOX rods and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly.

  14. LMFBR operation in the nuclear cycle without fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, S.I. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

    1997-12-01

    Substantiation is given to expediency of investigation of nuclear power (NP) development with fast reactors cooled by lead-bismuth alloy operating during extended time in the open nuclear fuel cycle with slightly enriched or depleted uranium make-up. 9 refs., 1 fig., 6 tabs.

  15. Electrochemical fluorination for processing of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2016-07-05

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  16. Galvanic cell for processing of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Diaz, Brenda L.; Martinez-Rodriguez, Michael J.; Gray, Joshua R.; Olson, Luke C.

    2017-02-07

    A galvanic cell and methods of using the galvanic cell is described for the recovery of uranium from used nuclear fuel according to an electrofluorination process. The galvanic cell requires no input energy and can utilize relatively benign gaseous fluorinating agents. Uranium can be recovered from used nuclear fuel in the form of gaseous uranium compound such as uranium hexafluoride, which can then be converted to metallic uranium or UO.sub.2 and processed according to known methodology to form a useful product, e.g., fuel pellets for use in a commercial energy production system.

  17. International nuclear fuel cycle fact book. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

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

    1984-03-01

    This Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids - international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  18. International Nuclear Fuel Cycle Fact Book. Revision 5

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-01-01

    This Fact Book has been compiled in an effort to provide: (1) an overview of worldwide nuclear power and fuel cycle programs; and (2) current data concerning fuel cycle and waste management facilities, R and D programs, and key personnel in countries other than the United States. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2. The Fact Book is organized as follows: (1) Overview section - summary tables which indicate national involvement in nuclear reactor, fuel cycle, and waste management development activities; (2) national summaries - a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; (3) international agencies - a section for each of the international agencies which has significant fuel cycle involvement; (4) energy supply and demand - summary tables, including nuclear power projections; (5) fuel cycle - summary tables; and (6) travel aids international dialing instructions, international standard time chart, passport and visa requirements, and currency exchange rate.

  19. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  20. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  1. ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    2017-06-19

    The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared to a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output

  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. Nuclear Fuel Design Technology Development for the Future Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Cheon, Jin Sik; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong; Lee, Byung Uk; Ko, Han Suk; So, Dong Sup; Koo, Dae Seo

    2006-04-15

    The test MOX fuels have been irradiated in the Halden reactor, and their burnup attained 40 GWd/t as of October 2005. The fuel temperature and internal pressure were measured by the sensors installed in the fuels and test rig. The COSMOS code, which was developed by KAERI, well predicted in-reactor behavior of MOX fuel. The COSMOS code was verified by OECD-NEA benchmarks, and the result confirmed the superiority of COSMOS code. MOX in-pile database (IFA-629.3, IFA-610.2 and 4) in Halden was also used for the verification of code. The COSMOS code was improved by introducing Graphic User Interface (GUI) and batch mode. The PCMI analysis module was developed and introduced by the new fission gas behavior model. The irradiation test performed under the arbitrary rod internal pressure could also be analyzed with the COSMOS code. Several presentations were made for the preparation to transfer MOX fuel performance analysis code to the industry, and the transfer of COSMOS code to the industry is being discussed. The user manual and COSMOS program (executive file) were provided for the industry to test the performance of COSMOS code. To envisage the direction of research, the MOX fuel research trend of foreign countries, specially focused on USA's GENP policy, was analyzed.

  4. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  5. 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.

  6. A cermet fuel reactor for nuclear thermal propulsion

    Science.gov (United States)

    Kruger, Gordon

    1991-01-01

    Work on the cermet fuel reactor done in the 1960's by General Electric (GE) and the Argonne National Laboratory (ANL) that had as its goal the development of systems that could be used for nuclear rocket propulsion as well as closed cycle propulsion system designs for ship propulsion, space nuclear propulsion, and other propulsion systems is reviewed. It is concluded that the work done in the 1960's has demonstrated that we can have excellent thermal and mechanical performance with cermet fuel. Thousands of hours of testing were performed on the cermet fuel at both GE and AGL, including very rapid transients and some radiation performance history. We conclude that there are no feasibility issues with cermet fuel. What is needed is reactivation of existing technology and qualification testing of a specific fuel form. We believe this can be done with a minimum development risk.

  7. A method for monitoring nuclear absorption coefficients of aviation fuels

    Science.gov (United States)

    Sprinkle, Danny R.; Shen, Chih-Ping

    1989-01-01

    A technique for monitoring variability in the nuclear absorption characteristics of aviation fuels has been developed. It is based on a highly collimated low energy gamma radiation source and a sodium iodide counter. The source and the counter assembly are separated by a geometrically well-defined test fuel cell. A computer program for determining the mass attenuation coefficient of the test fuel sample, based on the data acquired for a preset counting period, has been developed and tested on several types of aviation fuel.

  8. An analysis of international nuclear fuel supply options

    Science.gov (United States)

    Taylor, J'tia Patrice

    As the global demand for energy grows, many nations are considering developing or increasing nuclear capacity as a viable, long-term power source. To assess the possible expansion of nuclear power and the intricate relationships---which cover the range of economics, security, and material supply and demand---between established and aspirant nuclear generating entities requires models and system analysis tools that integrate all aspects of the nuclear enterprise. Computational tools and methods now exist across diverse research areas, such as operations research and nuclear engineering, to develop such a tool. This dissertation aims to develop methodologies and employ and expand on existing sources to develop a multipurpose tool to analyze international nuclear fuel supply options. The dissertation is comprised of two distinct components: the development of the Material, Economics, and Proliferation Assessment Tool (MEPAT), and analysis of fuel cycle scenarios using the tool. Development of MEPAT is aimed for unrestricted distribution and therefore uses publicly available and open-source codes in its development when possible. MEPAT is built using the Powersim Studio platform that is widely used in systems analysis. MEPAT development is divided into three modules focusing on: material movement; nonproliferation; and economics. The material movement module tracks material quantity in each process of the fuel cycle and in each nuclear program with respect to ownership, location and composition. The material movement module builds on techniques employed by fuel cycle models such as the Verifiable Fuel Cycle Simulation (VISION) code developed at the Idaho National Laboratory under the Advanced Fuel Cycle Initiative (AFCI) for the analysis of domestic fuel cycle. Material movement parameters such as lending and reactor preference, as well as fuel cycle parameters such as process times and material factors are user-specified through a Microsoft Excel(c) data spreadsheet

  9. 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.

  10. 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.

  11. 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.

  12. BISON Theory Manual The Equations behind Nuclear Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, R. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spencer, B. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stafford, D. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Perez, D. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Liu, W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact, and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.

  13. 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.

  14. Analysis of Proliferation Resistance of Nuclear Fuel Cycle Systems

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Hong Lae; Ko, Won Il; Kim, Ho Dong

    2009-11-15

    Proliferation resistance (PR) has been evaluated for the five nuclear fuel cycle systems, potentially deployable in Korea in the future, using the fourteen proliferation resistance attributes suggested in the TOPS report. Unidimensional Utility Theory (UUT) was used in the calculation of utility value for each of the fourteen proliferation resistance attributes, and Multi-Attribute Utility Theory (MAUT), a decision tool with multiple objectives, was used in the evaluation of the proliferation resistance of each nuclear fuel cycle system. Analytic Hierarchy Process (AHP) and Expert Elicitation (EE) were utilized in the derivation of weighting factors for the fourteen proliferation resistance attributes. Among the five nuclear fuel cycle systems evaluated, the once-through fuel cycle system showed the highest level of proliferation resistance, and Pyroprocessing-SFR fuel cycle system showed the similar level of proliferation resistance with the DUPIC fuel cycle system, which has two time higher level of proliferation resistance compared to that of the thermal MOX fuel cycle system. Sensitivity analysis was also carried out to make up for the uncertainty associated with the derivation of weighting factors for the fourteen proliferation resistance attributes.

  15. Material control in nuclear fuel fabrication facilities. Part I. Fuel descriptions and fabrication processes, P. O. 1236909 Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; Miller, C.L.

    1978-12-01

    The report presents information on foreign nuclear fuel fabrication facilities. Fuel descriptions and fuel fabrication information for three basic reactor types are presented: The information presented for LWRs assumes that Pu--U Mixed Oxide Fuel (MOX) will be used as fuel.

  16. International Nuclear Fuel Cycle Fact Book. Revision 12

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1992-05-01

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

  17. A study on the environmental friendliness of nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. J.; Lee, B. H.; Lee, S. Y.; Lim, C. Y.; Choi, Y. S.; Lee, Y. E.; Hong, D. S.; Cheong, J. H; Park, J. B.; Kim, K. K.; Cheong, H. Y; Song, M. C; Lee, H. J. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1998-01-01

    The purpose of this study is to develop methodologies for quantifying environmental and socio-political factors involved with nuclear fuel cycle and finally to evaluate nuclear fuel cycle options with special emphasis given to the factors. Moreover, methodologies for developing practical radiological health risk assessment code system will be developed by which the assessment could be achieved for the recycling and reuse of scrap materials containing residual radioactive contamination. Selected scenarios are direct disposal, DUPIC(Direct use of PWR spent fuel in CANDU), and MOX recycle, land use, radiological effect, and non-radiological effect were chosen for environmental criteria and public acceptance and non-proliferation of nuclear material for socio-political ones. As a result of this study, potential scenarios to be chosen in Korea were selected and methodologies were developed to quantify the environmental and socio-political criteria. 24 refs., 27 tabs., 29 figs. (author)

  18. International nuclear fuel cycle fact book. [Contains glossary

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.; Lakey, L.T.; Schneider, K.J.; Silviera, D.J.

    1987-01-01

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

  19. A study on the environmental friendliness of nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. J.; Lee, B. H.; Lee, S. Y.; Lim, C. Y.; Choi, Y. S.; Lee, Y. E.; Hong, D. S.; Cheong, J. H; Park, J. B.; Kim, K. K.; Cheong, H. Y; Song, M. C; Lee, H. J. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1998-01-01

    The purpose of this study is to develop methodologies for quantifying environmental and socio-political factors involved with nuclear fuel cycle and finally to evaluate nuclear fuel cycle options with special emphasis given to the factors. Moreover, methodologies for developing practical radiological health risk assessment code system will be developed by which the assessment could be achieved for the recycling and reuse of scrap materials containing residual radioactive contamination. Selected scenarios are direct disposal, DUPIC(Direct use of PWR spent fuel in CANDU), and MOX recycle, land use, radiological effect, and non-radiological effect were chosen for environmental criteria and public acceptance and non-proliferation of nuclear material for socio-political ones. As a result of this study, potential scenarios to be chosen in Korea were selected and methodologies were developed to quantify the environmental and socio-political criteria. 24 refs., 27 tabs., 29 figs. (author)

  20. International nuclear fuel cycle fact book: Revision 9

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.

    1989-01-01

    The International Nuclear Fuel Cycle Fact Book has been compiled in an effort to provide current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. The Fact Book contains: national summaries in which a section for each country which summarizes nuclear policy, describes organizational relationships and provides addresses, names of key personnel, and facilities information; and international agencies in which a section for each of the international agencies which has significant fuel cycle involvement, and a listing of nuclear societies. The national summaries, in addition to the data described above, feature a small map for each country as well as some general information. The latter is presented from the perspective of the Fact Book user in the United States.

  1. Temperature measuring analysis of the nuclear reactor fuel assembly

    Science.gov (United States)

    F., Urban; Ľ., Kučák; Bereznai, J.; Závodný, Z.; Muškát, P.

    2014-08-01

    Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuel assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.

  2. Nuclear fuel tax in court; Kernbrennstoffsteuer vor Gericht

    Energy Technology Data Exchange (ETDEWEB)

    Leidinger, Tobias [Gleiss Lutz Rechtsanwaelte, Duesseldorf (Germany)

    2014-07-15

    Besides the 'Nuclear Energy Moratorium' (temporary shutdown of eight nuclear power plants after the Fukushima incident) and the legally decreed 'Nuclear Energy Phase-Out' (by the 13th AtG-amendment), also the legality of the nuclear fuel tax is being challenged in court. After receiving urgent legal proposals from 5 nuclear power plant operators, the Hamburg fiscal court (4V 154/13) temporarily obliged on 14 April 2014 respective main customs offices through 27 decisions to reimburse 2.2 b. Euro nuclear fuel tax to the operating companies. In all respects a remarkable process. It is not in favour of cleverness to impose a political target even accepting immense constitutional and union law risks. Taxation 'at any price' is neither a statement of state sovereignty nor one for a sound fiscal policy. Early and serious warnings of constitutional experts and specialists in the field of tax law with regard to the nuclear fuel tax were not lacking. (orig.)

  3. Modeling Deep Burn TRISO particle nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, T.M., E-mail: besmanntm@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Stoller, R.E., E-mail: stollerre@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Samolyuk, G., E-mail: samolyukgd@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Schuck, P.C., E-mail: schuckpc@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Golubov, S.I., E-mail: golubovsi@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Rudin, S.P., E-mail: srudin@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Wills, J.M., E-mail: jxw@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Coe, J.D., E-mail: jcoe@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Wirth, B.D., E-mail: bdwirth@utk.edu [University of Tennessee, Knoxville, TN 37996-0750 (United States); Kim, S., E-mail: sungtae@cae.wisc.edu [University of Wisconsin, 1509 University Ave., Madison, WI 53706 (United States); Morgan, D.D., E-mail: ddmorgan@engr.wisc.edu [University of Wisconsin, 1509 University Ave., Madison, WI 53706 (United States); Szlufarska, I., E-mail: izabela@engr.wisc.edu [University of Wisconsin, 1509 University Ave., Madison, WI 53706 (United States)

    2012-11-15

    Under the DOE Deep Burn program TRISO fuel is being investigated as a fuel form for consuming plutonium and minor actinides, and for greater efficiency in uranium utilization. The result will thus be to drive TRISO particulate fuel to very high burn-ups. In the current effort the various phenomena in the TRISO particle are being modeled using a variety of techniques. The chemical behavior is being treated utilizing thermochemical analysis to identify phase formation/transformation and chemical activities in the particle, including kernel migration. Density functional theory is being used to understand fission product diffusion within the plutonia oxide kernel, the fission product's attack on the SiC coating layer, as well as fission product diffusion through an alternative coating layer, ZrC. Finally, a multiscale approach is being used to understand thermal transport, including the effect of radiation damage induced defects, in a model SiC material.

  4. Double-clad nuclear-fuel safety rod

    Science.gov (United States)

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

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

  5. 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.

  6. 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.

  7. Signatures of Extended Storage of Used Nuclear Fuel in Casks

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-28

    As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each casks neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

  8. Signatures of Extended Storage of Used Nuclear Fuel in Casks

    Energy Technology Data Exchange (ETDEWEB)

    Rauch, Eric Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-28

    As the amount of used nuclear fuel continues to grow, more and more used nuclear fuel will be transferred to storage casks. A consolidated storage facility is currently in the planning stages for storing these casks, where at least 10,000 MTHM of fuel will be stored. This site will have potentially thousands of casks once it is operational. A facility this large presents new safeguards and nuclear material accounting concerns. A new signature based on the distribution of neutron sources and multiplication within casks was part of the Department of Energy Office of Nuclear Energy’s Material Protection, Account and Control Technologies (MPACT) campaign. Under this project we looked at fingerprinting each cask's neutron signature. Each cask has a unique set of fuel, with a unique spread of initial enrichment, burnup, cooling time, and power history. The unique set of fuel creates a unique signature of neutron intensity based on the arrangement of the assemblies. The unique arrangement of neutron sources and multiplication produces a reliable and unique identification of the cask that has been shown to be relatively constant over long time periods. The work presented here could be used to restore from a loss of continuity of knowledge at the storage site. This presentation will show the steps used to simulate and form this signature from the start of the effort through its conclusion in September 2016.

  9. 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.

  10. Synergistic smart fuel for in-pile nuclear reactor measurements

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.A.; Kotter, D.K. [Idaho National Laboratories, Idaho Falls (United States); Ali, R.A.; Garrett, S.L. [Penn State University, University Park, State College, PA 16801 (United States)

    2013-07-01

    The thermo-acoustic fuel rod sensor developed in this research has demonstrated a novel technique for monitoring the temperature within the core of a nuclear reactor or the temperature of the surrounding heat-transfer fluid. It uses the heat from the nuclear fuel to generate sustained acoustic oscillations whose frequency will be indicative of the temperature. Converting a nuclear fuel rod into this type of thermo-acoustic sensor simply requires the insertion of a porous material (stack). This sensor has demonstrated a synergy with the elevated temperatures that exist within the nuclear reactor using materials that have only minimal susceptibility to high-energy particle fluxes. When the sensor is in operation, the sound waves radiated from the fuel rod resonator will propagate through the surrounding cooling fluid. The frequency of these oscillations is directly correlated with an effective temperature within the fuel rod resonator. This device is self-powered and is operational even in case of total loss of power of the reactor.

  11. HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER

    Energy Technology Data Exchange (ETDEWEB)

    BROWN,LC; BESENBRUCH,GE; LENTSCH,RD; SCHULTZ,KR; FUNK,JF; PICKARD,PS; MARSHALL,AC; SHOWALTER,SK

    2003-06-01

    OAK B202 HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER. Combustion of fossil fuels, used to power transportation, generate electricity, heat homes and fuel industry provides 86% of the world's energy. Drawbacks to fossil fuel utilization include limited supply, pollution, and carbon dioxide emissions. Carbon dioxide emissions, thought to be responsible for global warming, are now the subject of international treaties. Together, these drawbacks argue for the replacement of fossil fuels with a less-polluting potentially renewable primary energy such as nuclear energy. Conventional nuclear plants readily generate electric power but fossil fuels are firmly entrenched in the transportation sector. Hydrogen is an environmentally attractive transportation fuel that has the potential to displace fossil fuels. Hydrogen will be particularly advantageous when coupled with fuel cells. Fuel cells have higher efficiency than conventional battery/internal combustion engine combinations and do not produce nitrogen oxides during low-temperature operation. Contemporary hydrogen production is primarily based on fossil fuels and most specifically on natural gas. When hydrogen is produced using energy derived from fossil fuels, there is little or no environmental advantage. There is currently no large scale, cost-effective, environmentally attractive hydrogen production process available for commercialization, nor has such a process been identified. The objective of this work is to find an economically feasible process for the production of hydrogen, by nuclear means, using an advanced high-temperature nuclear reactor as the primary energy source. Hydrogen production by thermochemical water-splitting (Appendix A), a chemical process that accomplishes the decomposition of water into hydrogen and oxygen using only heat or, in the case of a hybrid thermochemical process, by a combination of heat and electrolysis, could meet these goals. Hydrogen produced from

  12. Sensitivity Analysis of Criticality for Different Nuclear Fuel Shapes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Hyun Sik; Jang, Misuk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    Rod-type nuclear fuel was mainly developed in the past, but recent study has been extended to plate-type nuclear fuel. Therefore, this paper reviews the sensitivity of criticality according to different shapes of nuclear fuel types. Criticality analysis was performed using MCNP5. MCNP5 is well-known Monte Carlo codes for criticality analysis and a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron or coupled neutron / photon / electron transport, including the capability to calculate eigenvalues for critical systems. We performed the sensitivity analysis of criticality for different fuel shapes. In sensitivity analysis for simple fuel shapes, the criticality is proportional to the surface area. But for fuel Assembly types, it is not proportional to the surface area. In sensitivity analysis for intervals between plates, the criticality is greater as the interval increases, but if the interval is greater than 8mm, it showed an opposite trend that the criticality decrease by a larger interval. As a result, it has failed to obtain the logical content to be described in common for all cases. The sensitivity analysis of Criticality would be always required whenever subject to be analyzed is changed.

  13. Technology Insights and Perspectives for Nuclear Fuel Cycle Concepts

    Energy Technology Data Exchange (ETDEWEB)

    S. Bays; S. Piet; N. Soelberg; M. Lineberry; B. Dixon

    2010-09-01

    The following report provides a rich resource of information for exploring fuel cycle characteristics. The most noteworthy trends can be traced back to the utilization efficiency of natural uranium resources. By definition, complete uranium utilization occurs only when all of the natural uranium resource can be introduced into the nuclear reactor long enough for all of it to undergo fission. Achieving near complete uranium utilization requires technologies that can achieve full recycle or at least nearly full recycle of the initial natural uranium consumed from the Earth. Greater than 99% of all natural uranium is fertile, and thus is not conducive to fission. This fact requires the fuel cycle to convert large quantities of non-fissile material into fissile transuranics. Step increases in waste benefits are closely related to the step increase in uranium utilization going from non-breeding fuel cycles to breeding fuel cycles. The amount of mass requiring a disposal path is tightly coupled to the quantity of actinides in the waste stream. Complete uranium utilization by definition means that zero (practically, near zero) actinide mass is present in the waste stream. Therefore, fuel cycles with complete (uranium and transuranic) recycle discharge predominately fission products with some actinide process losses. Fuel cycles without complete recycle discharge a much more massive waste stream because only a fraction of the initial actinide mass is burned prior to disposal. In a nuclear growth scenario, the relevant acceptable frequency for core damage events in nuclear reactors is inversely proportional to the number of reactors deployed in a fuel cycle. For ten times the reactors in a fleet, it should be expected that the fleet-average core damage frequency be decreased by a factor of ten. The relevant proliferation resistance of a fuel cycle system is enhanced with: decreasing reliance on domestic fuel cycle services, decreasing adaptability for technology misuse

  14. Fabrication of nuclear fuel assemblies in Mexico; Fabricacion de ensambles de combustible nuclear en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Medrano B, A. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: amb@nuclear.inin.mx

    2007-07-01

    In the Pilot Production Plant of Nuclear Fuel facilities (PPFCN) located in the Nuclear Center of Mexico; its were processed approximately 1000 Kg of powder of uranium dioxide with 11 different enrichments from 0.71 up to 3.95% U-235, the pellets were encapsulated in Zircaloy tubes and armed around 300 rods of nuclear fuel for to manufacture four assembles of nuclear fuel and a DUMMY for the qualification of processes, personnel and equipment. The project beginning in 1990 with the one agreement among General Electric, Federal Commission of Electricity (CFE) and the National Institute of Nuclear Research (ININ), after building the PPFCN, to install equipment, to design the parameters of production and to qualify us as suppliers of nuclear fuel; it was begins in 1994 the production of four GE9B assemblies that surrendered to the CNLV in May, 1996. In 1998 its were loaded in the unit 1 of the CNLV, assemble them of nuclear fuel with serial numbers INI002, INI003, INI004 and INI005 with an average enrichment of 3.03% U-235, four complete operational cycles worked including the central control cell. During the works of the ninth recharge of the unit 1 of the CNLV, September 20, 2002 were removed these assemblies from the reactor core reaching a burnt of 35313 MWD/TMU. (Author)

  15. Safety research in nuclear fuel cycle at PNC

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This report collects the results of safety research in nuclear fuel cycle at Power Reactor and Nuclear Fuel Development Corporation, in order to answer to the Questionnaire of OECD/NEA. The Questionnaire request to include information concerning to research topic, description, main results (if available), reference documents, research institutes involved, sponsoring organization and other pertinent information about followings: a) Recently completed research projects. b) Ongoing (current) research projects. Achievements on following items are omitted by the request of OECD/NEA, uranium mining and milling, uranium refining and conversion to UF{sub 6}, uranium enrichment, fuel manufacturers, spent fuel storage, radioactive waste management, transport of radioactive materials, decommissioning. We select topics from the fields of a) nuclear installation, b) seismic, and c) PSA, in projects from frame of annual safety research plan for nuclear installations established by Nuclear Safety Commission. We apply for the above a) and b) projects as follows: a) Achievements in Safety Research, fiscal 1991-1995, b) fiscal 1996 Safety Research Achievements: Progress. (author)

  16. Safety research in nuclear fuel cycle at PNC

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This report collects the results of safety research in nuclear fuel cycle at Power Reactor and Nuclear Fuel Development Corporation, in order to answer to the Questionnaire of OECD/NEA. The Questionnaire request to include information concerning to research topic, description, main results (if available), reference documents, research institutes involved, sponsoring organization and other pertinent information about followings: a) Recently completed research projects. b) Ongoing (current) research projects. Achievements on following items are omitted by the request of OECD/NEA, uranium mining and milling, uranium refining and conversion to UF{sub 6}, uranium enrichment, fuel manufacturers, spent fuel storage, radioactive waste management, transport of radioactive materials, decommissioning. We select topics from the fields of a) nuclear installation, b) seismic, and c) PSA, in projects from frame of annual safety research plan for nuclear installations established by Nuclear Safety Commission. We apply for the above a) and b) projects as follows: a) Achievements in Safety Research, fiscal 1991-1995, b) fiscal 1996 Safety Research Achievements: Progress. (author)

  17. 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.

  18. Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

    Science.gov (United States)

    Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

    2015-12-01

    The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.

  19. Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Yang Zhong; Robert C. O' Brien; Steven D. Howe; Nathan D. Jerred; Kristopher Schwinn; Laura Sudderth; Joshua Hundley

    2011-11-01

    The feasibility of the fabrication of tungsten based nuclear fuel cermets via Spark Plasma Sintering (SPS) is investigated in this work. CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar properties of these materials. This study shows that after a short time sintering, greater than 90 % density can be achieved, which is suitable to possess good strength as well as the ability to contain fission products. The mechanical properties and the densities of the samples are also investigated as functions of the applied pressures during the sintering.

  20. Use of silicide fuel in the Ford Nuclear Reactor - to lengthen fuel element lifetimes

    Energy Technology Data Exchange (ETDEWEB)

    Bretscher, M.M.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Burn, R.R.; Lee, J.C. [Univ. of Michigan, Ann Arbor, MI (United States). Phoenix Memorial Lab.

    1995-12-31

    Based on economic considerations, it has been proposed to increase the lifetime of LEU fuel elements in the Ford Nuclear Reactor by raising the {sup 235}U plate loading from 9.3 grams in aluminide (UAl{sub x}) fuel to 12.5 grams in silicide (U{sub 3}Si{sub 2}) fuel. For a representative core configuration, preliminary neutronic depletion and steady state thermal hydraulic calculations have been performed to investigate core characteristics during the transition from an all-aluminide to an all-silicide core. This paper discusses motivations for this fuel element upgrade, results from the calculations, and conclusions.

  1. Characterization of Nuclear Fuel using Multivariate Statistical Analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-27

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

  2. Nuclear Fuel Cycle Reasoner: PNNL FY13 Report

    Energy Technology Data Exchange (ETDEWEB)

    Hohimer, Ryan E.; Strasburg, Jana D.

    2013-09-30

    In Fiscal Year 2012 (FY12) PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In Fiscal Year 2013 (FY13) the SNAP demonstration was enhanced with respect to query and navigation usability issues.

  3. Evaluation of thorium based nuclear fuel. Extended summary

    Energy Technology Data Exchange (ETDEWEB)

    Franken, W.M.P.; Bultman, J.H.; Konings, R.J.M.; Wichers, V.A.

    1995-04-01

    Application of thorium based nuclear fuels has been evaluated with emphasis on possible reduction of the actinide waste. As a result three ECN-reports are published, discussing in detail: - The reactor physics aspects, by comparing the operation characteristics of the cores of Pressurized Water Reactors and Heavy Water Reactors with different fuel types, including equilibrium thorium/uranium free, once-through uranium fuel and equilibrium uranium/plutonium fuel, - the chemical aspects of thorium based fuel cycles with emphasis on fuel (re)fabrication and fuel reprocessing, - the possible reduction in actinide waste as analysed for Heavy Water Reactors with various types of thorium based fuels in once-through operation and with reprocessing. These results are summarized in this report together with a short discussion on non-proliferation and uranium resource utilization. It has been concluded that a substantial reduction of actinide radiotoxicity of the disposed waste may be achieved by using thorium based fuels, if very efficient partitioning and multiple recycling of uranium and thorium can be realized. This will, however, require large efforts to develop the technology to the necessary industrial scale of operation. (orig.).

  4. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

    The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

  5. Restriction of Civilian Nuclear Fuel Cycle and Effectiveness of Nuclear Nonproliferation

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, JaeSoo; Lee, HanMyung; Ko, HanSuk; Yang, MaengHo; Oh, KunBae [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    Many efforts have been made to prevent the spread of nuclear weapons since the nuclear era. Recent revelation such as Dr. A.Q. Khan Network showed that some states had acquired sensitive nuclear technologies including uranium enrichment which could be used for making nuclear weapons. In addition, with the advancement of industrial technology, it has become easier to have access to those technologies. In this context, proliferation risks are being increased more and more. As a result, various proposals to respond to proliferation risks by sensitive technologies have been made: Multilateral Nuclear Approaches (MNAs) by IAEA Director General El Baradei, non-transfer of sensitive nuclear technologies by the U.S. President George W. Bush, international center for nuclear fuel cycle service by Russian President Vladimir V. Putin, Global Nuclear Energy Partnership (GNEP) by Bush's administration and a concept for a multilateral mechanism for reliable access to nuclear fuel by 6 member states of the IAEA. Theses proposals all share the idea that the best way to reduce risk is to prevent certain states from having control over an indigenous civilian fuel cycle while still finding ways to confer the benefits of nuclear energy, and seem to imply that the current nonproliferation regime is fundamentally flawed and needs to be altered. However, these proposals are a center of controversy because they can restrict the inalienable right for the peaceful purposes of nuclear energy inscribed in Article IV of the NPT. Therefore, this paper analyzes the key challenges of these proposals and effectiveness of the goal of nuclear nonproliferation in practical term by restricting civilian nuclear fuel cycle.

  6. India's nuclear fuel cycle unraveling the impact of the U.S.-India nuclear accord

    CERN Document Server

    Woddi, Taraknath VK

    2009-01-01

    An analysis of the current (February 2009) status and future potential of India's nuclear fuel cycle is presented in this book. Such a fuel cycle assessment is important, but relatively opaque because India regards various aspects of its nuclear fuel cycle as strategically sensitive. Any study therefore necessarily depends upon reverse calculations based on the information that is available, expert assessments, engineering judgment and anecdotal information. In this work every effort is made to provide transparency to these foundations, so that changes can be made in light of alternative expec

  7. DIONISIO 2.0: New version of the code for simulating a whole nuclear fuel rod under extended irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Soba, Alejandro, E-mail: soba@cnea.gov.ar; Denis, Alicia

    2015-10-15

    Highlights: • A new version of the DIONISIO code is developed. • DIONISIO is devoted to simulating the behavior of a nuclear fuel rod in operation. • The formerly two-dimensional simulation of a pellet-cladding segment is now extended to the whole rod length. • An acceptable and more realistic agreement with experimental data is obtained. • The prediction range of our code is extended up to average burnup of 60 MWd/kgU. - Abstract: The version 2.0 of the DIONISIO code, that incorporates diverse new aspects, has been recently developed. One of them is referred to the code architecture that allows taking into account the axial variation of the conditions external to the rod. With this purpose, the rod is divided into a number of axial segments. In each one the program considers the system formed by a pellet and the corresponding cladding portion and solves the numerous phenomena that take place under the local conditions of linear power and coolant temperature, which are given as input parameters. To do this a bi-dimensional domain in the r–z plane is considered where cylindrical symmetry and also symmetry with respect to the pellet mid-plane are assumed. The results obtained for this representative system are assumed valid for the complete segment. The program thus produces in each rod section the values of the temperature, stress, strain, among others as outputs, as functions of the local coordinates r and z. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated. Moreover, new calculation tools designed to extend the application range of the code to high burnup, which were reported elsewhere, have also been incorporated to DIONISIO 2.0 in recent times. With these improvements, the code results are compared with some 33 experiments compiled in the IFPE data base, that cover more than 380 fuel rods irradiated up to average burnup levels of 40–60 MWd/kgU. The results of these

  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. Modeling Deep Burn TRISO Particle Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Besmann, Theodore M [ORNL; Stoller, Roger E [ORNL; Samolyuk, German D [ORNL; Schuck, Paul C [ORNL; Rudin, Sven [Los Alamos National Laboratory (LANL); Wills, John [Los Alamos National Laboratory (LANL); Wirth, Brian D. [University of California, Berkeley; Kim, Sungtae [University of Wisconsin, Madison; Morgan, Dane [University of Wisconsin, Madison; Szlufarska, Izabela [University of Wisconsin, Madison

    2012-01-01

    Under the DOE Deep Burn program TRISO fuel is being investigated as a fuel form for consuming plutonium and minor actinides, and for greater efficiency in uranium utilization. The result will thus be to drive TRISO particulate fuel to very high burn-ups. In the current effort the various phenomena in the TRISO particle are being modeled using a variety of techniques. The chemical behavior is being treated utilizing thermochemical analysis to identify phase formation/transformation and chemical activities in the particle, including kernel migration. First principles calculations are being used to investigate the critical issue of fission product palladium attack on the SiC coating layer. Density functional theory is being used to understand fission product diffusion within the plutonia oxide kernel. Kinetic Monte Carlo techniques are shedding light on transport of fission products, most notably silver, through the carbon and SiC coating layers. The diffusion of fission products through an alternative coating layer, ZrC, is being assessed via DFT methods. Finally, a multiscale approach is being used to understand thermal transport, including the effect of radiation damage induced defects, in a model SiC material.

  10. CLAD CARBIDE NUCLEAR FUEL, THERMIONIC POWER, MODULES.

    Science.gov (United States)

    The general objective is to evaluate a clad carbide emitter, thermionic power module which simulates nuclear reactor installation, design, and...performance. The module is an assembly of two series-connected converters with a single common cesium reservoir. The program goal is 500 hours

  11. Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

    Energy Technology Data Exchange (ETDEWEB)

    Ketusky, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-09

    The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

  12. Nuclear Fuel Cycle Reasoner: PNNL FY12 Report

    Energy Technology Data Exchange (ETDEWEB)

    Hohimer, Ryan E.; Pomiak, Yekaterina G.; Neorr, Peter A.; Gastelum, Zoe N.; Strasburg, Jana D.

    2013-05-03

    Building on previous internal investments and leveraging ongoing advancements in semantic technologies, PNNL implemented a formal reasoning framework and applied it to a specific challenge in nuclear nonproliferation. The Semantic Nonproliferation Analysis Platform (SNAP) was developed as a preliminary graphical user interface to demonstrate the potential power of the underlying semantic technologies to analyze and explore facts and relationships relating to the nuclear fuel cycle (NFC). In developing this proof of concept prototype, the utility and relevancy of semantic technologies to the Office of Defense Nuclear Nonproliferation Research and Development (DNN R&D) has been better understood.

  13. 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...

  14. Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Weber, William J. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering; Zhang, Yanwen [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Materials Science and Engineering

    2016-09-20

    This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effects of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.

  15. Nuclear characteristics of Pu fueled LWR and cross section sensitivities

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ., Suita (Japan). Faculty of Engineering

    1998-03-01

    The present status of Pu utilization to thermal reactors in Japan, nuclear characteristics and topics and cross section sensitivities for analysis of Pu fueled thermal reactors are described. As topics we will discuss the spatial self-shielding effect on the Doppler reactivity effect and the cross section sensitivities with the JENDL-3.1 and 3.2 libraries. (author)

  16. 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.

  17. Method of controlling crystallite size in nuclear-reactor fuels

    Science.gov (United States)

    Lloyd, M.H.; Collins, J.L.; Shell, S.E.

    Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

  18. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    Science.gov (United States)

    2010-03-05

    with uranium to make mixed-oxide ( MOX ) fuel, in which the 239Pu largely substitutes for 235U. Two French reprocessing plants at La Hague can each...and France also have older plants to reprocess gas-cooled reactor fuel, and India has a 275-ton plant.53 About 200 metric tons of MOX fuel is used...to make MOX fuel for today’s nuclear power plants are modest. Existing commercial light water reactors use ordinary water to slow down, or “moderate

  19. HEISHI: A fuel performance model for space nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Young, M.F.

    1994-08-01

    HEISHI is a Fortran computer model designed to aid in analysis, prediction, and optimization of fuel characteristics for use in Space Nuclear Thermal Propulsion (SNTP). Calculational results include fission product release rate, fuel failure fraction, mode of fuel failure, stress-strain state, and fuel material morphology. HEISHI contains models for decay chain calculations of retained and released fission products, based on an input power history and release coefficients. Decay chain parameters such as direct fission yield, decay rates, and branching fractions are obtained from a database. HEISHI also contains models for stress-strain behavior of multilayered fuel particles with creep and differential thermal expansion effects, transient particle temperature profile, grain growth, and fuel particle failure fraction. Grain growth is treated as a function of temperature; the failure fraction depends on the coating tensile strength, which in turn is a function of grain size. The HEISHI code is intended for use in analysis of coated fuel particles for use in particle bed reactors; however, much of the code is geometry-independent and applicable to fuel geometries other than spherical.

  20. 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.

  1. 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.

  2. Modeling and Simulation of a Nuclear Fuel Element Test Section

    Science.gov (United States)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  3. Long-term global nuclear energy and fuel cycle strategies

    Energy Technology Data Exchange (ETDEWEB)

    Krakowski, R.A. [Los Alamos National Lab., NM (United States). Technology and Safety Assessment Div.

    1997-09-24

    The Global Nuclear Vision Project is examining, using scenario building techniques, a range of long-term nuclear energy futures. The exploration and assessment of optimal nuclear fuel-cycle and material strategies is an essential element of the study. To this end, an established global E{sup 3} (energy/economics/environmental) model has been adopted and modified with a simplified, but comprehensive and multi-regional, nuclear energy module. Consistent nuclear energy scenarios are constructed using this multi-regional E{sup 3} model, wherein future demands for nuclear power are projected in price competition with other energy sources under a wide range of long-term demographic (population, workforce size and productivity), economic (price-, population-, and income-determined demand for energy services, price- and population-modified GNP, resource depletion, world-market fossil energy prices), policy (taxes, tariffs, sanctions), and top-level technological (energy intensity and end-use efficiency improvements) drivers. Using the framework provided by the global E{sup 3} model, the impacts of both external and internal drivers are investigated. The ability to connect external and internal drivers through this modeling framework allows the study of impacts and tradeoffs between fossil- versus nuclear-fuel burning, that includes interactions between cost, environmental, proliferation, resource, and policy issues.

  4. 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.

  5. 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.

  6. 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.

  7. Enduring Nuclear Fuel Cycle, Proceedings of a panel discussion

    Energy Technology Data Exchange (ETDEWEB)

    Walter, C. E., LLNL

    1997-11-18

    The panel reviewed the complete nuclear fuel cycle in the context of alternate energy resources, energy need projections, effects on the environment, susceptibility of nuclear materials to theft, diversion, and weapon proliferation. We also looked at ethical considerations of energy use, as well as waste, and its effects. The scope of the review extended to the end of the next century with due regard for world populations beyond that period. The intent was to take a long- range view and to project, not forecast, the future based on ethical rationales, and to avoid, as often happens, long-range discussions that quickly zoom in on only the next few decades. A specific nuclear fuel cycle technology that could satisfy these considerations was described and can be applied globally.

  8. Impact of the Taxes on Used Nuclear Fuel on the Fuel Cycle Economics in Spain

    Directory of Open Access Journals (Sweden)

    B. Yolanda Moratilla Soria

    2015-02-01

    Full Text Available In 2013, the Spanish government created two new taxes on used nuclear fuel. This article aims to present the results of an economic study carried out to compare the costs of long-term storage of used nuclear fuel –open cycle strategy–, with the cost of the strategy of reprocessing and recycling used fuel– closed cycle strategy– taking into account the impact of the new taxes on the global cost of the fuel cycle. The results show that the costs of open-cycle and closed-cycle spent fuel management, evaluated in Spain after the introduction of the taxes, are sufficiently similar (within the bounds of uncertainty, that the choice between both is predicated on other than purely economic criteria.

  9. Evaluation of conventional power systems. [emphasizing fossil fuels and nuclear energy

    Science.gov (United States)

    Smith, K. R.; Weyant, J.; Holdren, J. P.

    1975-01-01

    The technical, economic, and environmental characteristics of (thermal, nonsolar) electric power plants are reviewed. The fuel cycle, from extraction of new fuel to final waste management, is included. Emphasis is placed on the fossil fuel and nuclear technologies.

  10. Characteristics and behavior of emulsion at nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Gonda, K.; Nemoto, T.; Oka, K.

    1982-05-01

    The characteristics and behavior of the emulsion formed in mixer-settlers during nuclear fuel reprocessing were studied with the dissolver solution of spent fuel burned up to 28,000 MWd/MTU and a palladium colloidal solution, respectively. The emulsion was observed to be oil in water where nonsoluble residues of spent fuel were condensed as emulsifiers. Emulsion formed at interfaces in the settler showed electric conductivity due to continuity of the aqueous phase of the emulsion and viscosity due to the creamy state of the emulsion. The higher the palladium particle concentration was, the larger the amount of emulsion formed. This result agreed well with experience obtained in the Tokai Reprocessing Plant operation that both nonsoluble residues and emulsion formation increased remarkably on fuels in which burnup exceeded 20 000 MWd/MTU.

  11. Target-fueled nuclear reactor for medical isotope production

    Science.gov (United States)

    Coats, Richard L.; Parma, Edward J.

    2017-06-27

    A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

  12. Laser shockwave technique for characterization of nuclear fuel plate interfaces

    Science.gov (United States)

    Perton, M.; Lévesque, D.; Monchalin, J.-P.; Lord, M.; Smith, J. A.; Rabin, B. H.

    2013-01-01

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  13. Analysis of Corrosion Residues Collected from the Aluminum Basket Rails of the High-Burnup Demonstration Cask.

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.

  14. 1997 Scientific Report[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Govaerts, P.

    1998-07-01

    The 1997 Scientific Report of the Belgian Nuclear Research Centre SCK-CEN describes progress achieved in nuclear safety, radioactive waste management, radiation protection and safeguards. In the field of nuclear research, the main projects concern the behaviour of high-burnup and MOX fuel, the embrittlement of reactor pressure vessels, the irradiation-assisted stress corrosion cracking of reactor internals, and irradiation effects on materials of fusion reactors. In the field of radioactive waste management, progress in the following domains is reported: the disposal of high-level radioactive waste and spent fuel in a clay formation, the decommissioning of nuclear installations, the study of alternative waste-processing techniques. For radiation protection and safeguards, the main activities reported on are in the field of site and environmental restoration, emergency planning and response and scientific support to national and international programmes.

  15. Modern new nuclear fuel characteristics and radiation protection aspects.

    Science.gov (United States)

    Terry, Ian R

    2005-01-01

    The glut of fissile material from reprocessing plants and from the conclusion of the cold war has provided the opportunity to design new fuel types to beneficially dispose of such stocks by generating useful power. Thus, in addition to the normal reactor core complement of enriched uranium fuel assemblies, two other types are available on the world market. These are the ERU (enriched recycled uranium) and the MOX (mixed oxide) fuel assemblies. Framatome ANP produces ERU fuel assemblies by taking feed material from reprocessing facilities and blending this with highly enriched uranium from other sources. MOX fuel assemblies contain plutonium isotopes, thus exploiting the higher neutron yield of the plutonium fission process. This paper describes and evaluates the gamma, spontaneous and alpha reaction neutron source terms of these non-irradiated fuel assembly types by defining their nuclear characteristics. The dose rates which arise from these terms are provided along with an overview of radiation protection aspects for consideration in transporting and delivering such fuel assemblies to power generating utilities.

  16. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    Science.gov (United States)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. This approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.

  17. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

    Abstract—The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of reprocessing streams in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor), initial enrichment, burn up, and cooling time. Simulated gamma spectra were used to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type. Locally weighted PLS models were fitted on-the-fly to estimate continuous fuel characteristics. Burn up was predicted within 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment within approximately 2% RMSPE. This automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters and material diversions.

  18. Advanced Nuclear Fuel Cycle Transitions: Optimization, Modeling Choices, and Disruptions

    Science.gov (United States)

    Carlsen, Robert W.

    Many nuclear fuel cycle simulators have evolved over time to help understan the nuclear industry/ecosystem at a macroscopic level. Cyclus is one of th first fuel cycle simulators to accommodate larger-scale analysis with it liberal open-source licensing and first-class Linux support. Cyclus also ha features that uniquely enable investigating the effects of modeling choices o fuel cycle simulators and scenarios. This work is divided into thre experiments focusing on optimization, effects of modeling choices, and fue cycle uncertainty. Effective optimization techniques are developed for automatically determinin desirable facility deployment schedules with Cyclus. A novel method fo mapping optimization variables to deployment schedules is developed. Thi allows relationships between reactor types and scenario constraints to b represented implicitly in the variable definitions enabling the usage o optimizers lacking constraint support. It also prevents wasting computationa resources evaluating infeasible deployment schedules. Deployed power capacit over time and deployment of non-reactor facilities are also included a optimization variables There are many fuel cycle simulators built with different combinations o modeling choices. Comparing results between them is often difficult. Cyclus flexibility allows comparing effects of many such modeling choices. Reacto refueling cycle synchronization and inter-facility competition among othe effects are compared in four cases each using combinations of fleet of individually modeled reactors with 1-month or 3-month time steps. There are noticeable differences in results for the different cases. The larges differences occur during periods of constrained reactor fuel availability This and similar work can help improve the quality of fuel cycle analysi generally There is significant uncertainty associated deploying new nuclear technologie such as time-frames for technology availability and the cost of buildin advanced reactors

  19. New approaches to reprocessing of oxide nuclear fuel.

    Science.gov (United States)

    Myasoedov, B F; Kulyako, Yu M

    Dissolution of UO2, U3O8, and solid solutions of actinides in UO2 in subacid aqueous solutions (pH 0.9-1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL(-1)). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-08-10

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

  1. 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

  2. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  3. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    Science.gov (United States)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  4. 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

  5. A combined gas cooled nuclear reactor and fuel cell cycle

    Science.gov (United States)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  6. Thermodynamic treatment of noble metal fission products in nuclear fuel

    Science.gov (United States)

    Kaye, M. H.; Lewis, B. J.; Thompson, W. T.

    2007-06-01

    Based on a critical evaluation of the literature, a comprehensive thermodynamic model has been developed for the complete quinary system involving the noble metal fission products in nuclear fuel: Mo-Pd-Rh-Ru-Tc. This treatment was based on the foundation of ten binary systems and an interpolation scheme. The thermodynamic model has been demonstrated to fit the available experimental data for the ternary sub-systems. This work can be used with other models for potentially non-stoichiometric UO 2+ x containing fission products, as well as data for other phases, to assess the chemical form of fission products in irradiated fuel material.

  7. 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.

  8. 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.

  9. Nuclear reactor fuel element with vanadium getter on cladding

    Science.gov (United States)

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  10. Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jaramillo, Roger A [ORNL; Hendrich, WILLIAM R [ORNL; Packan, Nicolas H [ORNL

    2007-03-01

    A method for evaluating the room temperature ductility behavior of irradiated Zircaloy-4 nuclear fuel cladding has been developed and applied to evaluate tensile hoop strength of material irradiated to different levels. The test utilizes a polyurethane plug fitted within a tubular cladding specimen. A cylindrical punch is used to compress the plug axially, which generates a radial displacement that acts upon the inner diameter of the specimen. Position sensors track the radial displacement of the specimen outer diameter as the compression proceeds. These measurements coupled with ram force data provide a load-displacement characterization of the cladding response to internal pressurization. The development of this simple, cost-effective, highly reproducible test for evaluating tensile hoop strain as a function of internal pressure for irradiated specimens represents a significant advance in the mechanical characterization of irradiated cladding. In this project, nuclear fuel rod assemblies using Zircaloy-4 cladding and two types of mixed uranium-plutonium oxide (MOX) fuel pellets were irradiated to varying levels of burnup. Fuel pellets were manufactured with and without thermally induced gallium removal (TIGR) processing. Fuel pellets manufactured by both methods were contained in fuel rod assemblies and irradiated to burnup levels of 9, 21, 30, 40, and 50 GWd/MT. These levels of fuel burnup correspond to fast (E > 1 MeV) fluences of 0.27, 0.68, 0.98, 1.4 and 1.7 1021 neutrons/cm2, respectively. Following irradiation, fuel rod assemblies were disassembled; fuel pellets were removed from the cladding; and the inner diameter of cladding was cleaned to remove residue materials. Tensile hoop strength of this cladding material was tested using the newly developed method. Unirradiated Zircaloy-4 cladding was also tested. With the goal of determining the effect of the two fuel types and different neutron fluences on clad ductility, tensile hoop strength tests were

  11. 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.

  12. 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.

  13. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Science.gov (United States)

    2010-01-01

    ... fuel and nuclear waste. 71.97 Section 71.97 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of transportation of nuclear waste was published in the Federal Register on June...

  14. 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...

  15. Changing Perspectives on Nonproliferation and Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Choi, J; Isaacs, T H

    2005-03-29

    The concepts of international control over technologies and materials in the proliferation sensitive parts of the nuclear fuel cycle, specifically those related to enrichment and reprocessing, have been the subject of many studies and initiatives over the years. For examples: the International Fissionable Material Storage proposal in President Eisenhower's Speech on Atoms for Peace, and in the Charter of the International Atomic Energy Agency (IAEA) when the organization was formed in 1957; the regional nuclear fuel cycle center centers proposed by INFCE in the 80's; and most recently and notably, proposals by Dr. ElBaradei, the Director General of IAEA to limit production and processing of nuclear weapons usable materials to facilities under multinational control; and by U.S. President George W. Bush, to limit enrichment and reprocessing to States that have already full scale, functioning plants. There are other recent proposals on this subject as well. In this paper, the similarities and differences, as well as the effectiveness and challenges in proliferation prevention of these proposals and concepts will be discussed. The intent is to articulate a ''new nuclear regime'' and to develop concrete steps to implement such regime for future nuclear energy and deployment.

  16. Modeling of PWR fuel at extended burnup; Estudo de modelos para o comportamento a altas queimas de varetas combustiveis de reatores a agua leve pressurizada

    Energy Technology Data Exchange (ETDEWEB)

    Dias, Raphael Mejias

    2016-11-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-08-10

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

  18. Economic Analysis of Different Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Won Il Ko

    2012-01-01

    Full Text Available An economic analysis has been performed to compare four nuclear fuel cycle options: a once-through cycle (OT, DUPIC recycling, thermal recycling using MOX fuel in a pressurized water reactor (PWR-MOX, and sodium fast reactor recycling employing pyroprocessing (Pyro-SFR. This comparison was made to suggest an economic competitive fuel cycle for the Republic of Korea. The fuel cycle cost (FCC has been calculated based on the equilibrium material flows integrated with the unit cost of the fuel cycle components. The levelized fuel cycle costs (LFCC have been derived in terms of mills/kWh for a fair comparison among the FCCs, and the results are as follows: OT 7.35 mills/kWh, DUPIC 9.06 mills/kWh, PUREX-MOX 8.94 mills/kWh, and Pyro-SFR 7.70 mills/kWh. Due to unavoidable uncertainties, a cost range has been applied to each unit cost, and an uncertainty study has been performed accordingly. A sensitivity analysis has also been carried out to obtain the break-even uranium price (215$/kgU for the Pyro-SFR against the OT, which demonstrates that the deployment of the Pyro-SFR may be economical in the foreseeable future. The influence of pyrotechniques on the LFCC has also been studied to determine at which level the potential advantages of Pyro-SFR can be realized.

  19. Selenium electrochemistry. Applications in the nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Maslennikov, A.; Peretroukhine, V. [Russian Academy of Sciences, Moscow (Russian Federation). Inst. of Physical Chemistry; David, F. [Centre National de la Recherche Scientifique (CNRS), 91 - Orsay (France); Lecomte, M. [CEA Centre d' Etudes de la Valle du Rhone, 30 - Marcoule (France). Direction du Cycle du Combustible

    1999-07-01

    Modern state of selenium electrochemistry is reviewed in respect of the application of electrochemical methods for the study of the behavior of this element and its quantitative analysis in the solutions of nuclear fuel cycle. The review includes the data on the redox potentials of Se in aqueous solutions, and the data on Se redox reactions, occurring at mercury and solid electrodes. Analysis of the available literature data shows that the inverse stripping voltammetry technique for trace Se concentration and determination seems to be the most promising in application for the Se determination in PUREX solutions and in radioactive wastes. The adaptation of the ISV technique for the trace Se concentration and determination in the solutions of the nuclear fuel cycle is indicated as the most prospective goal of the future experimental study. (author)

  20. Technology development of nuclear material safeguards for DUPIC fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jong Sook; Kim, Ho Dong; Kang, Hee Young; Lee, Young Gil; Byeon, Kee Ho; Park, Young Soo; Cha, Hong Ryul; Park, Ho Joon; Lee, Byung Doo; Chung, Sang Tae; Choi, Hyung Rae; Park, Hyun Soo

    1997-07-01

    During the second phase of research and development program conducted from 1993 to 1996, nuclear material safeguards studies system were performed on the technology development of DUPIC safeguards system such as nuclear material measurement in bulk form and product form, DUPIC fuel reactivity measurement, near-real-time accountancy, and containment and surveillance system for effective and efficient implementation of domestic and international safeguards obligation. By securing in advance a optimized safeguards system with domestically developed hardware and software, it will contribute not only to the effective implementation of DUPIC safeguards, but also to enhance the international confidence build-up in peaceful use of spent fuel material. (author). 27 refs., 13 tabs., 89 figs.

  1. Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Hongbing [Univ. of Texas, Austin, TX (United States); Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott

    2014-01-09

    Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear

  2. 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.

  3. 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.

  4. Letter Report: Looking Ahead at Nuclear Fuel Resources

    Energy Technology Data Exchange (ETDEWEB)

    J. Stephen Herring

    2013-09-01

    The future of nuclear energy and its ability to fulfill part of the world’s energy needs for centuries to come depend on a reliable input of nuclear fuel, either thorium or uranium. Obviously, the present nuclear fuel cycle is completely dependent on uranium. Future thorium cycles will also depend on 235U or fissile isotopes separated from used fuel to breed 232Th into fissile 233U. This letter report discusses several emerging areas of scientific understanding and technology development that will clarify and enable assured supplies of uranium and thorium well into the future. At the most fundamental level, the nuclear energy community needs to appreciate the origins of uranium and thorium and the processes of planetary accretion by which those materials have coalesced to form the earth and other planets. Secondly, the studies of geophysics and geochemistry are increasing understanding of the processes by which uranium and thorium are concentrated in various locations in the earth’s crust. Thirdly, the study of neutrinos and particularly geoneutrinos (neutrinos emitted by radioactive materials within the earth) has given an indication of the overall global inventories of uranium and thorium, though little indication for those materials’ locations. Crustal temperature measurements have also given hints of the vertical distribution of radioactive heat sources, primarily 238U and 232Th, within the continental crust. Finally, the evolving technologies for laser isotope separation are indicating methods for reducing the energy input to uranium enrichment but also for tailoring the isotopic vectors of fuels, burnable poisons and structural materials, thereby adding another tool for dealing with long-term waste management.

  5. Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

    Science.gov (United States)

    Govers, K.; Verwerft, M.

    2016-09-01

    The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

  6. Nuclear fuel materials processing in reactive gas plasma

    Energy Technology Data Exchange (ETDEWEB)

    Min, Jin Young; Yang, Myung Seung; Seo, Yong Dae; Kim Yong Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-07-01

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and have been under post irradiation examination. (Hong, J. S.)

  7. DUPIC nuclear fuel manufacturing and process technology development at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yim, Sung Paal; Lee, Jung Won; Kim, Jong Ho; Kim, Soo Sung; Kim, Woong Ki; Yang, Myung Seung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2000-07-01

    DUPIC fuel cycle development project in KAERI of Korea was initiated in 1991 and has advanced in relevant technologies for last 10 years. The project includes five different topics such as nuclear fuel manufacturing, compatibility evaluation, performance evaluation, manufacturing facility management, and safeguards. The contents and results of DUPIC R and D up to now are as follow: - the basic foundation was established for the critically required pelletizing technology and powder treatment technology for DUPIC. - development of DUPIC process line and deployment of 20 each process equipment and examination instruments in DFDF. - powder and pellet characterization study was done at PIEF based on the simfuel study results, and 30 DUPIC pellets were successfully produced. - the manufactured pellets were used for sample fuel rods irradiated in July,2000 in HANARO research reactor in KAERI and has been under post irradiation examination. (Hong, J. S.)

  8. 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

  9. Impact of Multilateral Approaches for Assurances of Nuclear Fuel Supply

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Myung; Lee, B. W.; Ko, H. S.; Ryu, J. S.; Yang, M. H.; Oh, K. B.; Lee, K. S

    2007-12-15

    This study consists of 3 parts : analysis of the characteristics of the recent proposals for a nuclear fuel supply and the progress of them, responses from various sectors in the world, and measures for them. In response to recent proposals, majority of countries possessing sensitive nuclear fuel facilities are supportive in general. In contrast, many countries not possessing such facilities are reluctant about the proposals. To satisfy both parties, an ideal proposal could suggest measures to assure a non-proliferation as well as measures to acquire confidence from the so-called user nations. To get strong support from all countries concerned, the proposal should contain some critical elements such as clear attractiveness for a participation, equal opportunities for the participating countries, voluntarily in decision on a participation, and a gradual approach to remove any future obstacles encountered. The criteria to judge a legitimate need of a country for the introduction of nuclear fuel facilities should be prepared by a consensus. Compliance of a nonproliferation obligation, scale of an economy, and an energy security can be proposed as such criteria.

  10. Evaluation Indicators for Analysis of Nuclear Fuel Cycle Sustainability

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Ko, Won Il; Chang, Hong Lae

    2008-01-15

    In this report, an attempt was made to derive indicators for the evaluation of the sustainability of the nuclear fuel cycle, using the methodologies developed by the INPRO, OECD/NEA and Gen-IV. In deriving the indicators, the three main elements of the sustainability, i.e., economics, environmental impact, and social aspect, as well as the technological aspect of the nuclear fuel cycle, considering the importance of the safety, were selected as the main criteria. An evaluation indicator for each criterion was determined, and the contents and evaluation method of each indicator were proposed. In addition, a questionnaire survey was carried out for the objectivity of the selection of the indicators in which participated some experts of the Korea Energy Technology and Emergency Management Institute (KETEMI) . Although the proposed indicators do not satisfy the characteristics and requirements of general indicators, it is presumed that they can be used in the analysis of the sustainability of the nuclear fuel cycle because those indicators incorporate various expert judgment and public opinions. On the other hand, the weighting factor of each indicator should be complemented in the future, using the AHP method and expert advice/consultations.

  11. 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.

  12. Nuclear decay data for radionuclides occurring in routine releases from nuclear fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kocher, D.C.

    1977-08-01

    This report gives tabulations of the atomic and nuclear radiations emitted by 240 radionuclides. Most of the radionuclides are those expected to occur in routine releases of effluents from nuclear fuel cycle facilities. For each radionuclide are given the half-life and recommended values for the energies, intensities, and equilibrium absorbed-dose constants for each of the atomic and nuclear radiations. Also given are the daughter radionuclides produced and recommended values for decay branching ratios, where applicable. The radioactivity decay chains and branching ratios are displayed in diagram form.

  13. 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.

  14. National briefing summaries: Nuclear fuel cycle and waste management

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Bradley, D.J.; Fletcher, J.F.; Konzek, G.J.; Lakey, L.T.; Mitchell, S.J.; Molton, P.M.; Nightingale, R.E.

    1991-04-01

    Since 1976, the International Program Support Office (IPSO) at the Pacific Northwest Laboratory (PNL) has collected and compiled publicly available information concerning foreign and international radioactive waste management programs. This National Briefing Summaries is a printout of an electronic database that has been compiled and is maintained by the IPSO staff. The database contains current information concerning the radioactive waste management programs (with supporting information on nuclear power and the nuclear fuel cycle) of most of the nations (except eastern European countries) that now have or are contemplating nuclear power, and of the multinational agencies that are active in radioactive waste management. Information in this document is included for three additional countries (China, Mexico, and USSR) compared to the prior issue. The database and this document were developed in response to needs of the US Department of Energy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-01

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

  16. 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.

  17. Managing the Nuclear Fuel Cycle, The Big Picture

    Energy Technology Data Exchange (ETDEWEB)

    Brett W Carlsen

    2010-07-01

    The nuclear industry, at least in the United States, has failed to deliver on its promise of cheap, abundant energy. After pioneering the science and application and becoming a primary exporter of nuclear technologies, domestic use of nuclear power fell out-of-favor with the public and has been relatively stagnant for several decades. Recently, renewed interest has generated optimism and talk of a nuclear renaissance characterized by a new generation of safe, clean nuclear plants in this country. But, as illustrated by recent policy shifts regarding closure of the fuel cycle and geologic disposal of high-level radioactive wastes, significant hurdles have yet to be overcome. Using the principles of system dynamics, this paper will take a holistic look at the nuclear industry and the interactions between the key players to explore both the intended and unintended consequences of efforts to address the issues that have impeded the growth of the industry and also to illustrate aspects which must be effectively addressed if the renaissance of our industry is to be achieved and sustained.

  18. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    Science.gov (United States)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  19. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  20. The Technology Trend of Japanese Patent for the Nuclear Fuel Assembly Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jai Wan; Choi, Young Soo; Lee, Nam Ho; Jeong, Kyung Min; Suh, Yong Chil; Kim, Chang Hoi; Shin, Jung Cheol

    2008-06-15

    Japanese technology patents for the nuclear fuel assembly inspection unit, from the year 1993 to the year 2006, were investigated. The fuel rods which contain fissile material are grouped together in a closely-spaced array within the fuel assembly. Various kinds of reactor including the PWR reactor are being operated in Japan. There are many kinds of nuclear fuel assemblies in Japan, and the shape and the size of these nuclear fuel assemblies are various also. As the structure of these various fuel assemblies is a regular square as the same as the Korean one, the inspection method described in Japanese technology patent can be applied to the inspection of the nuclear fuel assembly of the Korea. This report focuses on advances in VIT(visual inspection test) of nuclear fuel assembly using the state-of-the-art CCD camera system.

  1. 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.

  2. Future nuclear fuel cycles: Prospect and challenges for actinide recycling

    Science.gov (United States)

    Warin, Dominique

    2010-03-01

    The global energy context pleads in favour of a sustainable development of nuclear energy since the demand for energy will likely increase, whereas resources will tend to get scarcer and the prospect of global warming will drive down the consumption of fossil fuel. In this context, nuclear power has the worldwide potential to curtail the dependence on fossil fuels and thereby to reduce the amount of greenhouse gas emissions while promoting energy independence. How we deal with nuclear radioactive waste is crucial in this context. In France, the public's concern regarding the long-term waste management made the French Governments to prepare and pass the 1991 and 2006 Acts, requesting in particular the study of applicable solutions for still minimizing the quantity and the hazardousness of final waste. This necessitates High Active Long Life element (such as the Minor Actinides MA) recycling, since the results of fuel cycle R&D could significantly change the challenges for the storage of nuclear waste. HALL recycling can reduce the heat load and the half-life of most of the waste to be buried to a couple of hundred years, overcoming the concerns of the public related to the long-life of the waste and thus aiding the "burying approach" in securing a "broadly agreed political consensus" of waste disposal in a geological repository. This paper presents an overview of the recent R and D results obtained at the CEA Atalante facility on innovative actinide partitioning hydrometallurgical processes. For americium and curium partitioning, these results concern improvements and possible simplifications of the Diamex-Sanex process, whose technical feasibility was already demonstrated in 2005. Results on the first tests of the Ganex process (grouped actinide separation for homogeneous recycling) are also discussed. In the coming years, next steps will involve both better in-depth understanding of the basis of these actinide partitioning processes and, for the new promising

  3. 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

  4. Implementation of integrated safeguards at nuclear fuel plant Pitesti Romania

    Energy Technology Data Exchange (ETDEWEB)

    Olaru, Vasilica; Tiberiu, Ivana; Epure, Gheorghe [Nuclear Safety Department, Nuclear Fuel Plant Pitesti, Cimpului, No 1, 115400 Mioveni (Romania)

    2010-07-01

    The nuclear activity in Romania was for many years under Traditional Safeguards (TS) and has developed in good conditions this type of nuclear safeguards. Now it has the opportunity to improve the performance and quality of the safeguards activity and increase the accountancy and control of nuclear material by passing to Integrated Safeguards (IS). The legal framework is Law 100/2000 for ratification of the Protocol between Romania and International Atomic Energy Agency (IAEA), additional to the Agreement between the Socialist Republic of Romania Government and IAEA related to safeguards as part of the Treaty on the non-proliferation of nuclear weapons published in the Official Gazette no. 3/31 January 1970, and the Additional Protocol content published in the Official Gazette no. 295/ 29.06.2000. The first discussion about Integrated Safeguards (IS) between Nuclear Fuel Plant (FCN) representatives and IAEA inspectors was in June 2005. In Feb. 2007 an IAEA mission visited FCN and established the main steps for implementing the IS. There were visited the storages, technological flow, and was reviewed the residence times for different nuclear materials, the applied chemical analysis, metrological methods, weighting method and elaborating the documents and lists. At that time the IAEA and FCN representatives established the main points for starting the IS at FCN: perform the Short Notice Random Inspections (SNRI), communicate the eligible days for SNRI for each year, communicate the estimated deliveries and shipments for first quarter and then for the rest of the year, daily mail box declaration (DD) with respect to the residence time for several nuclear material, advance notification (AN) for each nuclear material transfer (shipments and receipts), others. At 01 June 2007 Romania has passed officially to Integrated Safeguards and FCN (RO-D) has taken all measures to realize this objective. (authors)

  5. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  6. Science based integrated approach to advanced nuclear fuel development - vision, approach, and overview

    Energy Technology Data Exchange (ETDEWEB)

    Unal, Cetin [Los Alamos National Laboratory; Pasamehmetoglu, Kemal [IDAHO NATIONAL LAB; Carmack, Jon [IDAHO NATIONAL LAB

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Rcactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems is critical. In order to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. The purpose of this paper is to identify the modeling and simulation approach in order to deliver predictive tools for advanced fuels development. The coordination between experimental nuclear fuel design, development technical experts, and computational fuel modeling and simulation technical experts is a critical aspect of the approach and naturally leads to an integrated, goal-oriented science-based R & D approach and strengthens both the experimental and computational efforts. The Advanced Fuels Campaign (AFC) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) Fuels Integrated Performance and Safety Code (IPSC) are working together to determine experimental data and modeling needs. The primary objective of the NEAMS fuels IPSC project is to deliver a coupled, three-dimensional, predictive computational platform for modeling the fabrication and both normal and abnormal operation of nuclear fuel pins and assemblies, applicable to both existing and future reactor fuel designs. The science based program is pursuing the development of an integrated multi-scale and multi-physics modeling and simulation platform for nuclear fuels. This overview paper discusses the vision, goals and approaches how to develop and implement the new approach.

  7. Thermoacoustic enhancements for nuclear fuel rods and other high temperature applications

    Energy Technology Data Exchange (ETDEWEB)

    Garrett, Steven L.; Smith, James A.; Kotter, Dale K.

    2017-05-09

    A nuclear thermoacoustic device includes a housing defining an interior chamber and a portion of nuclear fuel disposed in the interior chamber. A stack is disposed in the interior chamber and has a hot end and a cold end. The stack is spaced from the portion of nuclear fuel with the hot end directed toward the portion of nuclear fuel. The stack and portion of nuclear fuel are positioned such that an acoustic standing wave is produced in the interior chamber. A frequency of the acoustic standing wave depends on a temperature in the interior chamber.

  8. Metrological evaluation of characterization methods applied to nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Faeda, Kelly Cristina Martins; Lameiras, Fernando Soares; Camarano, Denise das Merces; Ferreira, Ricardo Alberto Neto; Migliorini, Fabricio Lima; Carneiro, Luciana Capanema Silva; Silva, Egonn Hendrigo Carvalho, E-mail: kellyfisica@gmail.co, E-mail: fernando.lameiras@pq.cnpq.b, E-mail: dmc@cdtn.b, E-mail: ranf@cdtn.b, E-mail: flmigliorini@hotmail.co, E-mail: lucsc@hotmail.co, E-mail: egonn@ufmg.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2010-07-01

    In manufacturing the nuclear fuel, characterizations are performed in order to assure the minimization of harmful effects. The uranium dioxide is the most used substance as nuclear reactor fuel because of many advantages, such as: high stability even when it is in contact with water at high temperatures, high fusion point, and high capacity to retain fission products. Several methods are used for characterization of nuclear fuels, such as thermogravimetric analysis for the ratio O / U, penetration-immersion method, helium pycnometer and mercury porosimetry for the density and porosity, BET method for the specific surface, chemical analyses for relevant impurities, and the laser flash method for thermophysical properties. Specific tools are needed to control the diameter and the sphericity of the microspheres and the properties of the coating layers (thickness, density, and degree of anisotropy). Other methods can also give information, such as scanning and transmission electron microscopy, X-ray diffraction, microanalysis, and mass spectroscopy of secondary ions for chemical analysis. The accuracy of measurement and level of uncertainty of the resulting data are important. This work describes a general metrological characterization of some techniques applied to the characterization of nuclear fuel. Sources of measurement uncertainty were analyzed. The purpose is to summarize selected properties of UO{sub 2} that have been studied by CDTN in a program of fuel development for Pressurized Water Reactors (PWR). The selected properties are crucial for thermalhydraulic codes to study basic design accidents. The thermal characterization (thermal diffusivity and thermal conductivity) and the penetration immersion method (density and open porosity) of UO{sub 2} samples were focused. The thermal characterization of UO{sub 2} samples was determined by the laser flash method between room temperature and 448 K. The adaptive Monte Carlo Method was used to obtain the endpoints of

  9. LOAD-CHECK, program supported optimization of the fuel element disposal in cask CASTOR {sup registered} V casks; LOAD-CHECK, programmunterstuetzte Optimierung der Brennelemententsorgung in CASTOR {sup registered} V-Behaeltern

    Energy Technology Data Exchange (ETDEWEB)

    Amian, D.; Braun, A. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany); Graf, R.; Hoffmann, V. [GNS Gesellschaft fuer Nuklear-Service mbH, Essen (Germany)

    2010-05-15

    LOAD-CHECK is an interactive program module for the systematic and strategic spent fuel disposal planning. Using physical fuel element data the loading scenarios for the routine operation and the post-closure operation phase can be simulated for free selectable time periods. The basis for the loading license application are the available spent fuel casks according to the regulations of the interim storage facility. LOAD-CHECK allows the optimization of the loading campaigns with respect to the time schedule and the number of casks including the planning of optimized disposal of special spent fuel (MOX fuel elements or high-burnup fuel elements). Possibilities for a reduced post-closure operating phase of nuclear power plants might be the consequence.

  10. 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

  11. Ablation study of tungsten-based nuclear thermal rocket fuel

    Science.gov (United States)

    Smith, Tabitha Elizabeth Rose

    The research described in this thesis has been performed in order to support the materials research and development efforts of NASA Marshall Space Flight Center (MSFC), of Tungsten-based Nuclear Thermal Rocket (NTR) fuel. The NTR was developed to a point of flight readiness nearly six decades ago and has been undergoing gradual modification and upgrading since then. Due to the simplicity in design of the NTR, and also in the modernization of the materials fabrication processes of nuclear fuel since the 1960's, the fuel of the NTR has been upgraded continuously. Tungsten-based fuel is of great interest to the NTR community, seeking to determine its advantages over the Carbide-based fuel of the previous NTR programs. The materials development and fabrication process contains failure testing, which is currently being conducted at MSFC in the form of heating the material externally and internally to replicate operation within the nuclear reactor of the NTR, such as with hot gas and RF coils. In order to expand on these efforts, experiments and computational studies of Tungsten and a Tungsten Zirconium Oxide sample provided by NASA have been conducted for this dissertation within a plasma arc-jet, meant to induce ablation on the material. Mathematical analysis was also conducted, for purposes of verifying experiments and making predictions. The computational method utilizes Anisimov's kinetic method of plasma ablation, including a thermal conduction parameter from the Chapman Enskog expansion of the Maxwell Boltzmann equations, and has been modified to include a tangential velocity component. Experimental data matches that of the computational data, in which plasma ablation at an angle shows nearly half the ablation of plasma ablation at no angle. Fuel failure analysis of two NASA samples post-testing was conducted, and suggestions have been made for future materials fabrication processes. These studies, including the computational kinetic model at an angle and the

  12. Heat transfer in nuclear fuels: Measurements of gap conductance

    Science.gov (United States)

    Cho, Chun Hyung

    Heat transfer in the fuel-clad gap in a nuclear reactor impacts the overall temperature distribution, stored energy and the mechanical properties of a nuclear fuel rod. Therefore, an accurate estimation of the gap conductance between the fuel and the clad is critically important for reactor design and operations. To obtain the requisite accuracy in the gap conductance estimation, it is important to understand the effects of the convective heat transfer coefficient, the gas composition, pressure and temperature, and so forth. The objectives of this study are to build a bench-scale experimental apparatus for the measurement of thermal gap conductances and to develop a better understanding of the differences that have been previously observed between such measured values and those predicted theoretically. This is accomplished by employing improved analyses of the experiments and improved theoretical models. Using laser heating of slightly separated stainless-steel plates, the gap conductance was measured using a technique that compares the theoretical and experimental time dependent temperatures at the back surface of the second plate. To consider the effects of surface temperature and gas pressure, the theoretical temperatures were calculated using a convective heat transfer coefficient that was dependent upon both the temperature and the gas pressure.

  13. Project of a new circuit for nuclear fuel irradiation; Projeto de um novo circuito para irradiacao de combustiveis nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Zeituni, Carlos A.; Terremoto, Luis A.A.; Perrotta, Jose A.; Silva, Jose E.R. da [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear. Div. de Engenharia do Combustivel. E-mail: czeituni@usp.br

    2000-07-01

    This paper reports information about the operation of the old Irradiated Fuel Assembly for nuclear miniplates irradiation in the reactor IEA-R1, named CICON (Circuit for Nuclear Fuels Irradiation), and presents the project of the new one. This paper also describes the problems of the old capsule and which details we will change in the new project. (author)

  14. 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)

  15. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-01

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

  17. Ground test facilities for evaluating nuclear thermal propulsion engines and fuel elements

    Science.gov (United States)

    Allen, G. C.; Beck, D. F.; Harmon, C. D.; Shipers, L. R.

    Interagency panels evaluating nuclear thermal propulsion development options have consistently recognized the need for constructing a major new ground test facility to support fuel element and engine testing. This paper summarizes the requirements, configuration, and design issues of a proposed ground test complex for evaluating nuclear thermal propulsion engines and fuel elements being developed for the Space Nuclear Thermal Propulsion (SNTP) program.

  18. 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.

  19. Uranium in the Nuclear Fuel Cycle: Creation of Plutonium (Invited)

    Science.gov (United States)

    Ewing, R. C.

    2009-12-01

    One of the important properties of uranium is that it can be used to “breed” higher actinides, particularly plutonium. During the past sixty years, more than 1,800 metric tonnes of Pu, and substantial quantities of the “minor” actinides, such as Np, Am and Cm, have been generated in nuclear reactors - a permanent record of nuclear power. Some of these transuranium elements can be a source of energy in fission reactions (e.g., 239Pu), a source of fissile material for nuclear weapons (e.g., 239Pu and 237Np), and of environmental concern because of their long-half lives and radiotoxicity (e.g., 239Pu and 237Np). In fact, the new strategies of the Advance Fuel Cycle Initiative (AFCI) are, in part, motivated by an effort to mitigate some of the challenges of the disposal of these long-lived actinides. There are two basic strategies for the disposition of these heavy elements: 1.) to “burn” or transmute the actinides using nuclear reactors or accelerators; 2.) to “sequester” the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. There has been substantial interest in the use of actinide-bearing minerals, such as zircon or isometric pyrochlore, A2B2O7 (A= rare earths; B = Ti, Zr, Sn, Hf), for the immobilization of actinides, particularly plutonium, both as inert matrix fuels and nuclear waste forms. Systematic studies of rare-earth pyrochlores have led to the discovery that certain compositions (B = Zr, Hf) are stable to very high doses of alpha-decay event damage1. The radiation stability of these compositions is closely related to the structural distortions that can be accommodated for specific pyrochlore compositions and the electronic structure of the B-site cation. Recent developments in the understanding of the properties of heavy element solids have opened up new possibilities for the design of advanced nuclear fuels and waste forms.

  20. Metallic fuels: The EBR-II legacy and recent advances

    Energy Technology Data Exchange (ETDEWEB)

    Douglas L. Porter; Steven L. Hayes; J. Rory Kennedy

    2012-09-01

    Experimental Breeder Reactor – II (EBR-II) metallic fuel was qualified for high burnup to approximately 10 atomic per cent. Subsequently, the electrometallurgical treatment of this fuel was demonstrated. Advanced metallic fuels are now investigated for increased performance, including ultra-high burnup and actinide burning. Advances include additives to mitigate the fuel/cladding chemical interaction and uranium alloys that combine Mo, Ti and Zr to improve alloy performance. The impacts of the advances—on fabrication, waste streams, electrorefining, etc.—are found to be minimal and beneficial. Owing to extensive research literature and computational methods, only a modest effort is required to complete their development.

  1. In-Pile Thermal Conductivity Measurement Method for Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Joy L. Rempe; Brandon Fox; Heng Ban; Joshua E. Daw; Darrell L. Knudson; Keith G. Condie

    2009-08-01

    Thermophysical properties of advanced nuclear fuels and materials during irradiation must be known prior to their use in existing, advanced, or next generation reactors. Thermal conductivity is one of the most important properties for predicting fuel and material performance. A joint Utah State University (USU) / Idaho National Laboratory (INL) project, which is being conducted with assistance from the Institute for Energy Technology at the Norway Halden Reactor Project, is investigating in-pile fuel thermal conductivity measurement methods. This paper focuses on one of these methods – a multiple thermocouple method. This two-thermocouple method uses a surrogate fuel rod with Joule heating to simulate volumetric heat generation to gain insights about in-pile detection of thermal conductivity. Preliminary results indicated that this method can measure thermal conductivity over a specific temperature range. This paper reports the thermal conductivity values obtained by this technique and compares these values with thermal property data obtained from standard thermal property measurement techniques available at INL’s High Test Temperature Laboratory. Experimental results and material properties data are also compared to finite element analysis results.

  2. 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.

  3. 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.

  4. Globalisation of the nuclear fuel cycle - impact of developments on fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Durpel, L. van den; Bertel, E. [OECD Nuclear Energy Agency, 92 - Issy-les-Moulineaux (France)

    2000-02-01

    Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the deregulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to complete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according to the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economic perspective including environmental and social considerations. (orig.) [German] Die Kernenergie wird sich mehr und mehr in einem Umfeld behaupten muessen, das durch schnelle Veraenderungen auf Grund des Wettbewerbsdrucks in der Wirtschaft und des Liberalisierungsprozesses gekennzeichnet ist. Im heutigen Wirtschaftsumfeld muessen sich die Energieversorgungsunternehmen hauptsaechlich auf die Senkung ihrer Stromerzeugungs-Gesamtkosten konzentrieren. Darunter fallen auch die Brennstoffkreislaufkosten, die sie nur zum Teil beeinflussen koennen. Kurzfristig gesehen, duerften die Entwicklungen im Brennstoffkreislauf eher evolutionaer verlaufen und den jeweiligen Beduerfnissen der EVUs entsprechen. Im Zusammenhang mit einer

  5. 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.

  6. 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.

  7. 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)

  8. Models and simulations of nuclear fuel materials properties

    Energy Technology Data Exchange (ETDEWEB)

    Stan, M. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)], E-mail: mastan@lanl.gov; Ramirez, J.C. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States); Cristea, P. [University of Bucharest, Faculty of Physics, Bucuresti-Magurele (Romania); Hu, S.Y.; Deo, C.; Uberuaga, B.P.; Srivilliputhur, S.; Rudin, S.P.; Wills, J.M. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)

    2007-10-11

    To address the complexity of the phenomena that occur in a nuclear fuel element, a multi-scale method was developed. The method incorporates theory-based atomistic and continuum models into finite element simulations to predict heat transport phenomena. By relating micro and nano-scale models to the macroscopic equilibrium and non-equilibrium simulations, the predictive character of the method is improved. The multi-scale approach was applied to calculations of point defect concentration, helium bubbles formation, oxygen diffusivity, and simulations of heat and mass transport in UO{sub 2+x}.

  9. 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

  10. 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.

  11. Degree of Sustainability of Various Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R.; Krakowski, R.A. [Los Alamos National Laboratory, New Mexico (United States)

    2002-08-01

    The focus of this study is on a 'top-level' examination of the sustainability of nuclear energy in the context of the overall nuclear fuel cycle (NFC). This evaluation is conducted according to a set of established sustainability criteria that encompasses key economic (energy generation costs), environmental (resource utilization, long-term waste accumulations), and societal (nuclear-weapons proliferation risk) concerns associated with present and future NFC approaches. In this study, key NFCs are assessed according to a simplified and limited set of criteria that attempts to quantify NFC concerns related to cost, resource, waste, and proliferation. The overarching aim of this study is to examine a representative set of NFC options on a relative basis according to the adopted set of criteria to aid in the assessment and decision-making process. These criteria were then aggregated into a single, composite metric to examine the impacts of specific 'stakeholder' preferences. The study architecture is based on sets of nuclear process components. These sets are assembled around a particular nuclear reactor technology for the generation of electricity. Selections are made from the resulting sets of reactor-centric technologies and grouped to form nine central NFC scenarios. The above-described sustainability metrics are evaluated using a steady-state (equilibrium), highly aggregated model that is applied through mass and energy conservation to evaluate each NFC scenario. Six NFC scenarios examined to varying degrees are adaptations or extensions of scenarios used in a recent OECD study (OECD, 2002) of partitioning and transmutation (P and T) schemes based on accelerator-driven systems (ADS) or fast reactors (FR). Three NFC scenarios are based entirely on present-day or near-term LWR technologies. In addition to these near-term scenarios, more advanced systems considered in the original OECD study on which this model is based were retained using a

  12. Multiscale Modeling and Uncertainty Quantification for Nuclear Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Estep, Donald [Colorado State Univ., Fort Collins, CO (United States); El-Azab, Anter [Florida State Univ., Tallahassee, FL (United States); Pernice, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Peterson, John W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Polyakov, Peter [Univ. of Wyoming, Laramie, WY (United States); Tavener, Simon [Colorado State Univ., Fort Collins, CO (United States); Xiu, Dongbin [Purdue Univ., West Lafayette, IN (United States); Univ. of Utah, Salt Lake City, UT (United States)

    2017-03-23

    In this project, we will address the challenges associated with constructing high fidelity multiscale models of nuclear fuel performance. We (*) propose a novel approach for coupling mesoscale and macroscale models, (*) devise efficient numerical methods for simulating the coupled system, and (*) devise and analyze effective numerical approaches for error and uncertainty quantification for the coupled multiscale system. As an integral part of the project, we will carry out analysis of the effects of upscaling and downscaling, investigate efficient methods for stochastic sensitivity analysis of the individual macroscale and mesoscale models, and carry out a posteriori error analysis for computed results. We will pursue development and implementation of solutions in software used at Idaho National Laboratories on models of interest to the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program.

  13. French investigations of high burnup effect on LOCA thermomechanical behavior: Part 1. Experimental programmes in support of LOCA design methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Waeckel, N. [EDF/SEPTEN Villeurbanne (France); GrandJean, C. [IPSN, Cadarache (France); Cauvin, R.; Lebuffe, C. [EDF/SCMI, Chinon (France)

    1997-01-01

    Within the framework of Burn-Up extension request, EDF, FRAMATOME, CEA and IPSN have carried out experimental programmes in order to provide the design of fuel rods under LOCA conditions with relevant data. The design methods used in France for LOCA are based on standard Appendix K methodology updated to take into account some penalties related to the actual conditions of the Nuclear Power Plant. Best-Estimate assessments are used as well. Experimental programmes concern plastic deformation and burst behavior of advanced claddings (EDGAR) and thermal shock quenching behavior of highly irradiated claddings (TAGCIR). The former reveals the important role played by the {alpha}/{beta} transformation kinetics related to advanced alloys (Niobium alloys) and the latter the significative impact of hydrogen charged during in-reactor corrosion on oxidation kinetics and failure behavior in terms of cooling rates.

  14. Non-destructive Testing Dummy Nuclear Fuel Rods by Neutron Radiography

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng

    2013-01-01

    As a unique non-destructive testing technique,neutron radiography can be used to measure nuclear fuel rods with radioactivity.The images of the dummy nuclear fuel rods were obtained at the CARR.Through imaging analysis methods,the structure defections,the hydrogen accumulation in the cladding and the 235U enrichment of the pellet were studied and analyzed.Experiences for non-destructive testing real PWR nuclear fuel rods by NR

  15. 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.

  16. Environmental Impact Statement on the concept for disposal of Canada's nuclear fuel waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-01

    This report describes the many fundamental issues relating to the strategy being proposed by Atomic Energy of Canada Limited for the long-term management of nuclear fuel waste. It discusses the need for a method for disposal of nuclear fuel waste that would permanently protect human health and the natural environment and that would not unfairly burden future generations. It also describes the background and mandate of the Nuclear Fuel Waste Management Program in Canada.

  17. ENUSA in the international market of nuclear fuel; ENUSA en el mercado mundial de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez, J. E.; Gonzalez, R.

    2002-07-01

    ENUSA Industrias Avanzadas, s. a. has a promising future into the nuclear fuel world market evolving from a starting model of a state owned company focused on the Spanish market and using exclusively the technology coming from their American partners, to a current model of a modern company where technological innovation, the competitiveness and the quality are key factors in the developments of his business. ENUSA is a company oriented to the export sustained by the integrated services provided to the Spanish nuclear sector,where Europe is his natural market, but open to seek opportunities together with his technological partners into the American and Japanese market. (Author)

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-25

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

  20. 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.

  1. Impact of thicker cladding on the nuclear parameters of the NPP Krsko fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kromar, Marjan, E-mail: marjan.kromar@ijs.s [Jozef Stefan Institute, Reactor Physics Department, Jamova 39, 1001 Ljubljana (Slovenia); Kurincic, Bojan [Nuclear Power Plant Krsko, Engineering Division, Nuclear Fuel and Reactor Core, Vrbina 12, 8270 Krsko (Slovenia)

    2011-04-15

    To make fuel rods more resistant to grid-to-rod fretting or other cladding penetration failures, the cladding thickness could be increased or strengthened. Implementation of thicker fuel rod cladding was evaluated for the NPP Krsko that uses 16 x 16 fuel design. Cladding thickness of the Westinghouse standard fuel design (STD) and optimized fuel design (OFA) is increased. The reactivity effect during the fuel burnup is determined. To obtain a complete realistic view of the fuel behaviour a typical, near equilibrium, 18-month fuel cycle is investigated. The most important nuclear core parameters such as critical boron concentrations, isothermal temperature coefficient and rod worth are determined and compared.

  2. 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.

  3. SPOUTED BED DESIGN CONSIDERATIONS FOR COATED NUCLEAR FUEL PARTICLES

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas W.

    2017-07-01

    High Temperature Gas Cooled Reactors (HTGRs) are fueled with tristructural isotropic (TRISO) coated nuclear fuel particles embedded in a carbon-graphite fuel body. TRISO coatings consist of four layers of pyrolytic carbon and silicon carbide that are deposited on uranium ceramic fuel kernels (350µm – 500µm diameters) in a concatenated series of batch depositions. Each layer has dedicated functions such that the finished fuel particle has its own integral containment to minimize and control the release of fission products into the fuel body and reactor core. The TRISO coatings are the primary containment structure in the HTGR reactor and must have very high uniformity and integrity. To ensure high quality TRISO coatings, the four layers are deposited by chemical vapor deposition (CVD) using high purity precursors and are applied in a concatenated succession of batch operations before the finished product is unloaded from the coating furnace. These depositions take place at temperatures ranging from 1230°C to 1550°C and use three different gas compositions, while the fuel particle diameters double, their density drops from 11.1 g/cm3 to 3.0 g/cm3, and the bed volume increases more than 8-fold. All this is accomplished without the aid of sight ports or internal instrumentation that could cause chemical contamination within the layers or mechanical damage to thin layers in the early stages of each layer deposition. The converging section of the furnace retort was specifically designed to prevent bed stagnation that would lead to unacceptably high defect fractions and facilitate bed circulation to avoid large variability in coating layer dimensions and properties. The gas injection nozzle was designed to protect precursor gases from becoming overheated prior to injection, to induce bed spouting and preclude bed stagnation in the bottom of the retort. Furthermore, the retort and injection nozzle designs minimize buildup of pyrocarbon and silicon carbide on the

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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......

  10. ENVIRONMENTAL ASSESSMENT METHODOLOGY FOR THE NUCLEAR FUEL CYCLE

    Energy Technology Data Exchange (ETDEWEB)

    Brenchley, D. L.; Soldat, J. K.; McNeese, J. A.; Watson, E. C.

    1977-07-01

    This report describes the methodology for determining where environmental control technology is required for the nuclear fuel cycle. The methodology addresses routine emission of chemical and radioactive effluents, and applies to mining, milling, conversion, enrichment, fuel fabrication, reactors (LWR and BWR) and fuel reprocessing. Chemical and radioactive effluents are evaluated independently. Radioactive effluents are evaluated on the basis of maximum exposed individual dose and population dose calculations for a 1-year emission period and a 50-year commitment. Sources of radionuclides for each facility are then listed according to their relative contribution to the total calculated dose. Effluent, ambient and toxicology standards are used to evaluate the effect of chemical effluents. First, each chemical and source configuration is determined. Sources are tagged if they exceed existirrg standards. The combined effect of all chemicals is assessed for each facility. If the additive effects are unacceptable, then additional control technology is recommended. Finally, sources and their chemicals at each facility are ranked according to their relative contribution to the ambient pollution level. This ranking identifies those sources most in need of environmental control.

  11. 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.

  12. A Concept of An Accelerator Closed Nuclear Fuel Cycle

    Science.gov (United States)

    Eremeev, I. P.

    1997-05-01

    The physical approach (I.P.Eremeev. Proc. of the PAC-95. Vol.1, p.98.) is applied for technology of nuclear fuel cycle. It is proposed the cycle to be closed by such an accelerator based process link, which would allow, on the one hand, the most hazardous of "equilibrium" radionuclides to be transmuted to stable isotopes or incinerated and, on the other hand, additional fissile fuel to be produced to compensate the energy consumption. Parameters of the technology, such as an intensity and energy "cost" of a transmutation event, a flux of photoneutrons produced have been determined for model targets. It is shown that the approach allows the above fission/transuranium radionuclides to be transmuted/ incinerated at a much greater rate than that of their build-up in operating NPP reactors at a much less energy consumption than an energy produced under their formation and at considerable compensation of the consumed energy through breeding fissile isotopes. A possibility of going to a closed Th-U fuel cycle is discussed. To realize the technology proposed requirements to a system of electron accelerators are formulated.

  13. Used nuclear fuel separations process simulation and testing

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, C.; Krebs, J.F.; Copple, J.M.; Frey, K.E.; Maggos, L.E.; Figueroa, J.; Willit, J.L.; Papadias, D.D. [Argonne National Laboratory: 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2013-07-01

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

  14. 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.

  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. A review on the status of development in thorium-based nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Woo; Na, S. H.; Lee, Y. W.; Kim, H. S.; Kim, S. H.; Joung, C.Y

    2000-02-01

    Thorium as an alternative nuclear energy source had been widely investigated in the 1950s-1960s because it is more abundant than uranium, but the studies of thorium nuclear fuel cycle were discontinued by political and economic reasons in the 1970s. Recently, however, renewed interest was vested in thorium-based nuclear fuel cycle because it may generate less long-lived minor actinides and has a lower radiotoxicity of high level wastes after reprocessing compared with the thorium fuel cycle. In this state-of the art report, thorium-based nuclear cycle. In this state-of the art report, thorium-based nuclear fuel cycle and fuel fabrication processes developed so far with different reactor types are reviewed and analyzed to establish basic technologies of thorium fuel fabrication which could meet our situation. (author)

  17. 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...

  18. Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Trammell, Michael P [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Qualls, A L [ORNL; Harrison, Thomas J [ORNL

    2013-01-01

    ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

  19. 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...

  20. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  1. 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.

  2. 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.

  3. Environmental Justice, Place and Nuclear Fuel Waste Management in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Kuhn, Richard G. [Univ. of Guelph (Canada). Dept. of Geography; Murphy, Brenda L. [Wilfrid Launer Univ., Brantford (Canada)

    2006-09-15

    The purpose of this paper is to outline the basis of a Nuclear Fuel Waste management strategy for Canada, taking into account the unique legal tenets (Aboriginal rights; federal - provincial jurisdiction) and the orientation that the Nuclear Waste Management Organization (NWMO) has taken to date. The focus of the paper are grounded in notions of environmental justice. Bullard's definition provides a useful guideline: 'the fair treatment and meaningful involvement of all people regardless of race, colour, national origin or income with respect to the development, implementation and enforcement of environmental laws, regulations and policies'. The overriding concern is to work towards a process that is inclusive and just. Prior to developing a specific strategy to site a NFW disposal facility, we maintain that the NWMO needs to first address three fundamental issues: Expand its mandate to include the future of nuclear energy in Canada; Provide an inclusive role for First Nations (Aboriginal people) in all stages of the process; Adhere to the requirement of specifying an economic region and deal more overtly with the transportation of NF.

  4. 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)

  5. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Shao, Lin [Texas A & M Univ., College Station, TX (United States); Tsvetkov, Pavel [Texas A & M Univ., College Station, TX (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  6. Development and verification of NRC`s single-rod fuel performance codes FRAPCON-3 AND FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, C.E.; Cunningham, M.E.; Lanning, D.D. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    The FRAPCON and FRAP-T code series, developed in the 1970s and early 1980s, are used by the US Nuclear Regulatory Commission (NRC) to predict fuel performance during steady-state and transient power conditions, respectively. Both code series are now being updated by Pacific Northwest National Laboratory to improve their predictive capabilities at high burnup levels. The newest versions of the codes are called FRAPCON-3 and FRAPTRAN. The updates to fuel property and behavior models are focusing on providing best estimate predictions under steady-state and fast transient power conditions up to extended fuel burnups (> 55 GWd/MTU). Both codes will be assessed against a data base independent of the data base used for code benchmarking and an estimate of code predictive uncertainties will be made based on comparisons to the benchmark and independent data bases.

  7. To Recycle or Not to Recycle? An Intergenerational Approach to Nuclear Fuel Cycles

    NARCIS (Netherlands)

    Taebi, B.; Kloosterman, J.L.

    2007-01-01

    AbstractThis paper approaches the choice between the open and closed nuclear fuel cycles as a matter of intergenerational justice, by revealing the value conflicts in the production of nuclear energy. The closed fuel cycle improve sustainability in terms of the supply certainty of uranium and involv

  8. The nuclear fuels tax is in conformity with constitutional law; Die Kernbrennstoffsteuer ist verfassungsgemaess

    Energy Technology Data Exchange (ETDEWEB)

    Faehrmann, Ingo; Ringwald, Roman [Sozietaet Becker Buettner Held, Berlin (Germany)

    2012-02-13

    There are rulings by three courts of finance concerning the conformity of the nuclear fuels tax with German constitutional law. While the FG Hamburg and FG Munich were in some doubt, the FG Baden-Wuerttemberg was of the opinion that the nuclear fuels tax act is compatible with German constitutional law.

  9. Preliminary Study on Method of Quantitative Measurement of Nuclear Fuel Rod by Neutron CT at CARR

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; WANG; Hong-li; HE; Lin-feng; WANG; Yu; WU; Mei-mei; LIU; Yun-tao; CHEN; Dong-feng

    2015-01-01

    Neutron CT technique was applied to the quantitative measurement of the key parameters of nuclear fuel rods at China Advanced Research Reactor(CARR).The sample of dummy nuclear fuel rod was rotated in 180°range,and 900neutron projections were obtained.The 3-D neutron

  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. Preliminary Nuclear Analysis for the HANARO Fuel Element with Burnable Absorber

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, So Young; In, Won Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Burnable absorber is used for reducing reactivity swing and power peaking in high performance research reactors. Development of the HANARO fuel element with burnable absorber was started in the U-Mo fuel development program at HANARO, but detailed full core analysis was not performed because the current HANARO fuel management system is uncertain to analysis the HANARO core with burnable absorber. A sophisticated reactor physics system is required to analysis the core. The McCARD code was selected and the detailed McCARD core models, in which the basic HANARO core model was developed by one of the McCARD developers, are used in this study. The development of nuclear fuel requires a long time and correct developing direction especially by the nuclear analysis. This paper presents a preliminary nuclear analysis to promote the fuel development. Based on the developed fuel, the further nuclear analysis will improve reactor performance and safety. Basic nuclear analysis for the HANARO and the AHR were performed for getting the proper fuel elements with burnable absorber. Addition of 0.3 - 0.4% Cd to the fuel meat is promising for the current HANARO fuel element. Small addition of burnable absorber may not change any fuel characteristics of the HANARO fuel element, but various basic tests and irradiation tests at the HANARO core are required.

  12. Minor Actinides Loading Optimization for Proliferation Resistant Fuel Design - BWR

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; Hongbin Zhang

    2009-09-01

    One approach to address the United States Nuclear Power (NP) 2010 program for the advanced light water reactor (LWR) (Gen-III+) intermediate-term spent fuel disposal need is to reduce spent fuel storage volume while enhancing proliferation resistance. One proposed solution includes increasing burnup of the discharged spent fuel and mixing minor actinide (MA) transuranic nuclides (237Np and 241Am) in the high burnup fuel. Thus, we can reduce the spent fuel volume while increasing the proliferation resistance by increasing the isotopic ratio of 238Pu/Pu. For future advanced nuclear systems, MAs are viewed more as a resource to be recycled, and transmuted to less hazardous and possibly more useful forms, rather than simply disposed of as a waste stream in an expensive repository facility. MAs play a much larger part in the design of advanced systems and fuel cycles, not only as additional sources of useful energy, but also as direct contributors to the reactivity control of the systems into which they are incorporated. A typical boiling water reactor (BWR) fuel unit lattice cell model with UO2 fuel pins will be used to investigate the effectiveness of adding MAs (237Np and/or 241Am) to enhance proliferation resistance and improve fuel cycle performance for the intermediate-term goal of future nuclear energy systems. However, adding MAs will increase plutonium production in the discharged spent fuel. In this work, the Monte-Carlo coupling with ORIGEN-2.2 (MCWO) method was used to optimize the MA loading in the UO2 fuel such that the discharged spent fuel demonstrates enhanced proliferation resistance, while minimizing plutonium production. The axial averaged MA transmutation characteristics at different burnup were compared and their impact on neutronics criticality and the ratio of 238Pu/Pu discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    FRANCIS,A.J.

    2006-10-18

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

  14. ALMA Reveals the Molecular Medium Fueling the Nearest Nuclear Starburst

    Science.gov (United States)

    Leroy, Adam K.; Bolatto, Alberto D.; Ostriker, Eve C.; Rosolowsky, Erik; Walter, Fabian; Warren, Steven R.; Donovan Meyer, Jennifer; Hodge, Jacqueline; Meier, David S.; Ott, Jürgen; Sandstrom, Karin; Schruba, Andreas; Veilleux, Sylvain; Zwaan, Martin

    2015-03-01

    We use ALMA observations to derive mass, length, and time scales associated with NGC 253's nuclear starburst. This region forms ~2 M ⊙ yr-1 of stars and resembles other starbursts in ratios of gas, dense gas, and star formation tracers, with star formation consuming the gas reservoir at a normalized rate 10 times higher than in normal galaxy disks. We present new ~35 pc resolution observations of bulk gas tracers (CO), high critical density transitions (HCN, HCO+, and CS), and their isotopologues. The starburst is fueled by a highly inclined distribution of dense gas with vertical extent factor implied by our cloud calculations is approximately Galactic, contrasting with results showing a low value for the whole starburst region. The contrast provides resolved support for the idea of mixed molecular ISM phases in starburst galaxies.

  15. 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.)

  16. 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.

  17. National briefing summaries: Nuclear fuel cycle and waste management

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

    1988-12-01

    The National Briefing Summaries is a compilation of publicly available information concerning the nuclear fuel cycle and radioactive waste management strategies and programs of 21 nations, including the United States and three international agencies that have publicized their activities in this field. It presents available highlight information with references that may be used by the reader for additional information. The information in this document is compiled primarily for use by the US Department of Energy and other US federal agencies and their contractors to provide summary information on radioactive waste management activities in other countries. This document provides an awareness to managers and technical staff of what is occurring in other countries with regard to strategies, activities, and facilities. The information may be useful in program planning to improve and benefit United States' programs through foreign information exchange. Benefits to foreign exchange may be derived through a number of exchange activities.

  18. Selection of materials in nuclear fuel: present and future; Seleccion de materiales en el combustible nuclear: presente y futuro

    Energy Technology Data Exchange (ETDEWEB)

    Munoz-Reja, C.; Fuentes, L.; Garcia de la Infanta, J. M.; Munoz Sicilia, A.

    2013-07-01

    One of the main aspects of the nuclear fuel is the selection of materials for the components. The operating conditions of the fuel elements impose a major challenge to materials: high temperature, corrosive aqueous environment, high mechanical properties, long periods of time under these extreme conditions and what is the differentiating factor; the effect of irradiation. The materials are selected to fulfill these severe requirements and also to be able to control and to predict its behavior in the working conditions. Their development, in terms of composition and processing, is based on the continuous follow-up of the operation behavior. Many of these materials are specific of the nuclear industry, such as the uranium dioxide and the zirconium alloys. This article presents the selection and development of the nuclear fuel materials as a function of the services requirements. It also includes a view of the new nuclear fuels materials that are being raised after Fukushima accident. (Author)

  19. The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    Science.gov (United States)

    Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

    2017-01-01

    The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

  20. 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...

  1. Nuclear Cryogenic Propulsion Stage (NCPS) Fuel Element Testing in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    Science.gov (United States)

    Emrich, William J., Jr.

    2017-01-01

    To satisfy the Nuclear Cryogenic Propulsion Stage (NCPS) testing milestone, a graphite composite fuel element using a uranium simulant was received from the Oakridge National Lab and tested in the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) at various operating conditions. The nominal operating conditions required to satisfy the milestone consisted of running the fuel element for a few minutes at a temperature of at least 2000 K with flowing hydrogen. This milestone test was successfully accomplished without incident.

  2. Sustainable Nuclear Fuel Cycles and World Regional Issues

    Directory of Open Access Journals (Sweden)

    Aleksandra Schwenk-Ferrero

    2012-06-01

    Full Text Available In the present paper we have attempted to associate quantified impacts with a forecasted nuclear energy development in different world regions, under a range of hypotheses on the energy demand growth. It gives results in terms of availability of uranium resources, required deployment of fuel cycle facilities and reactor types. In particular, the need to achieve short doubling times with future fast reactors is investigated and quantified in specific world regions. It has been found that a crucial feature of any world scenario study is to provide not only trends for an idealized “homogeneous” description of the global world, but also trends for different regions in the world. These regions may be selected using rather simple criteria (mostly of a geographical type, in order to apply different hypotheses for energy demand growth, fuel cycle strategies and the implementation of various reactor types for the different regions. This approach was an attempt to avoid focusing on selected countries, in particular on those where no new significant energy demand growth is expected, but instead to provide trends and conclusions that account for the features of countries that will be major players in the world energy development in the future.

  3. Measurement of nuclear fuel pin hydriding utilizing epithermal neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

    Miller, W.H. [Univ. of Missouri, Columbia, MO (United States); Farkas, D.M.; Lutz, D.R. [General Electric Co., Pleasanton, CA (United States)

    1996-12-31

    The measurement of hydrogen or zirconium hydriding in fuel cladding has long been of interest to the nuclear power industry. The detection of this hydrogen currently requires either destructive analysis (with sensitivities down to 1 {mu}g/g) or nondestructive thermal neutron radiography (with sensitivities on the order of a few weight percent). The detection of hydrogen in metals can also be determined by measuring the slowing down of neutrons as they collide and rapidly lose energy via scattering with hydrogen. This phenomenon is the basis for the {open_quotes}notched neutron spectrum{close_quotes} technique, also referred to as the Hysen method. This technique has been improved with the {open_quotes}modified{close_quotes} notched neutron spectrum technique that has demonstrated detection of hydrogen below 1 {mu}g/g in steel. The technique is nondestructive and can be used on radioactive materials. It is proposed that this technique be applied to the measurement of hydriding in zirconium fuel pins. This paper summarizes a method for such measurements.

  4. 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.

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

    Directory of Open Access Journals (Sweden)

    Nick R. Soelberg

    2013-01-01

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

  6. Model development for quantitative evaluation of proliferation resistance of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

    This study addresses the quantitative evaluation of the proliferation resistance which is important factor of the alternative nuclear fuel cycle system. In this study, model was developed to quantitatively evaluate the proliferation resistance of the nuclear fuel cycles. The proposed models were then applied to Korean environment as a sample study to provide better references for the determination of future nuclear fuel cycle system in Korea. In order to quantify the proliferation resistance of the nuclear fuel cycle, the proliferation resistance index was defined in imitation of an electrical circuit with an electromotive force and various electrical resistance components. The analysis on the proliferation resistance of nuclear fuel cycles has shown that the resistance index as defined herein can be used as an international measure of the relative risk of the nuclear proliferation if the motivation index is appropriately defined. It has also shown that the proposed model can include political issues as well as technical ones relevant to the proliferation resistance, and consider all facilities and activities in a specific nuclear fuel cycle (from mining to disposal). In addition, sensitivity analyses on the sample study indicate that the direct disposal option in a country with high nuclear propensity may give rise to a high risk of the nuclear proliferation than the reprocessing option in a country with low nuclear propensity.

  7. Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Lippek, H.E.; Schuller, C.R.

    1979-03-01

    A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.

  8. 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.

  9. Software Design Document for the AMP Nuclear Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Philip, Bobby [ORNL; Clarno, Kevin T [ORNL; Cochran, Bill [ORNL

    2010-03-01

    The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

  10. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  11. International Source Book: Nuclear Fuel Cycle Research and Development Vol 1 Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K. M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lakey, L. T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    1983-07-01

    This document starts with an overview that summarizes nuclear power policies and waste management activities for nations with significant commercial nuclear fuel cycle activities either under way or planned. A more detailed program summary is then included for each country or international agency conducting nuclear fuel cycle and waste management research and development. This first volume includes the overview and the program summaries of those countries listed alphabetically from Argentina to Italy.

  12. Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Massaro, Lawrence M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-09-30

    A preliminary evaluation of removing spent nuclear fuel (SNF) from 13 shutdown nuclear power reactor sites was conducted. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, and Vermont Yankee. The evaluation was divided into four components: (1) characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory, (2) a description of the on-site infrastructure and conditions relevant to transportation of SNF and GTCC waste, (3) an evaluation of the near-site transportation infrastructure and experience relevant to shipping transportation casks containing SNF and GTCC waste, including identification of gaps in information, and (4) an evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. Every site was found to have at least one off-site transportation mode option for removing its SNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an important source of information used to identify the transportation mode options for the sites. Especially important in conducting the evaluation were site visits, through which information was obtained that would not have been available otherwise. Extensive photographs taken during the site visits proved to be particularly useful in documenting the current conditions at or near the sites. It is expected that additional site visits will be conducted to add to the information presented in the evaluation.

  13. Global nuclear energy partnership fuels transient testing at the Sandia National Laboratories nuclear facilities : planning and facility infrastructure options.

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, John E.; Wright, Steven Alan; Tikare, Veena; MacLean, Heather J. (Idaho National Laboratory, Idaho Falls, ID); Parma, Edward J., Jr.; Peters, Curtis D.; Vernon, Milton E.; Pickard, Paul S.

    2007-10-01

    The Global Nuclear Energy Partnership fuels development program is currently developing metallic, oxide, and nitride fuel forms as candidate fuels for an Advanced Burner Reactor. The Advance Burner Reactor is being designed to fission actinides efficiently, thereby reducing the long-term storage requirements for spent fuel repositories. Small fuel samples are being fabricated and evaluated with different transuranic loadings and with extensive burnup using the Advanced Test Reactor. During the next several years, numerous fuel samples will be fabricated, evaluated, and tested, with the eventual goal of developing a transmuter fuel database that supports the down selection to the most suitable fuel type. To provide a comparative database of safety margins for the range of potential transmuter fuels, this report describes a plan to conduct a set of early transient tests in the Annular Core Research Reactor at Sandia National Laboratories. The Annular Core Research Reactor is uniquely qualified to perform these types of tests because of its wide range of operating capabilities and large dry central cavity which extents through the center of the core. The goal of the fuels testing program is to demonstrate that the design and fabrication processes are of sufficient quality that the fuel will not fail at its design limit--up to a specified burnup, power density, and operating temperature. Transient testing is required to determine the fuel pin failure thresholds and to demonstrate that adequate fuel failure margins exist during the postulated design basis accidents.

  14. 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.

  15. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    Science.gov (United States)

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  16. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    Energy Technology Data Exchange (ETDEWEB)

    Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

    2009-10-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  17. 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.

  18. 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.

  19. Nuclear Fuel Cycle Analysis by Integrated AHP and TOPSIS Method Using an Equilibrium Model

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, S. R. [University of Science and Technology, Daejeon (Korea, Republic of); Choi, S. Y. [UNIST, Ulju (Korea, Republic of); Koc, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Determining whether to break away from domestic conflict surrounding nuclear power and step forward for public consensus can be identified by transparent policy making considering public acceptability. In this context, deriving the best suitable nuclear fuel cycle for Korea is the key task in current situation. Assessing nuclear fuel cycle is a multicriteria decision making problem dealing with multiple interconnected issues on efficiently using natural uranium resources, securing an environment friendliness to deal with waste, obtaining the public acceptance, ensuring peaceful uses of nuclear energy, maintaining economic competitiveness compared to other electricity sources, and assessing technical feasibility of advanced nuclear energy systems. This paper performed the integrated AHP and TOPSIS analysis on three nuclear fuel cycle options against 5 different criteria including U utilization, waste management, material attractiveness, economics, and technical feasibility. The fuel cycle options analyzed in this paper are three different fuel cycle options as follows: PWR-Once through cycle(PWR-OT), PWR-MOX cycle, Pyro- SFR cycle. These fuel cycles are most likely to be adopted in the foreseeable future. Analytic Hierarchy Process (AHP) and TOPSIS (Technique for Order of Preference by Similarity to Ideal Solution). The analyzed nuclear fuel cycle options include the once-through cycle, the PWR-MOX recycle, and the Pyro-SFR recycle.

  20. 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

  1. Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

    Energy Technology Data Exchange (ETDEWEB)

    Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

    2011-11-14

    For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

  2. Design and manufacturing of 05F-01K instrumented capsule for nuclear fuel irradiation in Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, J. M.; Shin, Y. T.; Park, S. J. (and others)

    2007-07-15

    An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in Hanaro. The instrumented capsule(02F-11K) for measuring and monitoring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. It was successfully irradiated in the test hole OR5 of Hanaro from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the instrumented capsule(03F-05K) were designed and manufactured to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. This capsule was irradiated in the test hole OR5 of Hanaro reactor from April 26, 2004 to October 1, 2004 (59.5 EFPD at 24 {approx} 30 MW). The six typed dual instrumented fuel rods, which allow for two characteristics to be measured simultaneously in one fuel rod, have been designed and manufactured to enhance the efficiency of the irradiation test using the instrumented fuel capsule. The 05F-01K instrumented fuel capsule was designed and manufactured for a design verification test of the three dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of Hanaro.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-01

    Most of the aqueous processes developed, or under consideration worldwide for the recycling of used nuclear fuel (UNF) utilize the oxido-reduction properties of actinides to separate them from other radionuclides. Generally, after acid dissolution of the UNF, (essentially in nitric acid solution), actinides are separated from the raffinate by liquid-liquid extraction using specific solvents, associated along the process, with a particular reductant that will allow the separation to occur. For example, the industrial PUREX process utilizes hydroxylamine as a plutonium reductant. Hydroxylamine has numerous advantages: not only does it have the proper attributes to reduce Pu(IV) to Pu(III), but it is also a non-metallic chemical that is readily decomposed to innocuous products by heating. However, it has been observed that the presence of high nitric acid concentrations or impurities (such as metal ions) in hydroxylamine solutions increase the likelihood of the initiation of an autocatalytic reaction. Recently there has been some interest in the application of simple hydrophilic hydroxamic ligands such as acetohydroxamic acid (AHA) for the stripping of tetravalent actinides in the UREX process flowsheet. This approach is based on the high coordinating ability of hydroxamic acids with tetravalent actinides (Np and Pu) compared with hexavalent uranium. Thus, the use of AHA offers a route for controlling neptunium and plutonium in the UREX process by complexant based stripping of Np(IV) and Pu(IV) from the TBP solvent phase, while U(VI) ions are not affected by AHA and remain solvated in the TBP phase. In the European GANEX process, AHA is also used to form hydrophilic complexes with actinides and strip them from the organic phase into nitric acid. However, AHA does not decompose completely when treated with nitric acid and hampers nitric acid recycling. In lieu of using AHA in the UREX + process, formohydroxamic acid (FHA), although not commercially available, hold

  4. Nuclear fuel cycle assessment of India: A technical study for U.S.-India cooperation

    Science.gov (United States)

    Krishna, Taraknath Woddi Venkat

    The recent civil nuclear cooperation proposed by the Bush Administration and the Government of India has heightened the necessity of assessing India's nuclear fuel cycle inclusive of nuclear materials and facilities. This agreement proposes to change the long-standing U.S. policy of preventing the spread of nuclear weapons by denying nuclear technology transfer to non-NPT signatory states. The nuclear tests in 1998 have convinced the world community that India would never relinquish its nuclear arsenal. This has driven the desire to engage India through civilian nuclear cooperation. The cornerstone of any civilian nuclear technological support necessitates the separation of military and civilian facilities. A complete nuclear fuel cycle assessment of India emphasizes the entwinment of the military and civilian facilities and would aid in moving forward with the separation plan. To estimate the existing uranium reserves in India, a complete historical assessment of ore production, conversion, and processing capabilities was performed using open source information and compared to independent reports. Nuclear energy and plutonium production (reactor- and weapons-grade) was simulated using declared capacity factors and modern simulation tools. The three-stage nuclear power program entities and all the components of civilian and military significance were assembled into a flowsheet to allow for a macroscopic vision of the Indian fuel cycle. A detailed view of the nuclear fuel cycle opens avenues for technological collaboration. The fuel cycle that grows from this study exploits domestic thorium reserves with advanced international technology and optimized for the existing system. To utilize any appreciable fraction of the world's supply of thorium, nuclear breeding is necessary. The two known possibilities for production of more fissionable material in the reactor than is consumed as fuel are fast breeders or thermal breeders. This dissertation analyzes a thermal

  5. Reactor Fuel Isotopics and Code Validation for Nuclear Applications

    Energy Technology Data Exchange (ETDEWEB)

    Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-02-01

    Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for 133Cs and 153Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

  6. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed [Egyptian Atomic Energy Authority, Cairo (Egypt)

    2013-07-01

    The operations with the fissile materials such as U{sup 235} introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k{sub eff}) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed.

  7. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    E. R. Johnson; R. E. Best

    2009-12-28

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

  9. Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

    Science.gov (United States)

    Kozier, K. S.; Dyck, G. R.

    2006-04-01

    A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in alternate channels. Results were compared using different room-temperature data files for deuterium, various thermal-scattering-law data files for hydrogen bound in light water and deuterium bound in heavy water, and for pre-ENDF/B-VII and ENDF/B-VI.8 data for uranium. The reactivity differences observed were small (typically <1 mk) and increased with axial neutron leakage.

  10. On feasibility of a closed nuclear power fuel cycle with minimum radioactivity

    Science.gov (United States)

    Andrianova, E. A.; Davidenko, V. D.; Tsibulskiy, V. F.

    2015-12-01

    Practical implementation of a closed nuclear fuel cycle implies solution of two main tasks. The first task is creation of environmentally acceptable operating conditions of the nuclear fuel cycle considering, first of all, high radioactivity of the involved materials. The second task is creation of effective and economically appropriate conditions of involving fertile isotopes in the fuel cycle. Creation of technologies for management of the high-level radioactivity of spent fuel reliable in terms of radiological protection seems to be the hardest problem.

  11. Speciation, in the nuclear fuel cycle by spectroscopic techniques

    Energy Technology Data Exchange (ETDEWEB)

    Colette, S.; Plancque, G.; Allain, F.; Lamouroux, C.; Steiner, V.; Amekraz, B.; Moulin, C. [CEA/Saclay, Dept, des Procedes d' Enrichissement (DPE), 91 - Gif-sur-Yvette (France)

    2000-07-01

    New analytical techniques allowing to perform speciation in the framework of the nuclear fuel cycle are more and more needed. They have to be selective (since matrix encountered are very complex), sensitive (in order to work at representative concentration and below solubility limit), as well as non intrusive (in order to keep the image of the real solution). Among them, laser-based analytical techniques present these advantages together with the possibility to perform remote measurements via fiber optics. Hence, Time-Resolved Laser-Induced Fluorescence (TRLIF) has been used for actinides/lanthanides interaction and speciation studies in inorganic and organic matrices from the reprocessing to waste storage. Moreover, new ion detection methods such as Electro-Spray - Mass Spectrometry (ES-MS) seems promising for speciation studies. Hence, it is the first time that it is possible to directly couple a liquid at atmospheric pressure to a mass detection working at reduced pressure with a soft mode of ionisation that should allow to give informations on chemical species present. Principle, advantages and limitations as well as results obtained with the use of TRLIF and ES-MS on different systems of interest including actinides, lanthanides, fission products in interaction with simple organic molecules to very complex structure will be presented and discussed. (authors)

  12. 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.

  13. 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 al