WorldWideScience

Sample records for fuel damage accident

  1. Characteristics of severely damaged fuel from PBF tests and the TMI-2 accident

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cook, B.A.; Dallman, R.J.; Broughton, J.M.

    1986-01-01

    As a result of the TMI-2 reactor accident, the US Nuclear Regulatory Commission initiated a research program to investigate phenomena associated with severe fuel damage accidents. This program is sponsored by several countries and includes in-pile and out-of-pile experiments, separate effects studies, and computer code development. The principal in-pile testing portion of the program includes four integral severe fuel damage (SFD) tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The INEL is also responsible for examining the damaged core in the Three Mile Island-Unit 2 (TMI-2) reactor, which offers the unique opportunity to directly compare the findings of an experimental program to those of an actual reactor accident. The principal core damage phenomena which can occur during a severe accident are discussed, and examples from the INEL research programs are used to illustrate the characteristics of these phenomena. The preliminary results of the programs are presented, and their impact on plant operability during severe accidents is discussed

  2. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    Directory of Open Access Journals (Sweden)

    Igor L. Kozlov

    2015-03-01

    Full Text Available Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific thermal-hydrodynamic conditions and mechanisms of explosive situations arrival at different stages of a severe accident development, is substantiated. Based on the analysis of the known shell-type nuclear reactors accidents results the explosion safety thermodynamic criteria are presented, the parameters defining the steam and gas explosions conditions are found, the need to perform the further verification and validation of deterministic codes serving to simulate general accident processes behavior as well as phase-to-phase interaction calculated dependencies is established. The main parameters controlling and defining the criteria explosion safety effective regulation areas and their optimization conditions are found.

  3. Grain boundary sweeping and dissolution effects on fission product behaviour under severe fuel damage accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1986-01-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behaviour considers the migration and coalescence of fission gas bubbles in either molten uranium, or a Zircaloy-Uranium eutectic melt. Results of the analyses demonstrate that intragranular fission product behavior during the tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. Whereas fuel liquefaction leads to an enhanced release of fission products in trace-irradiated fuel, the occurrence of fuel liquefaction in normally-irradiated fuel can degrade fission product release. This phenomenon is due in part to reduced gas-bubble mobilities in a viscous medium as compared to vapor transport, and in part to a degradation of grain growth rates and the subsequent decrease in grain-boundary sweeping of intragranular fission products into the liquified lamina. The analysis shows that total UO 2 dissolution due to eutectic melting leads to increased release for both trace-irradiated and normally-irradiated fuel. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally

  4. Grain boundary sweeping and dissolution effects on fission product behavior under severe fuel damage accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1985-10-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and during U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. The grain-boundary-sweeping theory considers the interaction between the moving grain boundary and two distinct size classes of bubbles, those on grain faces and on grain edges. The theory of the effects of fuel liquefaction and U-Zr eutectic melting on fission product behavior considers the migration and coalescence of fission gas bubbles in either molten uranium, or a zircaloy-uranium eutectic melt. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in normally irradiated fuel are highlighted

  5. Severe fuel damage projects

    International Nuclear Information System (INIS)

    Sdouz, G.

    1987-10-01

    After the descriptions of the generation of a Severe Fuel Damage Accident in a LWR the hypothetical course of such an accident is explained. Then the most significant projects are described. At each project the experimental facility, the most important results and the concluding models and codes are discussed. The selection of the projects is concentrated on the German Projekt Nukleare Sicherheit (PNS), tests performed at the Idaho National Engineering Laboratory (INEL) and smaller projects in France and Great Britain. 25 refs., 26 figs. (Author)

  6. Evaluation of fuel rod damage in LWR under accident conditions using SSYST

    International Nuclear Information System (INIS)

    Meyder, R.

    1982-01-01

    After a short outline of the recent SSYST-development, the creep rupture model NORA 2 is presented. The effect of temperature and oxygen on Zircaloy 4 creep behaviour is shown. Examples on the effect of azimuthal varying gap width and wall thickness are given. Remarks on the extension of a single rod analysis on a bundle and the stepwise application of SSYST for investigation of fuel rod failure conclude the paper. (orig.) [de

  7. Features of RAPTA-SFD code modelling of chemical interactions of basic materials of the WWER active zone in accident conditions with severe fuel damage

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Sokolov, N.B.; Salatov, A.V.; Nechaeva, O.A.; Andreyeva-Andrievskaya, L.N.; Vlasov, F.Yu.

    1996-01-01

    A brief description of RAPTA-SFD code intended for computer simulations of WWER-type fuel elements (simulator or absorber element) in conditions of accident with severe damage of fuel. Presented are models of chemical interactions of basic materials of the active zone, emphasized are special feature of their application in carrying out of the CORA-W2 experiment within the framework of International Standard Problem ISP-36. Results obtained confirm expediency of phenomenological models application. (author). 6 refs, 7 figs, 1 tab

  8. Grain boundary sweeping and liquefaction-induced fission product behavior in nuclear fuel under severe-core damage accident conditions

    International Nuclear Information System (INIS)

    Rest, J.

    1984-05-01

    The theoretical FASTGRASS-VFP model has been used in the interpretation of fission gas, iodine, tellurium, and cesium release from: (1) irradiated high-burnup LWR fuel in a flowing steam atmosphere during high-temperature, in-cell heating tests performed at Oak Ridge National Laboratory; and (2) trace-irradiated and high-burnup LWR fuel during severe-fuel-damage (SFD) tests performed in the PBF reactor in Idaho. A theory of grain boundary sweeping of gas bubbles, gas bubble behavior during fuel liquefaction (destruction of grain boundaries due to formation of a U-rich melt phase), and U-Zr eutectic melting has been included within the FASTGRASS-VFP formalism. Results of the analyses demonstrate that intragranular fission product behavior during both types of tests can be interpreted in terms of a grain-growth/grain-boundary-sweeping mechanism that enhances the flow of fission products from within the grains to the grain boundaries. Whereas fuel liquefaction leads to an enhanced release of fission products in trace-irradiated fuel, the occurrence of fuel liquefaction in high-burnup fuel can degrade fission product release. This phenomenon is due in part to reduced gas-bubble mobilities in a viscous medium as compared to vapor transport, and in part to a degradation of grain growth rates and the subsequent decrease in grain-boundary sweeping of intragranular fission products into the liquefied lamina. The analysis shows that total UO 2 dissolution due to eutectic melting leads to increased release for both trace-irradiated and high-burnup fuel. The FASTGRASS-VFP predictions, measured release rates from the above tests, and previously published release rates are compared and differences between fission product behavior in trace-irradiated and in high-burnup fuel are highlighted

  9. Nuclear fuel in a reactor accident.

    Science.gov (United States)

    Burns, Peter C; Ewing, Rodney C; Navrotsky, Alexandra

    2012-03-09

    Nuclear accidents that lead to melting of a reactor core create heterogeneous materials containing hundreds of radionuclides, many with short half-lives. The long-lived fission products and transuranium elements within damaged fuel remain a concern for millennia. Currently, accurate fundamental models for the prediction of release rates of radionuclides from fuel, especially in contact with water, after an accident remain limited. Relatively little is known about fuel corrosion and radionuclide release under the extreme chemical, radiation, and thermal conditions during and subsequent to a nuclear accident. We review the current understanding of nuclear fuel interactions with the environment, including studies over the relatively narrow range of geochemical, hydrological, and radiation environments relevant to geological repository performance, and discuss priorities for research needed to develop future predictive models.

  10. Accident tolerant fuel analysis

    International Nuclear Information System (INIS)

    2014-01-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant

  11. Probability of spent fuel transportation accidents

    International Nuclear Information System (INIS)

    McClure, J.D.

    1981-07-01

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

  12. The composition of aerosols generated during a severe reactor accident: Experimental results from the Power Burst Facility Severe Fuel Damage Test 1-4

    International Nuclear Information System (INIS)

    Petti, D.A.; Hobbins, R.R.; Hagrman, D.L.

    1994-01-01

    Experimental results on fission product and aerosol release during the Power Burst Facility Severe Fuel Damages (SFD) Test 1-4 are examined to determine the composition of aerosols that would be generated during a severe reactor accident. The SFD 1-4 measured aerosol contained significant quantities of volatile fission products (VFPs) (cesium, iodine, tellurium), control materials (silver and cadmium), and structural materials (tin), indicating that fission product release, vaporization of control material, and release of tin from oxidized Zircaloy were all important aerosol sources. On average the aerosol composition is between one-quarter and one-half VFPs (especially cesium), with the remainder being control material (especially cadmium), and structural material (especially tin). Source term computer codes like CORSOR-M tend to overpredict the release of structural and control rod material relative to fission products by a factor of between 2 and 15 because the models do not account for relocation of molten control, fuel, and structural material during the degradation process, which tends to reduce the aerosol source. The results indicate that the aerosol generation in a severe reactor accident is intimately linked to the core degradation process. They recommend that these results be used to improve the models in source term computer codes

  13. Mechanisms of damage to the oxide layer of cladding of fuel rods under accident conditions like RI

    International Nuclear Information System (INIS)

    Busser, Vincent

    2009-01-01

    During reactivity initiated accident, the importance of cladding tube oxidation on its thermomechanical behavior has been investigated. After RIA tests in experimental reactors oxide damage including radial cracking and spallation of the outer oxide layer has been evidenced. This work aims at better understanding the key mechanisms controlling these phenomena. Laboratory air-oxidation of Zircaloy-4 cladding tubes has been performed at 470 C. SEM micrographs show that radial cracks are initiated from the outer surface of the oxide layer and propagated radially towards the oxide-metal interface. A model predicting the stress evolution within the oxide and the depth of crack has been developed and validated on literature tests and tests of this study. Ring compression tests were used for the experimental study of the oxide degradation under mechanical loading. Experimental data revealed three mechanisms: densification of the radial crack network, propagation of these radial cracks, branching and spallation of oxide fragments. The influence of the circumferential cracks, periodically distributed in the oxide layer, on the stress distribution in oxide fragments has been analysed using finite element modelling. The determining influence of these cracks on the maximum stress oxide fragments has been demonstrated. (author)

  14. Modelling Accident Tolerant Fuel Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Gamble, Kyle Allan Lawrence [Idaho National Laboratory

    2016-05-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether either of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced

  15. Criticality accident of nuclear fuel facility. Think back on JCO criticality accident

    International Nuclear Information System (INIS)

    Naito, Keiji

    2003-09-01

    This book is written in order to understand the fundamental knowledge of criticality safety or criticality accident of nuclear fuel facility by the citizens. It consists of four chapters such as critical conditions and criticality accident of nuclear facility, risk of criticality accident, prevention of criticality accident and a measure at an occurrence of criticality accident. A definition of criticality, control of critical conditions, an aspect of accident, a rate of incident, damage, three sufferers, safety control method of criticality, engineering and administrative control, safety design of criticality, investigation of failure of safety control of JCO criticality accident, safety culture are explained. JCO criticality accident was caused with intention of disregarding regulation. It is important that we recognize the correct risk of criticality accident of nuclear fuel facility and prevent disasters. On the basis of them, we should establish safety culture. (S.Y.)

  16. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    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

  17. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

  18. Behavior of fission products released from severely damaged fuel during the PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hagrman, D.L.; Broughton, J.M.; Rest, J.

    1984-01-01

    The results of fission product release behavior during the first two Power Burst Facility Severe Fuel Damage tests are presented. Measured fission product release is compared with calculated release using temperature dependent release rate correlations and FASTGRASS analysis. The test results indicate that release from fuel of the high volatility fission products (Xe, Kr, I, Cs, and Te) is strongly influenced by parameters other than fuel temperature; namely fuel/fission product morphology, fuel and cladding oxidation state, extent of fuel liquefaction, and quench induced fuel shattering. Fission product transport from the test fuel through the sample system was strongly influenced by chemical effects. Holdup of I and Cs was affected by fission product chemistry, and transport time while Te release was primarily influenced by the extent of zircaloy oxidation. Analysis demonstrates that such integral test data can be used to confirm physical, chemical, and mechanistic models of fission product behavior for severe accident conditions

  19. Accident analysis of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.; Chi, D. Y

    1998-03-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

  20. Accident tolerant fuels for LWRs: A perspective

    Energy Technology Data Exchange (ETDEWEB)

    Zinkle, S.J., E-mail: zinklesj@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States); Terrani, K.A.; Gehin, J.C.; Ott, L.J.; Snead, L.L. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

    2014-05-01

    The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

  1. Accident tolerant fuels for LWRs: A perspective

    International Nuclear Information System (INIS)

    Zinkle, S.J.; Terrani, K.A.; Gehin, J.C.; Ott, L.J.; Snead, L.L.

    2014-01-01

    The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms

  2. Spent fuel transportation accident: a state's involvement

    International Nuclear Information System (INIS)

    Neuweg, M.

    1978-01-01

    On February 9, 1978 at 8:20 p.m., the duty officer for the Illinois Radiological Assistance Team was notified that a shipment containing uranium and plutonium was involved in an accident near Gibson City, Illinois on Route 54. It was reported that a pig containing an unknown amount of uranium and plutonium was involved. The Illinois District 6A State Police were called to the scene and secured the area. The duty officer in the meantime learned after numerous telephone calls, approximately 1 hour after the first notice was received, that the pig actually was a 48,000 pound cask containing 6 spent fuel rods and the tractor-trailer had split apart and was blocking one lane of the highway. The shipment had departed from Dresden Nuclear Power Station, Morris, Illinois, enroute to Babcox and Wilcox in Lynchburg, Virginia. Initial reports indicated the vehicle had split apart. Actually, the semi-trailer bed had buckled beneath the cask due to apparent excess stress. The cask remained entirely intact and was not damaged, but the state highway was closed to traffic. The State Radiological Assistance Team was dispatched and arrived on the scene at 12:45 a.m. Immediate radiation monitoring revealed a reading of 4 milliroentgen per hour at 10 feet from the cask. No contamination existed nor was anyone exposed to radiation unnecessarily. The cask was transferred to a Tri-State semi-trailer vehicle the following morning at approximately 6:30 a.m. At 9:30 a.m., February 10, the new vehicle was again enroute to its destination. This incident demonstrated typical occurrences involving transportation radiation accident: misinformation and/or lack of information on the initial response notification, inaccuracies of radiation monitorings at the scene of the accident, inconsistencies concerning the occurrences of the accident and unfamiliar terminology utilized by personnel first on the scene, i.e., pig, cask, vehicle split apart, etc

  3. PBF severe fuel damage program: results and comparison to analysis

    International Nuclear Information System (INIS)

    McDonald, P.E.; Buescher, B.J.; Gruen, G.E.; Hobbins, R.R.; McCardell, R.K.

    1983-01-01

    The United States Nuclear Regulatory Commission has initiated a severe fuel damage research program in the Power Burst Facility (PBF) to investigate fuel rod and core response, and fission product and hydrogen release and transport under degraded core cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

  4. Severe fuel-damage scoping test performance

    International Nuclear Information System (INIS)

    Gruen, G.E.; Buescher, B.J.

    1983-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle

  5. Power Burst Facility severe-fuel-damage test program

    International Nuclear Information System (INIS)

    McCardell, R.K.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

  6. Analysis of severe core damage accident progression for the heavy water reactor

    International Nuclear Information System (INIS)

    Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

    2010-01-01

    In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

  7. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

    Fields, S.R.

    1975-12-01

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel

  8. Enhanced accident-tolerant fuel (EATF)

    International Nuclear Information System (INIS)

    Strumpell, John

    2013-01-01

    The Fukushima accident provided a strong reminder that the exothermic reaction between zirconium and steam, and the attendant hydrogen generation, can significantly affect the course of a severe accident. Part of the response to the accident was increased interest in the extent to which the fuel itself can mitigate the consequences of a severe accident. Improved fuel alone is not sufficient to provide the desired increase in reactor safety, but it can provide an important contribution. With support from the US Department of Energy, AREVA has brought together a team that includes researchers (AREVA, Electric Power Research Institute, Savannah River National Laboratory, University of Florida, and University of Wisconsin), a fuel vendor (AREVA), and utilities (Duke Energy and Tennessee Valley Authority). The goal of the project is to develop new technologies that can be deployed in a lead assembly within ten years. The researchers have proposed a variety of approaches for improving the performance of the fuel, including new cladding and structural materials, fuel pellets with improved thermal characteristics, and coatings on the fuel rods. The expected performance of fuels that apply these technologies will be judged against the requirements of the vendor and utilities to determine those that are most promising for immediate development and those that may be suited for development in the future. The first review will consider the manufacturability of the proposed designs; the second will focus on performance. Materials that are suitable for immediate development will be considered for irradiation in a test reactor and subsequent use in lead assembly designs

  9. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    1984-03-01

    The Severe Core Damage Research and Analysis Task Force was established in Nuclear Safety Research Center, Tokai Research Establishment, JAERI, in May, 1982 to make a quantitative analysis on the issues related with the severe core damage accident and also to survey the present status of the research and provide the required research subjects on the severe core damage accident. This report summarizes the results of the works performed by the Task Force during last one and half years. The main subjects investigated are as follows; (1) Discussion on the purposes and necessities of severe core damage accident research, (2) proposal of phenomenological research subjects required in Japan, (3) analysis of severe core damage accidents and identification of risk dominant accident sequences, (4) investigation of significant physical phenomena in severe core damage accidents, and (5) survey of the research status. (author)

  10. Nuclear Fuel Behaviour during Reactivity Initiated Accidents. Workshop Proceedings

    International Nuclear Information System (INIS)

    2010-01-01

    A reactivity initiated accident (RIA) is a nuclear reactor accident that involves an unwanted increase in fission rate and reactor power. The power increase may damage the reactor core. The main objective of the workshop was to review the current status of the experimental and analytical studies of the fuel behavior during the RIA transients in PWR and BWR reactors and the acceptance criteria for RIA in use and under consideration. The workshop was organized in an opening session and 5 technical sessions: 1) Recent experimental results and experimental techniques used; 2) Modelling and Data Interpretation; 3) Code Assessment; 4) RIA Core Analysis and 5) Revision and application of safety criteria

  11. Fuel Behaviour at High During RIA and LOCA Accidents

    International Nuclear Information System (INIS)

    Barrio del Juanes, M. T.; Garcia Cuesta, J. C.; Vallejo Diaz, I.; Herranz Puebla

    2001-01-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs

  12. Analysis of the kinetic behaviour of iodine and caesium isotopes in the primary circuit of LWR's during severe fuel damage accidents

    International Nuclear Information System (INIS)

    Alonso, A.; Fernandez, S.; Buron, J.M.; Lopez, J.V.

    1991-01-01

    This State of the Art report deals with the chemical behaviour of caesium and iodine in the primary system, focusing particularly on kinetic chemical aspects. In case of a postulated severe accident in a nuclear reactor, cesium and iodine fission products are among the major contributors to health harm because of their high volatility and radiotoxicity. The extent of the release of such fission products to the environment depends on the effectiveness of transport through different structures in the reactor coolant system and within the reactor building. The release from fuel has been briefly studied; only those aspects concerning to iodine and caesium chemical forms when released have been reviewed; nevertheless the emphasis has been put on the transport of such elements and their species through the primary system. Some thermochemical equilibrium studies, applied to primary circuit conditions in LWR's, have been analyzed. The revision of the few kinetic studies existing on this matter has shown that kinetic behaviour of iodine and caesium isotopes in the primary circuit is an aspect poorly studied, despite the fact that kinetic aspects could have great importance on the chemical species formed under certain conditions. Other phenomena affecting iodine and caesium transport, besides chemical reactions, such as interactions with surfaces, aerosols or other chemical species have also been examined from available information on diverse experiments

  13. Determinants of the property damage costs of tanker accidents

    International Nuclear Information System (INIS)

    Talley, W.K.

    1999-01-01

    This study investigates determinants of the vessel, oil cargo spillage, and other-property damage costs of tanker accidents. Tobit estimation of a three-equation recursive model suggests that, among types of tanker accidents, fire/explosion accidents incur the largest vessel damage costs, but the smallest oil cargo spillage costs. Alternatively, grounding accidents incur the smallest vessel damage costs, but the largest oil cargo spillage costs, reflecting the difficulty of controlling oil cargo spillage subsequent to such accidents. Also, oil cargo spillage costs are lower for US flag tanker accidents. A dollar of vessel damage cost increases other-property damage cost by 0.06 dollars, whereas a dollar of oil cargo spillage increases this cost by 1.55 dollars

  14. Improving performance with accident tolerant-fuels

    International Nuclear Information System (INIS)

    Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

    2017-01-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO 2 - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO 2 - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U 3 Si 2 , UN, and UC contain a higher uranium density and thermal conductivity than UO 2 , providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U 3 Si 2 , UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO 2 . These ATFs also have thermal conductivities approximately four times higher than that of UO 2 . A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO 2 - Zr as the fuel system of choice. (author)

  15. Truck accident involving unirradiated nuclear fuel

    International Nuclear Information System (INIS)

    Carlson, R.W.; Fischer, L.E.

    1992-07-01

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 nuclear fuel assemblies in 12 containers on Interstate 1-91 in Springfield, Massachusetts. This paper documents the mechanical circumstances of the accident and the physical environment to which the containers were exposed and the response of the containers and their contents. The accident involved four impacts where the truck was struck by the car, impacted on the center guardrail, impacted on the outer concrete barrier and came to rest against the center guardrail. The impacts were followed by a fire that began in the engine compartment, spread to the.tractor and cab, and eventually spread to the trailer and payload. The fire lasted for about three hours and the packages were involved in the fire for about two hours. As a result of the fire, the tractor-trailer was completely destroyed and the packages were exposed to flames with temperatures between 1300 degrees F and 1800 degrees F. The fuel assemblies remained intact during the accident and there was no release of any radioactive material during the accident. This was a very severe accident; however, the injuries were minor and at no time was the public health and safety at risk

  16. Truck accident involving unirradiated nuclear fuel

    International Nuclear Information System (INIS)

    Carlson, R.W.; Fischer, L.E.

    1993-01-01

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 unirradiated nuclear fuel assemblies in 12 containers on Interstate I-91 in Springfield, Massachusetts. This paper documents the mechanical circumstances of the accident and assesses the physical environment to which the containers were exposed and the response of the containers and their contents. The accident involved four impacts where the truck was struck by the car, impacted on the center guardrail, impacted on the outer concrete barrier and came to rest against the center guardrail. The impacts were followed by a fire that began in the engine compartment, spread to the tractor and cab, and eventually spread to the trailer and payload. The fire lasted for about three hours and the packages were involved in the fire for about two hours. As a result of the fire, the tractor-trailer was completely destroyed and the packages were exposed to flames with temperatures between 1,300 F and 1,800 F. The fuel assemblies remained intact during the accident and there was no release of any radioactive material during the accident. This was a very severe accident; however, the injuries were minor and at no time was the public health and safety at risk

  17. Instrumentation needs in LWR severe fuel damage experiments

    International Nuclear Information System (INIS)

    McCormick, R.D.

    1980-01-01

    The Class 9 type nuclear accident is defined and the Three Mile Island type accident and proposed Idaho National Engineering Laboratory experiment series are described in some detail. Different types of severe fuel damage experiments are briefly discussed in order to show typical measurement requirements. General instrumentation needs and problems encountered in Class 9 accident research are outlined. It is concluded that the extremely high temperatures, high nuclear radiation fields, and oxidizing atmosphere will necessitate instrument development programs. Noncontact type sensing will be necessary in most of the molten core experiments

  18. Improving performance with accident tolerant-fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

    2017-07-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO{sub 2} - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO{sub 2} - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U{sub 3}Si{sub 2}, UN, and UC contain a higher uranium density and thermal conductivity than UO{sub 2}, providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U{sub 3}Si{sub 2}, UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO{sub 2}. These ATFs also have thermal conductivities approximately four times higher than that of UO{sub 2}. A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO{sub 2} - Zr as the fuel system of choice. (author)

  19. Compensation for damages in case of a nuclear accident

    International Nuclear Information System (INIS)

    Leger, M.

    2011-01-01

    This article presents the system of compensation for damages in case of a nuclear accident. This system of civil liability for nuclear damage, as a specific regime, departs on several points from the common rules of civil liability, in order to provide an adequate and equitable compensation for the damages suffered by the victims of nuclear accidents. The French system of civil liability for nuclear damage results from two International Conventions integrated in French law (Paris convention 1960 and Brussels convention 1963) and the French law of 1968, October 30 on civil liability in the area of nuclear energy. These texts define the conditions under which a nuclear operator could be held liable in case of a nuclear accident. The protocols to amend the Paris and Brussels Conventions of 2004, not yet come into force, are also presented. They ensure that increased resources are available to compensate a greater number of victims of a nuclear accident. (author)

  20. The compensation of damage in Germany following the Chernobyl accident

    International Nuclear Information System (INIS)

    Eich, W.

    2003-01-01

    In the framework of the workshop on the indemnification of damage in the event of a nuclear accident, this paper presents the proceeding of the the discussion on the compensation of damage in Germany following the Chernobyl accident. This paper presents also the national experiences and opinions, a documentation of the Federal Office of Administration on the topic, the example of Tokai-mura accident third party liability and compensation and the third party liability in the field of nuclear law in Ireland. (A.L.B.)

  1. Management of severely damaged nuclear fuel and related waste

    International Nuclear Information System (INIS)

    1991-01-01

    This report is concerned primarily with severe fuel damage accidents in large electric power producing reactors such as those in the TMI and Chernobyl plants. It does include, as appropriate, knowledge gained from accidents in other power, research and military reactors. It is believed that the conclusions and recommendations apply to a large extent to severe fuel damage accidents in all types of reactors. The period considered in this publication begins after the initial crisis of an accident has been brought under control. (This initial crisis could be from one day to several weeks after the event, depending on the specific conditions). Accordingly, it is assumed that the plant is shut down, the reactor is under control and decay heat removal is in progress in a stable manner so that attention must be given to cleanup. This report addresses the principles involved in planning, engineering, construction, operation and other activities to characterize, clean up and dispose of the fuel and related waste. The end of the period under consideration is when the fuel and abnormal wastes are packaged either for interim storage or final disposal and activities are started either to restore the plant to service or to establish a safe state from which decommissioning planning can start. 36 refs, 3 figs, 4 tabs.

  2. Accident situations tests HTR fuel with the device Kufa

    International Nuclear Information System (INIS)

    Kellerbauer, A. I.; Freis, D.

    2010-01-01

    The ceramic and ceramic-like coating materials in modern high-temperature reactor fuel are designed to ensure mechanical stability and retention of fission products under normal and transient conditions, regardless of the radiation damage sustained in-pile. In hypothetical depressurization and loss-of-forced-circulation (D LOFC) accidents, fuel elements of modular high-temperate reactors are exposed to temperatures several hundred degrees higher than during normal operation, causing increased thermo-mechanical stress on the coating layers. At the Institute for Transuranium Elements of the European Commission, a vigorous experimental program is being pursued with the aim of characterizing the performance of irradiated HTR fuel under such accident conditions. A cold finger device (Kufa), operational in ITUs hot cells since 2006, has been used to perform heating experiments on eight irradiated HTR fuel pebbles from the AVR experimental reactor and from dedicated irradiation campaigns at the High-Flux Reactor in Petten, the Netherlands. Gaseous fission products are collected in a cryogenic charcoal trap, while volatiles,are plated out on a water-cooled condensate plate. A quantitative measurement of the release is obtained by gamma spectroscopy. We highlight experimental results from the Kufa testing as well as the on-going development of new experimental facilities. (Author) 9 refs.

  3. Assessment of WWER fuel condition in design basis accident

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.; Sokolov, N.; Andreeva-Andrievskaya, L.; Vlasov, Yu.; Nechaeva, O.; Salatov, A.

    1994-01-01

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR's. 12 figs., 11 refs

  4. Assessment of WWER fuel condition in design basis accident

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu; Sokolov, N; Andreeva-Andrievskaya, L; Vlasov, Yu; Nechaeva, O; Salatov, A [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation)

    1994-12-31

    The fuel behaviour in design basis accidents is assessed by means of the verified code RAPTA-5. The code uses a set of high temperature physico-chemical properties of the fuel components as determined for commercially produced materials, fuel rod simulators and fuel rod bundles. The WWER fuel criteria available in Russia for design basis accidents do not generally differ from the similar criteria adopted for PWR`s. 12 figs., 11 refs.

  5. Compensation for damage in the case of transfrontier reactor accidents

    International Nuclear Information System (INIS)

    Gornig, G.

    1986-01-01

    The author discusses possibilities to recover in German and Soviet courts claims for the compensation of damage for a German citizen arising from the reactor accident in Chernobyl. Concerning the claims for damage suffered in the Federal Republic of Germany he investigates possible breaches of bilateral or multilateral international agreements and of universal international law by the Soviet Union. (WG) [de

  6. Civil Cases Proof Peculiarities of Road Traffic Accidents Damages

    Directory of Open Access Journals (Sweden)

    Polyakov D. N.

    2012-11-01

    Full Text Available The author reveals proof peculiarities in civil cases of reparation of damages harmed by road traffic accident, in relation to the determination of a respondent (debtor. In the article are analyzed the appropriate norms of the RF Civil code regulating the rules and conditions of civil liability for damage caused by using a transport facility as a source of danger

  7. Economic damage caused by a nuclear reactor accident

    International Nuclear Information System (INIS)

    Goemans, T.; Schwarz, J.J.

    1988-01-01

    This study is directed towards the estimation of the economic damage which arises from a severe possible accident with a newly built 1000 MWE nuclear power plant in the Netherlands. A number of cases have been considered which are specified by the weather conditions during and the severity of the accident and the location of the nuclear power plant. For each accident case the economic damage has been estimated for the following impact categories: loss of the power plant, public health, evacuation and relocation of population, export of agricultural products, working and living in contaminated regions, decontamination, costs of transportation and incoming foreign tourism. The consequences for drinking water could not be quantified adequately. The total economic damage could reach 30 billion guilders. Besides the power plant itself, loss of export and decreasing incoming foreign tourism determine an important part of the total damage. 12 figs.; 52 tabs

  8. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  9. Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

  10. The American 'severe fuel damage program'

    International Nuclear Information System (INIS)

    Sdouz, G.

    1982-03-01

    The TMI-2 accident has initiated a new phase of safety research. It is necessary to consider severe accidents with degraded or molten core. For NRC there was a need for an improved understanding of this reactor behaviour and the 'Severe Fuel Dage Program' was initiated. Planned are in-pile experiments in PBF, NRU and ESSOR and in addition separate effects tests and results from TMI-2. The analytical component of the program is the development of different versions of the code SCDAP for the detailed analysis during severe accident transients. (Author) [de

  11. Review of progress on enhanced accident tolerant fuel

    International Nuclear Information System (INIS)

    McCoy, K.; Dunn, B.; Kochendarfer, R.

    2015-01-01

    The accident at Fukushima has resulted in renewed interest in understanding the performance of nuclear power plants under accident conditions. Part of that interest is directed toward determining how to improve the performance of fuel during an accident that involves long exposures of the fuel to high temperatures. This paper describes the method being used by AREVA to select and evaluate approaches for improving the accident tolerance of nuclear fuel. The method involves starting with a large number of approaches that might enhance accident tolerance, and reviewing how well each approach satisfies a set of engineering requirements and goals. Among the approaches investigated we have the development of fuel pellets that contain a second phase to improve thermal conductivity, the use of molybdenum alloy tubing as fuel cladding, the use of oxidation-resistant coatings to zirconium cladding, and the use of nanoparticles in the coolant to improve heat transfer

  12. PBF Severe Fuel-Damage Program: results and comparison to analysis

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Buescher, B.J.; Hobbins, R.R.; McCardell, R.K.; Gruen, G.E.

    1983-01-01

    The United States Nuclear Regulatory Commission has initiated a severe fuel-damage research program in the Power Burst Facility (PBF) to investigate fuel-rod and core response, and fission-product and hydrogen release and transport under degraded-core-cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

  13. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M T; Garcia Cuesta, J C; Vallejo Diaz, I; Puebla, Herranz

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  14. Fuel Behaviour at High During RIA and LOCA Accidents; Comportamiento del Combustible de Alto Quemado en Accidents RIA y LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Barrio del Juanes, M.T.; Garcia Cuesta, J.C.; Vallejo Diaz, I.; Herranz Puebla

    2001-07-01

    Safety analysis of high burnup fuel requires ensuring the acceptable performance under design basis accidents, in particular during conditions representative of Reactivity Accidents (RIA) and Loss-of-Coolant Accidents (LOCA). The report's objective is to compile the state of the art on these issues. This is mainly focused in the effort made to define the applicability of safety criteria to the high burnup fuel. Irradiation damage modifies fuel rod properties, thus the probability of fuel to withstand thermal and mechanical loads during an accident could be quite different compared with unirradiated fuel. From the thermal point of view, fuel conductivity is the most affected property, decreasing notably with irradiation. From the mechanical point of view, a change in the pellet microstructure at its periphery is observed at high burnup (remiffect). Cladding is also effected during operation, showing a significant external and internal corrosion. All these phenomena result in the decrease of efficiency in heat transfer an in the reduction of capability to accommodate mechanical loads; this situation is especially significant at high burnup, when pellet-cladding mechanical interaction is present. Knowledge about these phenomena is not possible without appropriate experimental programmes. The most relevant have been performed in France, Japan, United States and Russia. Results obtained with fuel at high burnup show significant differences with respect to the phenomena observed in fuel at the present discharge burnup. Indeed, this is the encouragement to research about this occurrence. This study is framed within the CSN-CIEMAT agreement, about Fuel Thermo-Mechanical Behaviour at High Burnup. (Author) 172 refs.

  15. Fuel temperature analysis method for channel-blockage accident in HTTR

    International Nuclear Information System (INIS)

    Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Hayakawa, Hitoshi

    1994-01-01

    During operation of the High Temperature Engineering Test Reactor (HTTR), coolability must be maintained without core damage under all postulated accident conditions. Channel blockage of a fuel element was selected as one of the design-basis accidents in the safety evaluation of the reactor. The maximum fuel temperature for such a scenario has been evaluated in the safety analysis and is compared to the core damage limits.For the design of the HTTR, an in-core thermal and hydraulic analysis code ppercase[flownet/trump] was developed. This code calculates fuel temperature distribution, not only for a channel blockage accident but also for transient conditions. The validation of ppercase[flownet/trump] code was made by comparison of the analytical results with the results of thermal and hydraulic tests by the Helium Engineering Demonstration Loop (HENDEL) multi-channel test rig (T 1-M ), which simulated one fuel column in the core. The analytical results agreed well with the experiments in which the HTTR operating conditions were simulated.The maximum fuel temperature during a channel blockage accident is 1653 C. Therefore, it is confirmed that the integrity of the core is maintained during a channel blockage accident. ((orig.))

  16. Calculation of spent fuel pool severe accident with MELCOR

    International Nuclear Information System (INIS)

    Deng Jian; Xiang Qing'an; Zhou Kefeng

    2014-01-01

    A calculation model was established for spent fuel pool (SFP) using MELCOR code to study the severe accident phenomena caused by the long term station black-out (SBO), including spent fuel heatup, zirconium cladding oxidation, and the injection into SFP to mitigate the severe accident. The results show that the severe accident progression is slow and relates directly with the initial water level in SFP. It is illustrated that the injection into SFP is one of the best mitigated measures for the SFP severe accident. (authors)

  17. Credible accident analyses for TRIGA and TRIGA-fueled reactors

    International Nuclear Information System (INIS)

    Hawley, S.C.; Kathren, R.L.

    1982-04-01

    Credible accidents were developed and analyzed for TRIGA and TRIGA-fueled reactors. The only potential for offsite exposure appears to be from a fuel-handling accident that, based on highly conservative assumptions, would result in dose equivalents of less than or equal to 1 mrem to the total body from noble gases and less than or equal to 1.2 rem to the thyroid from radioiodines. Credible accidents from excess reactivity insertions, metal-water reactions, lost, misplaced, or inadvertent experiments, core rearrangements, and changes in fuel morphology and ZrH/sub x/ composition are also evaluated, and suggestions for further study provided

  18. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  19. Fuel behaviour in the case of severe accidents and potential ATF designs. Fuel Behavior in Severe Accidents and Potential Accident Tolerance Fuel Designs

    International Nuclear Information System (INIS)

    Cheng, Bo

    2013-01-01

    This presentation reviews the conditions of fuel rods under severe loss of coolant conditions, approaches that may increase coping time for plant operators to recover, requirements of advanced fuel cladding to increase tolerance in accident conditions, potential candidate alloys for accident-tolerant fuel cladding and a novel design of molybdenum (Mo) -based fuel cladding. The current Zr-alloy fuel cladding will lose all its mechanical strength at 750-800 deg. C, and will react rapidly with high-pressure steam, producing significant hydrogen and exothermic heat at 700-1000 deg. C. The metallurgical properties of Zr make it unlikely that modifications of the Zr-alloy will improve the behaviour of Zr-alloys at temperatures relevant to severe accidents. The Mo-based fuel cladding is designed to (1) maintain fuel rod integrity, and reduce the release rate of hydrogen and exothermic heat in accident conditions at 1200-1500 deg. C. The EPRI research has thus far completed the design concepts, demonstration of feasibility of producing very thin wall (0.2 mm) Mo tubes. The feasibility of depositing a protective coating using various techniques has also been demonstrated. Demonstration of forming composite Mo-based cladding via mechanical reduction has been planned

  20. Melcor benchmarking against integral severe fuel damage tests

    Energy Technology Data Exchange (ETDEWEB)

    Madni, I.K. [Brookhaven National Lab., Upton, NY (United States)

    1995-09-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the U.S. Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC to provide independent assessment of MELCOR, and a very important part of this program is to benchmark MELCOR against experimental data from integral severe fuel damage tests and predictions of that data from more mechanistic codes such as SCDAP or SCDAP/RELAP5. Benchmarking analyses with MELCOR have been carried out at BNL for five integral severe fuel damage tests, namely, PBF SFD 1-1, SFD 14, and NRU FLHT-2, analyses, and their role in identifying areas of modeling strengths and weaknesses in MELCOR.

  1. Historical update of past and recent skin damage radiation accidents

    International Nuclear Information System (INIS)

    Lushbaugh, C.C.; Fry, S.A.; Ricks, R.C.; Hubner, K.F.; Burr, W.W.

    1986-01-01

    Records of radiation accidents worldwide since 1944 are maintained at the Radiation Accident Registry of the Radiation Emergency Assistance Center/Training Site (REAC/TS) in Oak Ridge. These records show that in 263 major radiation accidents there have been 150 severe local radiation injuries, of which 117 have been exposure to sealed radioactive sources. Most lesions resulted from the unsafe handling of 192 Ir radiography sources. Recent redesign of these devices, used for testing the integrity of welds, promises to eliminate these accidents. However, many other kinds of irradiators used in industry and scientific research still remain in the public domain, capable of causing irreparable dermal damage. Registry records reveal many unsolved physical and medical problems whose solution is urgently needed to improve the prognosis and therapy of such lesions. Pathologically, radiation-induced skin lesions are well described and an approximate dose-response relationship is univerally accepted even though the actual 'dose' is rarely known at first. Radiation dose is estimated biologically after the lesion has run its pathological course or after a medical physicist has prepared a retrospective 'mock-up' of the accident. (author)

  2. Development of Accident Scenario for Interim Spent Fuel Storage Facility Based on Fukushima Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongjin; Choi, Kwangsoon; Yoon, Hyungjoon; Park, Jungsu [KEPCO-E and C, Yongin (Korea, Republic of)

    2014-05-15

    700 MTU of spent nuclear fuel is discharged from nuclear fleet every year and spent fuel storage is currently 70.9% full. The on-site wet type spent fuel storage pool of each NPP(nuclear power plants) in Korea will shortly exceed its storage limit. Backdrop, the Korean government has rolled out a plan to construct an interim spent fuel storage facility by 2024. However, the type of interim spent fuel storage facility has not been decided yet in detail. The Fukushima accident has resulted in more stringent requirements for nuclear facilities in case of beyond design basis accidents. Therefore, there has been growing demand for developing scenario on interim storage facility to prepare for beyond design basis accidents and conducting dose assessment based on the scenario to verify the safety of each type of storage.

  3. Survey of potential light water reactor fuel rod failure mechanisms and damage limits

    International Nuclear Information System (INIS)

    Courtright, E.L.

    1979-07-01

    The findings and conclusions are presented of a survey to evaluate current information applicable to the development of fuel rod damage and failure limits for light water reactor fuel elements. The survey includes a review of past fuel failures, and identifies potential damage and failure mechanisms for both steady state operating conditions and postulated accident events. Possible relationships between the various damage and failure mechanisms are also proposed. The report identifies limiting criteria where possible, but concludes that sufficient data are not currently available in many important areas

  4. Analysis of tritium mission FMEF/FAA fuel handling accidents

    Energy Technology Data Exchange (ETDEWEB)

    Van Keuren, J.C.

    1997-11-18

    The Fuels Material Examination Facility/Fuel Assembly Area is proposed to be used for fabrication of mixed oxide fuel to support the Fast Flux Test Facility (FFTF) tritium/medical isotope mission. The plutonium isotope mix for the new mission is different than that analyzed in the FMEF safety analysis report. A reanalysis was performed of three representative accidents for the revised plutonium mix to determine the impact on the safety analysis. Current versions computer codes and meterology data files were used for the analysis. The revised accidents were a criticality, an explosion in a glovebox, and a tornado. The analysis concluded that risk guidelines were met with the revised plutonium mix.

  5. A highway accident involving unirradiated nuclear fuel in Springfield, Massachusetts, on December 16, 1991

    International Nuclear Information System (INIS)

    Carlson, R.W.; Fischer, L.E.

    1992-06-01

    In the early morning of Dec. 16, 1991, a severe accident occurred when a passenger vehicle traveling in the wrong direction collided with a tractor trailer carrying 24 unirradiated nuclear fuel assemblies in 12 containers on Interstate I-91 in Springfield, Massachusetts. The purpose of this report is to document the mechanical circumstances of the severe accident, confirm the nature and quantity of the radioactive materials involved, and assess the physical environment to which the containers were exposed and the response of the containers and their contents. The report consists of five major sections. The first section describes the circumstances and conditions of the accident and the finding of facts. The second describes the containers, the unirradiated nuclear fuel assemblies, and the tie down arrangement used for the trailer. The third describes the damage sustained during the accident to the tractor, trailer, containers, and unirradiated nuclear fuel assemblies. The fourth evaluates the accident environment and its effects on the containers and their contents. The final section gives conclusions derived from the analysis and fact finding investigation. During this severe accident, only minor injuries occurred, and at no time was the public health and safety at risk

  6. Behaviour of molten reactor fuels under accident conditions

    International Nuclear Information System (INIS)

    Xavier Swamikannu, A.; Mathews, C.K.

    1980-01-01

    The behaviour of molten reactor fuels under accident conditions has received considerable importance in recent times. The chemical processes that occur in the molten state among the fuel, the clad components and the concrete of the containment building under the conditions of a core melt down accident in oxide fuelled reactors have been reviewed with the purpose of identifying areas of developmental work required to be performed to assess and minimize the consequences of such an accident. This includes the computation and estimation of vapour pressure of various gaseous species over the fuel, the clad and the coolant, providing of sacrificial materials in the concrete in order to protect the containment building in order to prevent release of radioactive gases into the atmosphere and understanding the distribution and chemical state of fission products in the molten fuel in order to provide for the effective removal of their decay heats. (auth.)

  7. Environmental damage caused by fossil fuels consumption

    International Nuclear Information System (INIS)

    Barbir, F.; Veziroglu, T.N.

    1991-01-01

    This paper reports that the objectives of this study is to identify the negative effects of the fossil fuels use and to evaluate their economic significance. An economic value of the damage for each of the analyzed effects has been estimated in US dollars per unit energy of the fuel used ($/GJ). This external costs of fossil fuel use should be added to their existing market price, and such real costs should be compared with the real costs of other, environmentally acceptable, energy alternatives, such as hydrogen

  8. Researches of WWER fuel rods behaviour under RIA accident conditions

    International Nuclear Information System (INIS)

    Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

    2003-01-01

    Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

  9. Deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Parsons, P.D.; Hindle, E.D.; Mann, C.A.

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail.

  10. The deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Parsons, P.D.; Hindle, E.D.; Mann, C.A.

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail. (author)

  11. Safety demonstration analyses at JAERI for severe accident during overland transport of fresh nuclear fuel

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Kitao, Kohichi; Karasawa, Kiyonori; Yamada, Kenji; Takahashi, Satoshi; Watanabe, Kohji; Okuno, Hiroshi; Miyoshi, Yoshinori

    2005-01-01

    It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted in a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident postulated to occur during transportation, for the purpose of gaining acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and thus, accident conditions leading to mechanical damages and thermal failure were determined to characterize the scenarios. As a result, the worst-case conditions of run-off-the-road accidents were set up to define the impact against a concrete or asphalt surface. For fire accident scenarios to be set up, collisions were assumed to occur with an oil tanker carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside a tunnel without ventilation. Then the cask models were determined for these safety demonstration analyses to represent those commonly used for fresh nuclear fuel transported throughout Japan. Following the postulated accident scenarios, the mechanical damages were analyzed by using the general-purpose finite element code LS-DYNA with three-dimensional elements. It was found that leak tightness of the package be maintained even in the severe impact scenario. Then the thermal safety was analyzed by using the general-purpose finite element code ABAOUS with three-dimensional elements to describe cask geometry. As a result of the thermal analyses, the integrity of the containment

  12. Severe accident analysis and management in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Golshan, Mina

    2013-01-01

    Within the UK regulatory regime, assessment of risks arising from licensee's activities are expected to cover both normal operations and fault conditions. In order to establish the safety case for fault conditions, fault analysis is expected to cover three forms of analysis: design basis analysis (DBA), probabilistic safety assessment (PSA) and severe accident analysis (SAA). DBA should provide a robust demonstration of the fault tolerance of the engineering design and the effectiveness of the safety measures on a conservative basis. PSA looks at a wider range of fault sequences (on a best estimate basis) including those excluded from the DBA. SAA considers significant but unlikely accidents and provides information on their progression and consequences, within the facility, on the site and off site. The assessment of severe accidents is not limited to nuclear power plants and is expected to be carried out for all plant states where the identified dose targets could be exceeded. This paper sets out the UK nuclear regulatory expectation on what constitutes a severe accident, irrespective of the type of facility, and describes characteristics of severe accidents focusing on nuclear fuel cycle facilities. Key rules in assessment of severe accidents as well as the relationship to other fault analysis techniques are discussed. The role of SAA in informing accident management strategies and offsite emergency plans is covered. The paper also presents generic examples of scenarios that could lead to severe accidents in a range of nuclear fuel cycle facilities. (authors)

  13. Report on damaged FLIP TRIGA fuel

    International Nuclear Information System (INIS)

    Feltz, Donald E.; Randall, John D.; Schumacher, Robert F.

    1977-01-01

    Damaged FLIP elements were discovered, positioned adjacent to the transient rod. It then became apparent that this was not the failure of a defective, element but a heretofore unknown operating or design problem. The damaged elements are described as having bulges in the cladding and unevenly spaced dark rings along the fuelled portion of the element. Possible causes are investigated, including: defective fuel elements, incorrectly calculated power distributions in the core and in the elements, water leakage into the void follower of the transient rod, and improper safety limit for FLIP fuel. Based on measurements and calculations that have been experimentally verified it is concluded that the safety limit was not exceeded or even closely approached. It is also concluded that the problem is due entirely due to some phenomena occurring during pulsing, and that the steady state history of the fuel is not a factor

  14. In-pool damaged fuel bundle recovery

    International Nuclear Information System (INIS)

    Piascik, T.G.; Patenaude, R.S.

    1988-01-01

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power / Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved

  15. In-pool damaged fuel bundle recovery

    International Nuclear Information System (INIS)

    Piascik, T.G.; Patenaude, R.S.

    1988-01-01

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package, and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power/Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved. In spite of several emergent problems which a task of this nature presents, this small, close knit utility/vendor team completed the work on schedule and within the exposure and cost budgets

  16. Development of Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

  17. Transportation accident response of a high-capacity truck cask for spent fuel

    International Nuclear Information System (INIS)

    O'Connell, W.J.; Glaser, R.E.; Johnson, G.L.; Perfect, S.A.; McGuinn, E.J.; Lake, W.H.

    1995-11-01

    Two of the primary goals of this study were (i) to check the structural and thermal performance of the GA-4 cask in a broad range of accidents and (ii) to carry out a severe-accidents analysis as had been addressed in the Modal Study but now using a specific recent cask design and using current-generation computer models and capabilities. At the same time, it was desired to compare the accident performance of the Ga-4 cask to that of the generic truck cask analyzed in the Modal Study. The same range of impact and fire accidents developed in the Modal Study was adopted for this study. The accident-description data base of the Modal Study categorizes accidents into types of collisions with mobile or fixed objects, non-collision accidents, and fires. The mechanical modes of damage may be via crushing, impact, or puncture. The fire occurrences in the Modal Study data are based on truck accident statistics. The fire types are taken to be pool fires of petroleum products from fuel tanks and/or cargoes

  18. Multiscale Multiphysics Developments for Accident Tolerant Fuel Concepts

    International Nuclear Information System (INIS)

    Gamble, K. A.; Hales, J. D.; Yu, J.; Zhang, Y.; Bai, X.; Andersson, D.; Patra, A.; Wen, W.; Tome, C.; Baskes, M.; Martinez, E.; Stanek, C. R.; Miao, Y.; Ye, B.; Hofman, G. L.; Yacout, A. M.; Liu, W.

    2015-01-01

    U 3 Si 2 and iron-chromium-aluminum (Fe-Cr-Al) alloys are two of many proposed accident-tolerant fuel concepts for the fuel and cladding, respectively. The behavior of these materials under normal operating and accident reactor conditions is not well known. As part of the Department of Energy's Accident Tolerant Fuel High Impact Problem program significant work has been conducted to investigate the U 3 Si 2 and FeCrAl behavior under reactor conditions. This report presents the multiscale and multiphysics effort completed in fiscal year 2015. The report is split into four major categories including Density Functional Theory Developments, Molecular Dynamics Developments, Mesoscale Developments, and Engineering Scale Developments. The work shown here is a compilation of a collaborative effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory and Anatech Corp.

  19. Full-length fuel rod behavior under severe accident conditions

    International Nuclear Information System (INIS)

    Lombardo, N.J.; Lanning, D.D.; Panisko, F.E.

    1992-12-01

    This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors

  20. Indemnification of damage in the event of a nuclear accident

    International Nuclear Information System (INIS)

    2003-01-01

    The Workshop on the Indemnification of Damage in the Event of a Nuclear Accident, organised by the OECD Nuclear Energy Agency in close co-operation with the French authorities, was held in Paris from 26 to 28 November 2001. This event was an integral part of the International Nuclear Emergency Exercise INEX 2000. It attracted wide participation from national nuclear authorities, regulators, operators of nuclear installations, nuclear insurers and international organisations. The objective was to test the capacity of the existing nuclear liability and compensation mechanisms in the 29 countries represented at the workshop to manage the consequences of a nuclear emergency. This workshop was based upon the scenario used for the INEX 2000 Exercise, i.e. an accident simulated at the Gravelines nuclear power plant in the north of France in May 2001. These proceedings contain a comparative analysis of legislative and regulatory provisions governing emergency response and nuclear third party liability, based upon country replies to a questionnaire. This publication also includes the full responses provided to that questionnaire, as well as the texts of presentations made by special guests from Germany and Japan describing the manner in which the public authorities in their respective countries responded to two nuclear accidents of a very different nature and scale. (authors)

  1. EPRI nuclear fuel-cycle accident risk assessment

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The present results of the nuclear fuel-cycle accident risk assessment conducted by the Electric Power Research Institute show that the total risk contribution of the nuclear fuel cycle is only approx. 1% of the accident risk of the power plant; hence, with little error, the accident risk of nuclear electric power is essentially that of the power plant itself. The power-plant risk, assuming a very large usage of nuclear power by the year 2005 is only approx. 0.5% of the radiological risk of natural background. The smallness of the fuel-cycle risk relative to the power-plant risk may be attributed to the lack of internal energy to drive an accident and the small amount of dispersible material. This work aims at a realistic assessment of the process hazards, the effectiveness of confinement and mitigation systems and procedures, and the associated likelihood of errors and the estimated size of errors. The primary probabilistic estimation tool is fault-tree analysis, with the release source terms calculated using physicochemical processes. Doses and health effects are calculated with CRAC (Consequences of Reactor Accident Code). No evacuation or mitigation is considered; source terms may be conservative through the assumption of high fuel burnup (40,000 MWd/t) and short cooling period (90 to 150 d); high-efficiency particulate air filter efficiencies are derived from experiments

  2. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and

  3. Accident tolerant fuel cladding development: Promise, status, and challenges

    Science.gov (United States)

    Terrani, Kurt A.

    2018-04-01

    The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

  4. Catalogue of methods, tools and techniques for recovery from fuel damage events

    International Nuclear Information System (INIS)

    1991-10-01

    On the basis of the recommendations of the Advisory Group Meeting on Main Principles of Safe Management of Severely Damaged Nuclear Fuel and other Accident Generated Waste, held from 13 to 16 November 1989, the IAEA initiated a programme in 1990 to collect technical information on special tools and methods to deal with circumstances beyond the normal design basis of fuel damage. A Questionnaire was sent out to solicit information from the Member States and organizations which might have experience in this field. The responses to the Questionnaire were discussed at a Consultants Meeting and at an Advisory Group Meeting during 1990. The aim of this document is to disseminate the experience gained in Member States serving Article 5 of the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency and also filling a potential void in response to fuel damage events of less severe magnitude

  5. Modeling of in-vessel fission product release including fuel morphology effects for severe accident analyses

    International Nuclear Information System (INIS)

    Suh, K.Y.

    1989-10-01

    A new in-vessel fission product release model has been developed and implemented to perform best-estimate calculations of realistic source terms including fuel morphology effects. The proposed bulk mass transfer correlation determines the product of fission product release and equiaxed grain size as a function of the inverse fuel temperature. The model accounts for the fuel-cladding interaction over the temperature range between 770 K and 3000 K in the steam environment. A separate driver has been developed for the in-vessel thermal hydraulic and fission product behavior models that were developed by the Department of Energy for the Modular Accident Analysis Package (MAAP). Calculational results of these models have been compared to the results of the Power Burst Facility Severe Fuel Damage tests. The code predictions utilizing the mass transfer correlation agreed with the experimentally determined fractional release rates during the course of the heatup, power hold, and cooldown phases of the high temperature transients. Compared to such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation, the mass transfer correlation resulted in lower and less rapid releases in closer agreement with the on-line and grab sample data from the Severe Fuel Damage tests. The proposed mass transfer correlation can be applied for best-estimate calculations of fission products release from the UO 2 fuel in both nominal and severe accident conditions. 15 refs., 10 figs., 2 tabs

  6. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

    2017-07-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  7. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    International Nuclear Information System (INIS)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

    2017-01-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  8. Status of USNRC research on fuel behavior under accident conditions

    International Nuclear Information System (INIS)

    Johnston, W.V.

    1976-01-01

    The program of the Fuel Behaviour Research is directed at providing a detailed understanding of the response of nuclear fuel assemblies to off-normal or accident conditions. This understanding is expressed in physical and analytical correlations which are incorporated into computer codes. The results of these experiments and the resulting codes are available to the licensing authorities for use in evaluating utility submissions. (orig.) [de

  9. A methodology for the evaluation of fuel rod failures under transportation accidents

    International Nuclear Information System (INIS)

    Rashid, J.Y.R.; Machiels, A.J.

    2004-01-01

    Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

  10. Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gilles Youinou; R. Sonat Sen

    2013-09-01

    The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

  11. Use of fuel failure correlations in accident analysis

    International Nuclear Information System (INIS)

    O'Dell, L.D.; Baars, R.E.; Waltar, A.E.

    1975-05-01

    The MELT-III code for analysis of a Transient Overpower (TOP) accident in an LMFBR is briefly described, including failure criteria currently applied in the code. Preliminary results of calculations exploring failure patterns in time and space in the reactor core are reported and compared for the two empirical fuel failure correlations employed in the code. (U.S.)

  12. A new NEA expert group on accident-tolerant fuels

    International Nuclear Information System (INIS)

    Massara, Simone

    2014-01-01

    After the events at the Fukushima Daiichi nuclear power plant in March 2011, enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion. One outcome of those discussions has been to promote research into the development of advanced fuels and more robust reactor system technologies with improved performance, reliability and safety characteristics during normal operations and under accident conditions. The Fukushima Daiichi accident has highlighted in particular the importance of reducing hydrogen production rates and increasing fission product retention during extended loss of cooling accidents. In this context, the NEA organised two international workshops to share information and discuss technical and safety issues associated with the development of accident-tolerant fuels (ATFs) for LWRs. Presentations were given by experts from various organisations, industry and regulatory bodies of NEA member countries, as well as from representatives of international bodies. The presentations focused on lessons learnt from the Fukushima Daiichi accident, the desired characteristics of ATFs, potential design options and candidate materials, as well as the current state of the art in related modelling and simulation methods. During discussions following these workshop presentations, delegates agreed to establish a collaborative framework on ATFs within the NEA. Reporting to the Nuclear Science Committee, the Expert Group on Accident-tolerant Fuels for Light Water Reactors (EGATFL) will define and carry out a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with more enhanced accident tolerance compared to currently used zircaloy/UO 2 fuels. The group will foster information exchange on material properties and relevant phenomenological experiments, carry out state-of-the-art reviews, organise benchmark studies and foster international

  13. Safety demonstration analyses for severe accident of fresh nuclear fuel transport packages at JAERI

    International Nuclear Information System (INIS)

    Yamada, K.; Watanabe, K.; Nomura, Y.; Okuno, H.; Miyoshi, Y.

    2004-01-01

    It is expected in the near future that more and more fresh nuclear fuel will be transported in a variety of transport packages to cope with increasing demand from nuclear fuel cycle facilities. Accordingly, safety demonstration analyses of these methods are planned and conducted at JAERI under contract with the Ministry of Economy, Trade and Industry of Japan. These analyses are conducted part of a four year plan from 2001 to 2004 to verify integrity of packaging against leakage of radioactive material in the case of a severe accident envisioned to occur during transportation, for the purpose of gaining public acceptance of such nuclear fuel activities. In order to create the accident scenarios, actual transportation routes were surveyed, accident or incident records were tracked, international radioactive material transport regulations such as IAEA rules were investigated and, thus, accident conditions leading to mechanical damage and thermal failure were selected for inclusion in the scenario. As a result, the worst-case conditions of run-off-the-road accidents were incorporated, where there is impact against a concrete or asphalt surface. Fire accidents were assumed to occur after collision with a tank truck carrying lots of inflammable material or destruction by fire after collision inside a tunnel. The impact analyses were performed by using three-dimensional elements according to the general purpose impact analysis code LS-DYNA. Leak-tightness of the package was maintained even in the severe impact accident scenario. In addition, the thermal analyses were performed by using two-dimensional elements according to the general purpose finite element method computer code ABAQUS. As a result of these analyses, the integrity of the inside packaging component was found to be sufficient to maintain a leak-tight state, confirming its safety

  14. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  15. Mechanical injuries, burns and combination damage caused by reactor accidents

    International Nuclear Information System (INIS)

    Koslowski, L.

    1981-01-01

    In cases of combination damage the initial treatment of wounds is the same as with injuries without accompanying radiation exposure. In the beginning the general principles of surgical treatment apply. In case of a mass accident, the examination of the injured to decide on the necessary kind of treatment has priority. A common problem to all the decisions is that the extent of a radiation exposure that may have been sustained cannot be established at once. Whether the radiation exposure has been so heavy as to require the modification of the surgical measures can be seen only from the blood count, the bone marrow biopsy, the reticulocyte count or from a chromosome analysis. (DG) [de

  16. Indemnification of Damage in the Event of a Nuclear Accident

    International Nuclear Information System (INIS)

    2006-01-01

    The Second International Workshop on the Indemnification of Nuclear Damage was held in Bratislava, Slovak Republic, from 18 to 20 May 2005. The workshop was co-organised by the OECD Nuclear Energy Agency and the Nuclear Regulatory Authority of the Slovak Republic. It attracted wide participation from national nuclear authorities, regulators, operators of nuclear installations, nuclear insurers and international organisations. The purpose of the workshop was to assess the third party liability and compensation mechanisms that would be implemented by participating countries in the event of a nuclear accident taking place within or near their borders. To accommodate this objective, two fictitious accident scenarios were developed: one involving a fire in a nuclear installation located in the Slovak Republic and resulting in the release of significant amounts of radioactive materials off-site, and the other a fire on board a ship transporting enriched uranium hexafluoride along the Danube River. The first scenario was designed to involve the greatest possible number of countries, with the second being restricted to countries with a geographical proximity to the Danube. These proceedings contain the papers presented at the workshop, as well as reports on the discussion sessions held. (author)

  17. Scratch Behaviors of Cr-Coated Zr-Based Fuel Claddings for Accident-Tolerant Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Il-Hyun; Kim, Hyun-Gil; Kim, Hyung-Kyu; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As the progression of Fukushima accident is worsened by the runaway reaction at a high temperature above 1200 .deg. C, it is essential to ensure the stabilities of coating layers on conventional Zr-based alloys during normal operations as well as severe accident conditions. This is because the failures of coating layer result in galvanic corrosion phenomenon by potential difference between coating layer and Zr alloy. Also, it is possible to damage the coating layer during handling and manufacturing process by contacting structural components of a fuel assembly. So, adhesion strength is one of the key factors determining the reliability of the coating layer on conventional Zr-based alloy. In this study, two kinds of Cr-coated Zr-based claddings were prepared using arc ion plating (AIP) and direct laser (DL) coating methods. The objective is to evaluate the scratch deformation behaviors of each coating layers on Zr alloys. Large area spallation below normal load of about 15 N appeared to be the predominant mode of failure in the AIP coating during scratch test. However, no tensile crack were found in entire stroke length. In DL coating, small plastic deformation and grooving behavior are more dominant scratching results. It was observed that the change of the slope of the COF curve did not coincide with the failure of coating layer.

  18. Full scale simulations of accidents on spent-nuclear-fuel shipping systems

    International Nuclear Information System (INIS)

    Yoshimura, H.R.

    1978-01-01

    In 1977 and 1978, five first-of-a-kind full scale tests of spent-nuclear-fuel shipping systems were conducted at Sandia Laboratories. The objectives of this broad test program were (1) to assess and demonstrate the validity of current analytical and scale modeling techniques for predicting damage in accident conditions by comparing predicted results with actual test results, and (2) to gain quantitative knowledge of extreme accident environments by assessing the response of full scale hardware under actual test conditions. The tests were not intended to validate the present regulatory standards. The spent fuel cask tests fell into the following configurations: crashes of a truck-transport system into a massive concrete barrier (100 and 130 km/h); a grade crossing impact test (130 km/h) involving a locomotive and a stalled tractor-trailer; and a railcar shipping system impact into a massive concrete barrier (130 km/h) followed by fire. In addition to collecting much data on the response of cask transport systems, the program has demonstrated thus far that current analytical and scale modeling techniques are valid approaches for predicting vehicular and cask damage in accident environments. The tests have also shown that the spent casks tested are extremely rugged devices capable of retaining their radioactive contents in very severe accidents

  19. Prevention of criticality accidents. Fuel elements storage

    International Nuclear Information System (INIS)

    Canavese, S.I.; Capadona, N.M.

    1990-01-01

    Before the need to store fuel elements of the plate type MTR (Materials Testing Reactors), produced with enriched uranium at 20% in U235 for research reactors, it requires the design of a deposit for this purpose, which will give intrinsic security at a great extent and no complaints regarding its construction, is required. (Author) [es

  20. Accidents with damage to nuclear core. A perspective for TMI-2

    International Nuclear Information System (INIS)

    Alonso, A.

    1980-01-01

    The most direct consequence of the TMI-2 accident was the destruction of substantial fraction of the fuel element cladding. With the aim of given a certain perspective to that accident, an analysis is made of the causes by which the fuel element clad may lose its integrity. The Windscale, SL-1 and Enrico Fermi accidents constitute important examples to that end. These accidents are analyzed giving special emphasis to those aspects which apear later on at TMI-2. The general consequences of the latter are examined with a certain details, including the social, institutional, technological and economic aspects of the accident. (author)

  1. Prevention of criticality accidents in a fuel cycle plant

    International Nuclear Information System (INIS)

    Gatti, A.M.; Canavese, S.I.; Capadona, N.M.

    1990-01-01

    This work reports the basic considerations on criticality accidents applied to an uranium dioxide fuel cycle production plant. The different fabrication stages are briefly described, with the identification of the neutronically isolated areas. Once the areas have been defined, an evaluation is made, setting up the control parameters to be used in each of them and their variation ranges; normal operation limitations based on experimental data or validating calculations, applied specifically to 5% enriched uranium, are established. Afterwards, defined parameters deviations are analyzed due to incidental conditions in order to prevent criticality accidents under normal conditions and maintenance operations. (Author) [es

  2. Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

    International Nuclear Information System (INIS)

    Schwegler, E.C.; Maudlin, P.J.

    1983-01-01

    A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

  3. Fuel and control rod failure behavior during degraded core accidents

    International Nuclear Information System (INIS)

    Chung, K.S.

    1984-01-01

    As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

  4. Material Selection for Accident Tolerant Fuel Cladding

    International Nuclear Information System (INIS)

    Pint, Bruce A.; Terrani, Kurt A.; Yamamoto, Yukinori; Snead, Lance Lewis

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H 2 environments at ≥1473 K (1200°C) for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti 2 AlC form a protective alumina scale in steam. However, commercial Ti 2 AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO 2 , and therefore Ti 2 AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α' formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  5. Material Selection for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    Alternative cladding materials to Zr-based alloys are being investigated for accident tolerance, which can be defined as > 100X improvement (compared to Zr-based alloys) in oxidation resistance to steam or steam-H2 environments at ≥ 1200°C for short times. After reviewing a wide range of candidates, current steam oxidation testing is being conducted on Mo, MAX phases and FeCrAl alloys. Recently reported low mass losses for Mo in steam at 800°C could not be reproduced. Both FeCrAl and MAX phase Ti2AlC form a protective alumina scale in steam. However, commercial Ti2AlC that was not single phase, formed a much thicker oxide at 1200°C in steam and significant TiO2, and therefore Ti2AlC may be challenging to form as a cladding or a coating. Alloy development for FeCrAl is seeking to maintain its steam oxidation resistance to 1475°C, while reducing its Cr content to minimize susceptibility to irradiation-assisted α´ formation. The composition effects and critical limits to retaining protective scale formation at > 1400°C are still being evaluated.

  6. Fuel behavior and fission product release under HTGR accident conditions

    International Nuclear Information System (INIS)

    Fukuda, K.; Hayashi, K.; Shiba, K.

    1990-01-01

    In early 1989 a final decision was made over construction of a 30 MWth HTGR called the High Temperature Engineering Test Reactor, HTTR, in Japan in order to utilize it for high temperature gas engineering tests and various nuclear material tests. The HTTR fuel is a pin-in-block type fuel element which is composed of a hexagonal graphite block with dimension of 580 mm in length and 360 mm in face-to-face distance and about 30 of the fuel rods inserted into the coolant channels drilled in the block. The TRISO coated fuel particles for HTTR are incorporated with graphite powder and phenol resin into the fuel compacts, 19 of which are encased into a graphite sleeve as a fuel rod. It is necessary for the HTTR licensing to prove the fuel stability under predicted accidents related to the high temperature events. Therefore, the release of the fission products and the fuel failure have been investigated in the irradiation---and the heating experiments simulating these conditions at JAERI. This report describes the HTTR fuel behavior at extreme temperature, made clear in these experiments

  7. Behavior of LWR fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

    1977-01-01

    In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

  8. Westinghouse accident tolerant fuel program. Current results and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Ray, Sumit; Xu, Peng; Lahoda, Edward; Hallstadius, Lars; Boylan, Frank [Westinghouse Electric Company LLC, Hopkins, SC (United States)

    2016-07-15

    This paper discusses the current status, results from initial tests, as well as the future direction of the Westinghouse's Accident Tolerant Fuel (ATF) program. The current preliminary testing is addressed that is being performed on these samples at the Massachusetts Institute of Technology (MIT) test reactor, initial results from these tests, as well as the technical learning from these test results. In the Westinghouse ATF approach, higher density pellets play a significant role in the development of an integrated fuel system.

  9. Status report on the EPRI fuel cycle accident risk assessment

    International Nuclear Information System (INIS)

    Erdmann, R.C.; Fullwood, R.R.; Garcia, A.A.; Mendoza, Z.T.; Ritzman, R.L.; Stevens, C.A.

    1979-07-01

    This report summarizes and extends the work reported in five unpublished draft reports: the accidental radiological risk of reprocessing spent fuel, mixed oxide fuel fabrication, the transportation of materials within the fuel cycle, and the disposal of nuclear wastes, and the routine atmospheric radiological risk of mining and milling uranium-bearing ore. Results show that the total risk contribution of the fuel cycle is only about 1% of the accident risk of the power plant and hence, with little error, the accident risk of nuclear electric power is that of the power plant itself. The power plant risk, assuming a very large usage of nuclear power by the year 2005, is only about 0.5% of the radiological risk of natural background. This work aims at a realistic assessment of the process hazards, the effectiveness of confinement and mitigation systems and procedures, and the associated likelihoods and estimated errors. The primary probabilistic estimation tool is fault tree analysis with the release source terms calculated using physical--chemical processes. Doses and health effects are calculated with the CRAC code. No evacuation or mitigation is considered: source terms may be conservative through the assumption of high fuel burnup (40,000 MWd/T) and short cooling (90 to 150 d); HEPA filter efficiencies are derived from experiments

  10. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  12. Use of activity measurements in the plume from Chernobyl to deduce fuel state before, during and after the accident

    International Nuclear Information System (INIS)

    Longworth, J.P.; Tobias, A.

    1986-07-01

    Work performed at Berkely Nuclear Laboratories both prior to the meeting in Vienna at which USSR gave full details of the Chernobyl accident and after that meeting is recorded. Plume data from Western Europe were used to deduce the likely damage to the fuel and its previous irradiation history. The note concludes that the source to the environment consisted of an initial dispersion of fuel particulate followed by a prolonged release at a lower rate, the total release being some 3% of the core inventory of fuel. Early and late in the release period it was enhanced in volatile species. Damage to the fuel was thus due both to mechanical disruption and to high temperatures. During the early dispersive event high temperatures (probably approaching fuel melting) were reached in some of the core, though the proportion of the fuel affected may have been small. (UK)

  13. 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.

  14. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    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

  15. Use of classical criterions of a decision making for choice of measures on decrease of economic damage from nuclear and radiation accidents

    International Nuclear Information System (INIS)

    Rylov, M.I.; Kamynov, Sh.V.; Mozhaev, A.S.; Anisimov, N.A.; Nikitin, V.S.

    2004-01-01

    Application of classical criteria of decision making for choice of measures on the decrease of economic damage from possible nuclear and radiation accidents during spent fuel unloading from nuclear submarines and storage in the process of their utilization was demonstrated. Economic damage was chosen as optimization index, three versions of possible accidents and limited number of measures on the decrease of their effect were treated for illustration of the suggested approach. On the base of analysis of classical criteria the optimal strategy for decrease of economic damage was chosen [ru

  16. Development of Collision Accident Scenario during Nuclear Spent Fuel Maritime Transportation

    International Nuclear Information System (INIS)

    Yoo, Min; Kang, Hyun Gook

    2015-01-01

    Population density of South Korea is much higher than the other countries, and it is peninsula. Therefore, it is expected that major means of transportation of the spent fuel will be maritime transportation rather than overland transportation. Korea Maritime safety Tribunal (KMST) categorized various maritime accident, see table I. Among them, collision accident is one of the most important and complicated accident from Probabilistic Safety Analysis (PSA) point of view. We will show what will happen if the transportation ship is struck by other ship, how to calculate collision energy and probability of the branches on ship-ship collision with Event Tree Analysis (ETA) method. We selected and re-categorized maritime accident that KMST categorized for ship-ship collision analysis of spent fuel transportation ship. Event tree is constructed and collision energy distribution is derived from statistics and equation. And outer and inner hull fracture probabilities are calculated. If outer hull is broken but inner hull is fine, water will be flooded into the space between outer and inner hull. It will decrease mobility of the ship. If inner hull is fractured, water will be flooded into the ship inside. The ship has compartment structure to resist from foundering. Loss of mobility and compartment damage (ultimately it ends with sink) mechanism need to be analyzed to complete transportation ship collision event tree

  17. Collective radiation doses following a hypothetical, very severe accident to an irradiated fuel transport flask containing AGR fuel

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1985-05-01

    Studies of the consequences of very severe, although unlikely, accidents to irradiated fuel transport flasks are made in order to evaluate risks. If an irradiated fuel transport flask carrying AGR fuel were damaged in a hypothetical accident involving a severe impact followed by a prolonged fire, a small proportion of caesium and other fission products might be released to the atmosphere from the gap inventory of broken fuel pins. The consequent radiation dose to the public would arise predominantly by direct irradiation from ground deposits and the ingestion of slightly contaminated foodstuffs. Although these collective doses must generally be estimated with the aid of computer codes, it is shown here that the worst case, when a high proportion of the radioactivity is deposited in a densely population area, can be assessed approximately by a much simpler method, an approach which is of great value in explaining the calculation in a manner that can be readily understood. A comparison is made between the simple approach and equivalent results from the NECTAR code, the worst case is compared with an ensemble average over all weather conditions, and the relative contributions of the two main routes to collective dose are discussed. (author)

  18. Consequence analysis of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    Wahba, N.N.; Kim, Y.T.; Lie, S.G.

    1997-01-01

    The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

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

    International Nuclear Information System (INIS)

    Grundfelt, Bertil

    2013-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  1. Harvesting budworm-damaged stands for fuel

    Energy Technology Data Exchange (ETDEWEB)

    Henley, S.G. (York, Sunbury, Charlotte Wood Products Marketing Board, (Canada))

    1985-01-01

    This project was initiated to demonstrate the economics and logistics of harvesting budworm-damaged stands for use as fuel. Dead spruce and balsam fir were to be harvested from small private woodlots in southwestern New Brunswick, using an integrated, full-tree harvesting system to produce wood chip fuel and other forest products. The overall objectives of the study are listed. The harvesting equipment and the selection of sites are discussed. The most efficient methods of finding candidate woodlots was found to be by advertising and word of mouth. Contact was made with 85 woodlot owners, and 45 woodlots were visited and evaluated for their suitability. A further 150 management plans were screened and rejected for various reasons. Only 2 woodlots were initially recognized as potential sites; however, after showing some interest, the owners decided not to participate. The reasons for the rejection of the various woodlots are listed. The fact that a number of owners were against clearcutting, and, in some cases, against any cutting, and that others showed no interest in the study, is attributed to the high percentage of white-collar workers owning woodlots. Other strategies for harvesting dead or scrap wood are suggested. 1 ref., 1 tab.

  2. Experimental Setup for Reflood Quench of Accident Tolerant Fuel Claddings

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan; Lee, Kwan Geun; In, Wang Kee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The concept of accident tolerant fuel (ATF) is a solution to suppress the hydrogen generation in loss of coolant accident (LOCA) situation without safety injection, which was the critical incident in the severe accident in the Fukushima. The changes in fuel and cladding materials may cause a significant difference in reactor performance in long term operation. Properties in terms of material science and engineering have been tested and showed promising results. However, numerous tests are still required to ensure the design performance and safety. Thermal hydraulic tests including boiling and quenching are partly confirmed, but not yet complete. We have been establishing the experimental setup to confirm the properties in the terms of thermal hydraulics. Design considerations and preliminary tests are introduced in this paper. An experimental setup to test thermal hydraulic characteristics of new ATF claddings are established and tested. The W heater set inside the cladding is working properly, exceeding 690 W/m linear power with thermocouples and insulating ceramic sheaths inside. The coolant injection control was also working in good conditions. The setup is about to complete and going to simulate quenching behavior of the ATF in the LOCA situation.

  3. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  4. Accident Damage Analysis Module (ADAM) – Technical Guidance, Software tool for Consequence Analysis calculations

    OpenAIRE

    FABBRI LUCIANO; BINDA MASSIMO; BRUINEN DE BRUIN YURI

    2017-01-01

    This report provides a technical description of the modelling and assumptions of the Accident Damage Analysis Module (ADAM) software application, which has been recently developed by the Joint Research Centre (JRC) of the European Commission (EC) to assess physical effects of an industrial accident resulting from an unintended release of a dangerous substance

  5. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

  6. The behaviour of spherical HTR fuel elements under accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schenk, W; Naoumidis, A [Institute for Reactor Material, KFA Juelich (Germany)

    1985-07-01

    Hypothetical accidents may lead to significantly higher temperatures in HTR fuel than during normal operation. In order to obtain meaningful statements on fission product behaviour and release, irradiated spherical fuel elements containing a large number of coated particles (20,000-40,000) with burnups between 6 and 16% FIMA were heated at temperatures between 1400 and 2500 deg. C. HTI-pyrocarbon coating retains the gaseous fission products (e.g. Kr) very well up to about 2400 deg. C if the burnup does not exceed the specified value for THTR (11.5%). Cs diffuses through the pyrocarbon significantly faster than Kr and the diffusion is enhanced at higher fuel burnups because of irradiation induced kernel microstructure changes. Below about 1800 deg. C the Cs release rate is controlled by diffusion in the fuel kernel; above this temperature the diffusion in the pyrocarbon coating is the controlling parameter. An additional SiC coating interlayer (TRISO) ensures Cs retention up to 1600 deg. C. However, the release obtained in the examined fuel elements was only by a factor of three lower than through the HTI pyrocarbon. Solid fission products added to UO{sub 2}-TRISO particles to simulate high burnup behave in various ways and migrate to attack the SiC coating. Pd migrates fastest and changes the SiC microstructure making it permeable.

  7. Damage of fuel assembly premature changing in a power reactor

    International Nuclear Information System (INIS)

    Rudik, A.P.

    1987-01-01

    Material balance, including energy recovery and nuclear fuel flow rate, under conditions of premature FA extraction from power reactor is considered. It is shown that in cases when before and after FA extraction reactor operates not under optimal conditions damage of FA premature changing is proportional to the first degree of fuel incomplete burning. If normal operating conditions of reactor or its operation after FA changing is optimal, the damage is proportional to the square of fuel incomplete burning

  8. Dropped fuel damage prediction techniques and the DROPFU code

    International Nuclear Information System (INIS)

    Mottershead, K.J.; Beardsmore, D.W.; Money, G.

    1995-01-01

    During refuelling, and fuel handling, at UK Advanced Gas Cooled Reactor (AGR) stations it is recognised that the accidental dropping of fuel is a possibility. This can result in dropping individual fuel elements, a complete fuel stringer, or a whole assembly. The techniques for assessing potential damage have been developed over a number of years. This paper describes how damage prediction techniques have subsequently evolved to meet changing needs. These have been due to later fuel designs and the need to consider drops in facilities outside the reactor. The paper begins by briefly describing AGR fuel and possible dropped fuel scenarios. This is followed by a brief summary of the damage mechanisms and the assessment procedure as it was first developed. The paper then describes the additional test work carried out, followed by the detailed numerical modelling. Finally, the paper describes the extensions to the practical assessment methods. (author)

  9. Noble gas confinement for reactor fuel melting accidents

    International Nuclear Information System (INIS)

    Monson, P.R.

    1984-01-01

    In the unlikely event of a fuel melting accident, radioactive material would be released into the reactor room. This radioactive material would consist of particulate matter, iodine, tritium, and the noble gases krypton and xenon. In the case of reactors with containment domes the gases would be contained for subsequent cleanup. For reactors without contaiment the particulates and the iodine can be effectively removed with HEPA and carbon filters of current technology; however, noble gases cannot be easily removed and would be released to the atmosphere. In either case, it would be highly desirable to have a system that could be brought online to treat this contaminated air to minimize the population dose. A low temperature adsorption system has been developed at the Savannah River Laboratory to remove the airborne radioactive material from such a fuel melting accident. Over two dozen materials have been tested in extensive laboratory studies, and hydrogen mordenite and silver mordenite were found to be the most promising adsorbents. A full-scale conceptual design has also been developed. Results of the laboratory studies and the conceptual design are discussed along with plans for further development of this concept

  10. Noble gas confinement for reactor fuel melting accidents

    International Nuclear Information System (INIS)

    Monson, P.R.

    1985-01-01

    In the unlikely event of a fuel melting accident radioactive material would be released into the reactor room. This radioactive material would consist of particulate matter, iodine, tritium, and the noble gases krypton and xenon. In the case of reactors with containment domes, the gases would be contained for subsequent cleanup. For reactors without containment the particulates and the iodine can be effectively removed with HEPA and carbon filters of current technology; however, noble gases cannot be easily removed and would be released to the atmosphere. In either case, it would be highly desirable to have a system that could be brought online to treat this contaminated air to minimize the population dose. A low temperature adsorption system has been developed at the Savannah River Laboratory to remove the airborne radioactive material from such a fuel melting accident. Over two dozen materials have been tested in extensive laboratory studies, and hydrogen mordensite and silver mordenite were found to be the most promising absorbents. A full-scale conceptual design has also been developed. Results of the laboratory studies and the conceptual design will be discussed along with plans for further development of this concept

  11. Development Status of Accident Tolerant Fuels for Light Water Reactors in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jae Ho; Kim, Hyun Gil; In, Wang Kee; Kim, Weon Ju; Koo, Yang Hyum [KAERI, Daejeon (Korea, Republic of); Lee, Seung Jae [KEPCONF, Daejeon (Korea, Republic of)

    2016-05-15

    Research on accident tolerant fuels (ATFs) is aimed at developing innovative fuels, which can mitigate or prevent the consequences of accidents. In Korea, innovative concepts are being developed to improve fuel safety and reliability of LWRs during accident events and normal operations. ATF technologies will be developed and commercialized through a sequence of long-lead and extensive activities. The interim milestone for new fuel program is that we would be ready for an irradiation test in commercial reactor by 2021. This presentation deals with the status of ATF development in KOREA and plan to implement new fuel technology successfully in commercial nuclear power plants.

  12. Preliminary neutronic assessment for ATF (Accident Tolerant Fuel) based on iron alloy

    International Nuclear Information System (INIS)

    Abe, Alfredo; Carluccio, Thiago; Piovezan, Pamela; Giovedi, Claudia; Martins, Marcelo R.

    2015-01-01

    After Fukushima Daiichi nuclear accident in 2011, the nuclear fuel performance under accident condition became a very important issue and currently different research and development program are in progress toward to reliability and withstand under accident condition. These initiatives are known as ATF (Accident Tolerant Fuel) R and D program, which many countries with different research institutes, fuel vendors and others are nowadays involved. Accident Tolerant Fuel (ATF) can be defined as enhanced fuel which can tolerate loss of active cooling system capability for a considerably longer time period and the fuel/cladding system can be maintained without significant degradation and can also improve the fuel performance during normal operations and transients, as well as design-basis accident (DBA) and beyond design-basis (BDBA) accident. Different materials have being proposed as fuel cladding candidates considering thermo-mechanical properties and lower reaction kinetic with steam and slower hydrogen production. The aim of this work is to perform a neutronic assessment for several cladding candidates based on iron alloy considering a standard PWR fuel rod (fuel pellet and dimension). The purpose of the assessment is to address different parameters that might contribute for possible neutronic reactivity gain in order to overcome the penalty due to increase of neutron absorption in the cladding materials. All the neutronic assessment is performed using MCNP, Monte Carlo code. (author)

  13. Bruce B fuelling-with-flow operations: fuel damage investigation

    Energy Technology Data Exchange (ETDEWEB)

    Manzer, A.M. [CANTECH Associates Ltd., Burlington, Ontario (Canada); Morikawa, D. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Hains, A.J.; Cichowlas, W.M. [Nuclear Safety Solutions Limited, Toronto, Ontario (Canada); Roberts, J.G.; Wylie, J. [Bruce Power, Ontario (Canada)

    2005-07-01

    This paper summarizes the fuel bundle damage characterization done by Nuclear Safety Solutions Limited (NSS) and the out-reactor flow visualization tests done at Atomic Energy of Canada Limited (AECL) to reproduce the damage observed on irradiated fuel bundles. The bearing pad damage mechanism was identified and the tests showed that a minor change to the fuelling sequence would eliminate the mechanical interaction. The change was implemented in January 2005. Since then, the bearing pad damage appears to have been greatly reduced based on the small number of discharged bundles inspected to date. (author)

  14. Bruce B fuelling-with-flow operations: fuel damage investigation

    International Nuclear Information System (INIS)

    Manzer, A.M.; Morikawa, D.; Hains, A.J.; Cichowlas, W.M.; Roberts, J.G.; Wylie, J.

    2005-01-01

    This paper summarizes the fuel bundle damage characterization done by Nuclear Safety Solutions Limited (NSS) and the out-reactor flow visualization tests done at Atomic Energy of Canada Limited (AECL) to reproduce the damage observed on irradiated fuel bundles. The bearing pad damage mechanism was identified and the tests showed that a minor change to the fuelling sequence would eliminate the mechanical interaction. The change was implemented in January 2005. Since then, the bearing pad damage appears to have been greatly reduced based on the small number of discharged bundles inspected to date. (author)

  15. Simulation of the mechanical behavior of a spent fuel shipping cask in a rail accident environment

    International Nuclear Information System (INIS)

    Fields, S.R.

    1977-02-01

    A preliminary mathematical model has been developed to simulate the dynamic mechanical response of a large spent fuel shipping cask to the impact experienced in a hypothetical rail accident. The report was written to record the status of the development of the mechanical response model and to supplement an earlier report on spent fuel shipping cask accident evaluation

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

    International Nuclear Information System (INIS)

    Ostmeyer, R.M.

    1986-01-01

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

  17. Economic damage caused by a nuclear reactor accident

    International Nuclear Information System (INIS)

    Baan, P.J.A.

    1988-01-01

    The impacts of a nuclear reactor accident have been estimated for: the public water supply; the use of surface water for sprinkling in agriculture, for industry water supply, recreation, etc.; and fisheries. Contamination of water sources may affect the public water supply severely. In such a situation demand of water cannot always be met. Agriculture faces production losses, if demand for uncontaminated surface water cannot be met. The impacts on recreation can also be significant. The losses to other water users are less substantial. Fisheries may lose (export) markets, as people become reluctant to buy fish and fish products. 33 refs.; 3 figs.; 35 tabs

  18. Clinical peculiarities of the brain damage in the liquidators of the Chernobyl accident

    Energy Technology Data Exchange (ETDEWEB)

    Zozulya, Y A; Vinnitsky, A R; Stepanenko, I V [Institute of Neurosurgery, Academy of Medical Sciences, Kiev (Ukraine)

    1997-09-01

    Investigation into the features of the brain damage by the liquidators of the Chernobyl accident has become an urgent issue of today due to a number of circumstances. According to the classical concept dominating radiobiology until recently, the brain being composed of highly - differentiated nerve cells, present a radioresistant structure responsive to radiation injury induced by high and very high radiation doses (10000 rem and higher) only. The results of clinical examinations given to the Chernobyl accident recovery workers at Kiev Institute of Neurosurgery, Academy of Medical Sciences of Ukraine, show that even the so - called ``small - dose`` radiation, when consumed continuously, produces neurological sings of brain damage. 6 figs.

  19. Clinical peculiarities of the brain damage in the liquidators of the Chernobyl accident

    International Nuclear Information System (INIS)

    Zozulya, Y.A.; Vinnitsky, A.R.; Stepanenko, I.V.

    1997-01-01

    Investigation into the features of the brain damage by the liquidators of the Chernobyl accident has become an urgent issue of today due to a number of circumstances. According to the classical concept dominating radiobiology until recently, the brain being composed of highly - differentiated nerve cells, present a radioresistant structure responsive to radiation injury induced by high and very high radiation doses (10000 rem and higher) only. The results of clinical examinations given to the Chernobyl accident recovery workers at Kiev Institute of Neurosurgery, Academy of Medical Sciences of Ukraine, show that even the so - called ''small - dose'' radiation, when consumed continuously, produces neurological sings of brain damage. 6 figs

  20. Oxidation behavior of fuel cladding tube in spent fuel pool accident condition

    International Nuclear Information System (INIS)

    Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo; Tojo, Masayuki

    2017-01-01

    In spent fuel pool (SFP) under loss-of-cooling or loss-of-coolant severe accident condition, the spent fuels will be exposed to air and heated by their own residual decay heat. Integrity of fuel cladding is crucial for SFP safety therefore study on cladding oxidation in air at high temperature is important. Zircaloy-2 (Zry2) and zircaloy-4 (Zry4) were applied for thermogravimetric analyses (TGA) in different temperatures in air at different flow rates to evaluate oxidation behavior. Oxidation rate increased with testing temperature. In a range of flow rate of air which is predictable in spent fuel lack during a hypothetical SFP accident, influence of flow rate was not clearly observed below 950degC for the Zry2, or below 1050degC for Zry4. In higher temperature, oxidation rate was higher in high rate condition, and this trend was seen clearer when temperature increased. Oxide layers were carefully examined after the TGA analyses and compared with mass gain data to investigate detail of oxidation process in air. It was revealed that the mass gain data in pre-breakaway regime reflects growth of dense oxide film on specimen surface, meanwhile in post-breakaway regime, it reflects growth of porous oxide layer beneath fracture of the dense oxide film. (author)

  1. Removal of the damaged fuel from Paks-2 pit

    International Nuclear Information System (INIS)

    Cserhati, A.

    2007-01-01

    On 10 April 2003, during the outage period a chemical cleaning program for the fuel assemblies has been carried out at the unit 2, in a specially designed cleaning tank. The tank is located in a pit, near to the reactor. 30 fuel assemblies have been significantly damaged due to inadequate cooling. After the extensive preparation - lasting 3,5 years - the pickup and encapsulation of the damaged fuel has been preformed. All tasks have been carried out safely, during the planned 3 months without any substantial problems. This paper covers the events of this last implementation phase. The main topics are: initial conditions of the pit and the cleaning tank before the start of the recovery; tasks and responsibilities, organization, timing, control.; visual following for the fuel removal; technology features, steps made; short and long term tasks after the removal of the fuel; summary, achievements. (author)

  2. Risks and benefits of the interventions aimed at minimizing nuclear damage in the Chernobyl accident

    International Nuclear Information System (INIS)

    Rossiello, L.A.; Failla, L.

    1997-01-01

    The damages that the absorption of ionizing radiation (i.r.) can cause to humans may be classified as 1) nonstochastic (somatic or deterministic) or 2) stochastic (probabilistic) , which result, for example, from high doses of i.r. absorbed after a serious nuclear accident. Though the Chernobyl case involved both kinds of damage, this paper deals only with stochastic damage risk, and confine our considerations to individuals who were directly Affected and received high i.r. doses. The purpose of this paper is to provide elements on which to base future decisions on the evacuation and return of populations affected by serious nuclear accidents. Unlike the abundant literature on the subject, and as a necessary complement thereto within the bounds of a strict synthesis, to identify the most significant parameters applicable to single individuals rather than to the population at large, and referring solely to risks of stochastic damage

  3. Economic estimation of risk and compensation of damage from accidents in power engineering objects

    International Nuclear Information System (INIS)

    Lesnykh, V.V.

    1996-01-01

    Place and basic peculiarities of the task relative to compensation of damage due to accidents in the problem on technical-economical studies of the power engineering objects, including NPPs, are analyzed. Certain approaches in the task of the risk economical estimates and basic provisions of the economical damage compensation system are presented. Description of imitated and analytical approach in the task of estimating financial state is given and certain study results are presented. 11 refs., 8 figs

  4. Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Folsom, Charles Pearson [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, Giovanni [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-05-01

    One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

  5. Post-test investigation result on the WWER-1000 fuel tested under severe accident conditions

    International Nuclear Information System (INIS)

    Goryachev, A.; Shtuckert, Yu.; Zwir, E.; Stupina, L.

    1996-01-01

    The model bundle of WWER-type were tested under SFD condition in the out-of-pile CORA installation. The objective of the test was to provide an information on the WWER-type fuel bundles behaviour under severe fuel damage accident conditions. Also it was assumed to compare the WWER-type bundle damage mechanisms with these experienced in the PWR-type bundle tests with aim to confirm a possibility to use the various code systems, worked our for PWR as applied to WWER. In order to ensure the possibility of the comparison of the calculated core degradation parameters with the real state of the tested bundle, some parameters have been measured on the bundle cross-sections under examination. Quantitative parameters of the bundle degradation have been evaluated by digital image processing of the bundle cross-sections. The obtained results are shown together with corresponding results obtained by the other participants of this investigation. (author). 3 refs, 13 figs

  6. Aging, Loss-of-Coolant Accident (LOCA), and high potential testing of damaged cables

    International Nuclear Information System (INIS)

    Vigil, R.A.; Jacobus, M.J.

    1994-04-01

    Experiments were conducted to assess the effects of high potential testing of cables and to assess the survivability of aged and damaged cables under Loss-of-Coolant Accident (LOCA) conditions. High potential testing at 240 Vdc/mil on undamaged cables suggested that no damage was incurred on the selected virgin cables. During aging and LOCA testing, Okonite ethylene propylene rubber (EPR) cables with a bonded jacket experienced unexpected failures. The failures appear to be primarily related to the level of thermal aging and the presence of a bonded jacket that ages more rapidly than the insulation. For Brand Rex crosslinked polyolefin (XLPO) cables, the results suggest that 7 mils of insulation remaining should give the cables a high probability of surviving accident exposure following aging. The voltage necessary to detect when 7 mils of insulation remain on unaged Brand Rex cables is approximately 35 kVdc. This voltage level would almost certainly be unacceptable to a utility for use as a damage assessment tool. However, additional tests indicated that a 35 kvdc voltage application would not damage virgin Brand Rex cables when tested in water. Although two damaged Rockbestos silicone rubber cables also failed during the accident test, no correlation between failures and level of damage was apparent

  7. Application of Coating Technology for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature (RT), and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.

  8. Assessment of clad integrity of PHWR fuel pin following a postulated severe accident

    International Nuclear Information System (INIS)

    Dutta, B.K.; Kushwaha, H.S.; Venkat Raj, V.

    2000-01-01

    A mechanistic fuel performance analysis code FAIR has been developed. The code can analyse fuel pins with free standing as well as collapsible clad under normal, off-normal and accident conditions of reactors. The code FAIR is capable of analysing the effects of high burnup on fuel behaviour. The code incorporates finite element based thermo-mechanical module for computing transient temperature distribution and thermal-elastic-plastic stresses in the fuel pin. A number of high temperature thermo-physical and thermo-mechanical models also have been incorporated for analysing fuel pins subjected to severe accident scenario. The present paper describes salient features of code FAIR and assessment of clad integrity of PHWR fuel pins with different initial burnup subjected to severe accident scenario. (author)

  9. Experience with failed or damaged spent fuel and its impacts on handling

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1989-12-01

    Spent fuel management planning needs to include consideration of failed or damaged spent light-water reactor (LWR) fuel. Described in this paper, which was prepared under the Commercial Spent Fuel Management (CSFM) Program that is sponsored by the US Department of Energy (DOE), are the following: the importance of fuel integrity and the behavior of failed fuel, the quantity and burnup of failed or damaged fuel in storage, types of defects, difficulties in evaluating data on failed or damaged fuel, experience with wet storage, experience with dry storage, handling of failed or damaged fuel, transporting of fuel, experience with higher burnup fuel, and conclusions. 15 refs

  10. Compensation for damages in case of a nuclear accident; L'indemnisation des prejudices en cas d'accident nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Leger, M. [CEA Saclay, 91 - Gif sur Yvette (France)

    2011-01-15

    This article presents the system of compensation for damages in case of a nuclear accident. This system of civil liability for nuclear damage, as a specific regime, departs on several points from the common rules of civil liability, in order to provide an adequate and equitable compensation for the damages suffered by the victims of nuclear accidents. The French system of civil liability for nuclear damage results from two International Conventions integrated in French law (Paris convention 1960 and Brussels convention 1963) and the French law of 1968, October 30 on civil liability in the area of nuclear energy. These texts define the conditions under which a nuclear operator could be held liable in case of a nuclear accident. The protocols to amend the Paris and Brussels Conventions of 2004, not yet come into force, are also presented. They ensure that increased resources are available to compensate a greater number of victims of a nuclear accident. (author)

  11. Transuranium contamination in BWRs after fuel accidents and its impact on decommissioning exposures and costs

    Energy Technology Data Exchange (ETDEWEB)

    Lundgren, K.

    1996-12-01

    The theme of the present study is to quantify the amount of transuranium activity in different parts of the plant after various fuel accidents, and which impact such contamination has on radiation exposure and costs for decommissioning the plant. The consequences of four different accident degrees have been treated: Common fuel failures, e.g. in line with recent experiences from Swedish BWRs; Fuel channel obstruction resulting in partial melting of one fuel assembly; Total loss of electric power resulting in partial meltdown of the core, but with primary circuit intact preventing a massive contamination of the containment; A LOCA followed by a core meltdown and melting and penetration of the reactor pressure vessel. The amount of transuranium activity distributed, the form of this activity and the plant contamination are evaluated for these accidents. The costs and exposures have been split up on cleanup activities after the accident and decommissioning. 75 refs.

  12. Transuranium contamination in BWRs after fuel accidents and its impact on decommissioning exposures and costs

    International Nuclear Information System (INIS)

    Lundgren, K.

    1996-12-01

    The theme of the present study is to quantify the amount of transuranium activity in different parts of the plant after various fuel accidents, and which impact such contamination has on radiation exposure and costs for decommissioning the plant. The consequences of four different accident degrees have been treated: Common fuel failures, e.g. in line with recent experiences from Swedish BWRs; Fuel channel obstruction resulting in partial melting of one fuel assembly; Total loss of electric power resulting in partial meltdown of the core, but with primary circuit intact preventing a massive contamination of the containment; A LOCA followed by a core meltdown and melting and penetration of the reactor pressure vessel. The amount of transuranium activity distributed, the form of this activity and the plant contamination are evaluated for these accidents. The costs and exposures have been split up on cleanup activities after the accident and decommissioning. 75 refs

  13. Study on light water reactor fuel behavior under reactivity initiated accident condition in TREAT

    International Nuclear Information System (INIS)

    Ohnishi, Nobuaki; Ishijima, Kiyomi; Ochiai, Masaaki; Tanzawa, Sadamitsu; Uemura, Mutsumi

    1981-05-01

    This report reviews the results of the fuel failure experiments performed in TREAT in the U.S.A. simulating Reactivity Initiated Accidents. One of the main purposes of the TREAT experiments is the study of the fuel failure behavior, and the other is the study of the molten fuel-water coolant interaction and the consequent hydrogen behavior. This report mainly shows the results of the TREAT experiments studying the fuel failure behavior in Light Water Reactor, and then it describes the fuel failure threshold and the fuel failure mechanism, considering the results of the photographic experiments of the fuel failure behavior with transparent capsules. (author)

  14. Fuel solution criticality accident studies with the SILENE reactor: phenomenology, consequences and simulated intervention

    International Nuclear Information System (INIS)

    Barbry, F.

    1984-01-01

    After defining the content and the objectives of criticality accident studies, the SILENE reactor, a means of studying fuel solution criticality accidents, is presented. Information obtained from the CRAC and SILENE experimental programs are then presented; they concern power excursion phenomenology, radiological consequences, and finally guide-lines for current and future programs

  15. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  16. Accident-generated radioactive particle source term development for consequence assessment of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Sutter, S.L.; Ballinger, M.Y.; Halverson, M.A.; Mishima, J.

    1983-04-01

    Consequences of nuclear fuel cycle facility accidents can be evaluated using aerosol release factors developed at Pacific Northwest Laboratory. These experimentally determined factors are compiled and consequence assessment methods are discussed. Release factors can be used to estimate the fraction of material initially made airborne by postulated accident scenarios. These release fractions in turn can be used in models to estimate downwind contamination levels as required for safety assessments of nuclear fuel cycle facilities. 20 references, 4 tables

  17. CORA-13 experiment on severe fuel damage

    International Nuclear Information System (INIS)

    Firnhaber, M.; Trambauer, K.; Hagen, S.; Hofmann, P.; Schanz, G.; Sepold, L.

    1993-07-01

    The major objectives of the experiment were to investigate the behavior of PWR fuel elements during early core degradation and fast cooldown due to refill. Measured quantities are boundary conditions, bundle temperatures, hydrogen generation and the final bundle configuration. Boundary conditions which could not be measured, but which are necessary for simplified test simulation (axial power profile, shroud insulation temperature, bundle refill flow) were estimated using ATHLET-CD. The capability of the codes in calculating the main degradation phenomena has been clearly illustrated and weaknesses concerning the modelling of some degradation processes have been identified. Among the degradation phenomena involved in the test, the more severe limitations concern the UO 2 -ZrO 2 dissolution by molten Zr, the solubility limits in the resulting U-Zr-O mixture and the cladding failure by the molten mixture. There is a lack concerning the Inconel spacer-grid interactions with the rods, the material interaction between control rod material and fuel rods, and in the modelling of hydrogen generation during cooldown. (orig./DG)

  18. Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

    2009-03-01

    The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

  19. Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2015-11-01

    The demands on nuclear fuel have recently been increasing, and include transient regimes, higher discharge burnup and longer fuel cycles. This has resulted in an increase of loads on fuel and core internals. In order to satisfy these demands while ensuring compliance with safety criteria, new national and international programmes have been launched and advanced modelling codes are being developed. The Fukushima Daiichi accident has particularly demonstrated the need for adequate analysis of all aspects of fuel performance to prevent a failure and also to predict fuel behaviour were an accident to occur.This publication presents the Proceedings of the Technical Meeting on Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents, which was hosted by the Nuclear Power Institute of China (NPIC) in Chengdu, China, following the recommendation made in 2013 at the IAEA Technical Working Group on Fuel Performance and Technology. This recommendation was in agreement with IAEA mid-term initiatives, linked to the post-Fukushima IAEA Nuclear Safety Action Plan, as well as the forthcoming Coordinated Research Project (CRP) on Fuel Modelling in Accident Conditions. At the technical meeting in Chengdu, major areas and physical phenomena, as well as types of code and experiment to be studied and used in the CRP, were discussed. The technical meeting provided a forum for international experts to review the state of the art of code development for modelling fuel performance of nuclear fuel for water cooled reactors with regard to steady state and transient conditions, and for design basis and early phases of severe accidents, including experimental support for code validation. A round table discussion focused on the needs and perspectives on fuel modelling in accident conditions. This meeting was the ninth in a series of IAEA meetings, which reflects Member States’ continuing interest in nuclear fuel issues. The previous meetings were held in 1980 (jointly with

  20. Medical treatment of radiation damages and medical emergency planning in case of nuclear power plant incidents and accidents

    International Nuclear Information System (INIS)

    Ohlenschlaeger, L.

    1981-03-01

    Medical measures in case of radiation damages are discussed on the basis of five potential categories of radiation incidents and accidents, respectively, viz. contaminations, incorporations, external local and general radiation over-exposures, contaminated wounds, and combinations of radiation damages and conventional injuries. Considerations are made for diagnostic and therapeutic initial measures especially in case of minor and moderate radiation accidents. The medical emergency planning is reviewed by means of definations used in the practical handling of incidents or accidents. The parameters are: extent of the incident or accident, number of persons involved, severity of radiation damage. Based on guiding symptoms the criteria for the classification into minor, moderate or severe radiation accidents are discussed. Reference is made to the Medical Radiation Protection Centers existing in the Federal Republic of Germany and the possibility of getting advices in case of radiation incidents and accidents. (orig.) [de

  1. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  2. Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Sevy, R.H.; Su, S.F.

    1985-01-01

    This paper presents the results of a study of the effectivness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). Results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs

  3. Core damage frequency estimation using accident sequence precursor data: 1990-1993

    International Nuclear Information System (INIS)

    Martz, H.F.

    1998-01-01

    The Nuclear Regulatory Commission's (NRC's) ongoing Accident Sequence Precursor (ASP) program uses probabilistic risk assessment (PRA) techniques to assess the potential for severe core damage (henceforth referred to simply as core damage) based on operating events. The types of operating events considered include accident sequence initiators, safety equipment failures, and degradation of plant conditions that could increase the probability that various postulated accident sequences occur. Such operating events potentially reduce the margin of safety available for prevention of core damage an thus can be considered as precursors to core damage. The current process for identifying, analyzing, and documenting ASP events is described in detail in Vanden Heuval et al. The significance of a Licensee Event Report (LER) event (or events) is measured by means of the conditional probability that the event leads to core damage, the so-called conditional core damage probability or, simply, CCDP. When the first ASP study results were published in 1982, it covered the period 1969--1979. In addition to identification and ranking of precursors, the original study attempted to estimate core damage frequency (CDF) based on the precursor events. The purpose of this paper is to compare the average annual CDF estimates calculated using the CCDP sum, Cooke-Goossens, Bier, and Abramson estimators for various reactor classes using the combined ASP data for the four years, 1990--1993. An important outcome of this comparison is an answer to the persistent question regarding the degree and effect of the positive bias of the CCDP sum method in practice. Note that this paper only compares the estimators with each other. Because the true average CDF is unknown, the estimation error is also unknown. Therefore, any observations or characterizations of bias are based on purely theoretical considerations

  4. K Basins floor sludge retrieval system knockout pot basket fuel burn accident

    International Nuclear Information System (INIS)

    HUNT, J.W.

    1998-01-01

    The K Basins Sludge Retrieval System Preliminary Hazard Analysis Report (HNF-2676) identified and categorized a series of potential accidents associated with K Basins Sludge Retrieval System design and operation. The fuel burn accident was of concern with respect to the potential release of contamination resulting from a runaway chemical reaction of the uranium fuel in a knockout pot basket suspended in the air. The unmitigated radiological dose to an offsite receptor from this fuel burn accident is calculated to be much less than the offsite risk evaluation guidelines for anticipated events. However, because of potential radiation exposure to the facility worker, this accident is precluded with a safety significant lifting device that will prevent the monorail hoist from lifting the knockout pot basket out of the K Basin water pool

  5. Validation of the metal fuel version of the SAS4A accident analysis code

    International Nuclear Information System (INIS)

    Tentner, A.M.

    1991-01-01

    This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

  6. Modelling of Zircaloy-steam-oxidation under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Malang, S.; Neitzel, H.J.

    1983-01-01

    Small break loss-of-coolant accidents and special transients in an LWR, in combination with loss of required safety systems, may lead to an uncovered core for an extended period of time. As a consequence, the cladding temperature could rise up to the melting point due to the decay heat, resulting in severely damaged fuel rods. During heat-up the claddings oxidize due to oxygen uptake from the steam atmosphere in the core. The modeling and assessment of the Zircaloy-steam oxidation under such conditions is important, mainly for two reasons: The oxidation of the cladding influences the temperature transients due to the exothermic heat of reaction; the amount of liquified fuel depends on the oxide layer thickness and the oxygen content of the remaining Zircaloy metal when the melting point is reached. (author)

  7. Visual observations of fuel disruption in in-pile LMFBR accident experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Mast, P.K.

    1982-01-01

    Sandia National Laboratories has been investigating initiation phase phenomena in a series of Fuel Disruption (FD) experiments since 1977. In this program high speed cinematography is used to observe fuel disruption in in-pile experiments that simulate loss of flow accidents. Thus, these experiments provide high resolution measurements of initial fuel and clad motion with prototypic materials and prototypic heating conditions. The main objective of the FD experiment is to determine the timing (relative to fuel temperature) and the mode of fuel disruption under LOF heating conditions. Observed modes of disruption include fuel swelling, solid state breakup, cracking, ejection of a molten fuel jet, slumping, and rapid expansion of small particles. Because the temperature and character of the fuel at disruption are known, disruption can be correlated with the mechanisms driving the disruption such as fuel vapor pressure, molten fuel expansion, fission gases, and impurity gases

  8. TREAT experimental data base regarding fuel dispersals in LMFBR loss-of-flow accidents

    International Nuclear Information System (INIS)

    Simms, R.; Fink, C.L.; Stanford, G.S.; Regis, J.P.

    1981-01-01

    The reactivity feedback from fuel relocation is a central issue in the analysis of loss-of-flow (LOF) accidents in LMFBRs. Fuel relocation has been studied in a number of LOF simulations in the TREAT reactor. In this paper the results of these tests are analyzed, using, as the principal figure of merit, the changes in equivalent fuel worth associated with the fuel motion. The equivalent fuel worth was calculated from the measured axial fuel distributions by weighting the data with a typical LMFBR fuel-worth function. At nominal power, the initial fuel relocation resulted in increases in equivalent fuel worth. Above nominal power the fuel motion was dispersive, but the dispersive driving forces could not unequivocally be identified from the experimental data

  9. Radiation damage of UO{sub 2} fuel; Radijaciono ostecenje UO{sub 2} goriva

    Energy Technology Data Exchange (ETDEWEB)

    Stevanovic, M; Sigulinski, F [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    Radiation damage study of fuel and fuel elements covers: study of radiation damage methods in Sweden; analysis of testing the fuel and fuel elements at the RA reactor; feasibility study of irradiation in the Institute compared to irradiation abroad in respect to the reactor possibilities. Tasks included in this study are relater to testing of irradiated UO{sub 2} and ceramic fuel elements.

  10. Development of LWR Fuels with Enhanced Accident Tolerance

    International Nuclear Information System (INIS)

    Lahoda, Edward J.; Boylan, Frank A.

    2015-01-01

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company's Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15 N and U 3 Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3 Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3 Si 2 will represent a 15% increase in U235 loadings over those in UO fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3 Si 2 /68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2 , UN, and U 3 Si 2 the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2 . Cold spray application of either the amorphous steel or the Ti 2 AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the

  11. Development of LWR Fuels with Enhanced Accident Tolerance

    Energy Technology Data Exchange (ETDEWEB)

    Lahoda, Edward J. [Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States); Boylan, Frank A. [Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States)

    2015-10-30

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U3Si2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO2 pellets. Pellets and powders of UO2, UN, and U3Si2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO2. Cold spray application of either the amorphous steel or the Ti2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing

  12. Fuel damage during off-normal transients in metal-fueled fast reactors

    International Nuclear Information System (INIS)

    Kramer, J.M.; Bauer, T.H.

    1990-01-01

    Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

  13. Local governments' roles of the compensation for damage by the Tokai JCO criticality accident

    Energy Technology Data Exchange (ETDEWEB)

    Tanabe, Tomoyuki [Central Research Inst. of Electric Power Industry, Tokyo (Japan). Socio-Economic Research Center

    2003-03-01

    The Tokai JCO criticality accident on September 30, 1999 was the first case to which The Law on Compensation for Nuclear Damage was applied. Although the Law on Compensation for Nuclear Damage formulates the outline of the institutional framework for nuclear third party liability together with operator's insurance scheme, details of actual compensation procedure are not specified. By this reason, the compensation procedure in the Tokai accident had been executed without a concrete legal specification and a precedent. In spite of this situation, the compensation procedure with the accident led to an unexpectedly successful result. We observe the several reasons why the compensation procedure was implemented successfully despite the lack of concrete legal specification and a precedent. One of the reasons is that the local governments, Tokai Village and Ibaraki Prefecture, immediately took the leadership in implementing a temporary regime of compensation procedure without wasting time for waiting national government's directives. Upon practicing this compensation procedure, the local governments implemented the following steps. (1) Initial estimation of the amount and scope of damage. (2) Providing the criteria and heads of damage subject to compensation. (3) Unitary compensation procedure at the local levels. (4) Distribution of emergency payments for the victims. (5) Facilitating compensatory negotiation between the victims and JCO as arbitrator. However, some concerns are also pointed out about the fact that the local government directed the whole procedure without sufficient adjustment with the national government for compensation policy. Among all, in the compensation led by the local governments, it was difficult to guarantee fairness of compensation because victims who are influential on the local government such as industrial associations would have unfairly strong negotiation power in the compensatory negotiation, while the operator being

  14. Local governments' roles of the compensation for damage by the Tokai JCO criticality accident

    International Nuclear Information System (INIS)

    Tanabe, Tomoyuki

    2003-01-01

    The Tokai JCO criticality accident on September 30, 1999 was the first case to which The Law on Compensation for Nuclear Damage was applied. Although the Law on Compensation for Nuclear Damage formulates the outline of the institutional framework for nuclear third party liability together with operator's insurance scheme, details of actual compensation procedure are not specified. By this reason, the compensation procedure in the Tokai accident had been executed without a concrete legal specification and a precedent. In spite of this situation, the compensation procedure with the accident led to an unexpectedly successful result. We observe the several reasons why the compensation procedure was implemented successfully despite the lack of concrete legal specification and a precedent. One of the reasons is that the local governments, Tokai Village and Ibaraki Prefecture, immediately took the leadership in implementing a temporary regime of compensation procedure without wasting time for waiting national government's directives. Upon practicing this compensation procedure, the local governments implemented the following steps. (1) Initial estimation of the amount and scope of damage. (2) Providing the criteria and heads of damage subject to compensation. (3) Unitary compensation procedure at the local levels. (4) Distribution of emergency payments for the victims. (5) Facilitating compensatory negotiation between the victims and JCO as arbitrator. However, some concerns are also pointed out about the fact that the local government directed the whole procedure without sufficient adjustment with the national government for compensation policy. Among all, in the compensation led by the local governments, it was difficult to guarantee fairness of compensation because victims who are influential on the local government such as industrial associations would have unfairly strong negotiation power in the compensatory negotiation, while the operator being responsible for the

  15. Assessment of Neutronic Characteristics of Accident-Tolerant Fuel and Claddings for CANDU Reactors

    Directory of Open Access Journals (Sweden)

    Simon Younan

    2018-01-01

    Full Text Available The objective of this study was to evaluate accident-tolerant fuel (ATF concepts being considered for CANDU reactors. Several concepts, including uranium dioxide/silicon carbide (UO2-SiC composite fuel, dense fuels, microencapsulated fuels, and ATF cladding, were modelled in Serpent 2 to obtain reactor physics parameters, including important feedback parameters such as coolant void reactivity and fuel temperature coefficient. In addition, fuel heat transfer was modelled, and a simple accident model was tested on several ATF cases to compare with UO2. Overall, several concepts would require enrichment of uranium to avoid significant burnup penalties, particularly uranium-molybdenum (U-Mo and fully ceramic microencapsulated (FCM fuels. In addition, none of the fuel types have a significant advantage over UO2 in terms of overall accident response or coping time, though U-9Mo fuel melts significantly sooner due to its low melting point. Instead, the different ATF concepts appear to have more modest advantages, such as reduced fission product release upon cladding failure, or reduced hydrogen generation, though a proper risk assessment would be required to determine the magnitude of these advantages to weigh against economic disadvantages. The use of uranium nitride (UN enriched in N15 would increase exit burnup for natural uranium, providing a possible economic advantage depending on fuel manufacturing costs.

  16. Power Burst Facility Severe Fuel Damage test series

    International Nuclear Information System (INIS)

    Buescher, B.J.; Osetek, D.J.; Ploger, S.A.

    1982-01-01

    The Severe Fuel Damage (SFD) tests planned for the Power Burst Facility (PBF) are described. Bundles containing 32 zircaloy-clad, PWR-type fuel rods will be subjected to severe overheating transients in a high-pressure, superheated-steam environment. Cladding temperatures are expected to reach 2400 0 K, resulting in cladding ballooning and rupture, severe cladding oxidation, cladding melting, fuel dissolution, fuel rod fragmentation, and possibly, rubble bed formation. An experiment effluent collection system is being installed and the PBF fission product monitoring system is being upgraded to meet the special requirements of the SFD tests. Scoping calculations were performed to evaluate performance of the SFD test design and to establish operational requirements for the PBF loop

  17. Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

    International Nuclear Information System (INIS)

    1996-10-01

    This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

  18. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.

    1985-01-01

    SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

  19. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

    1986-01-01

    The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

  20. International collaboration for development of accident-resistant LWR fuel. International Collaboration for Development of Accident Resistant Light Water Reactor Fuel

    International Nuclear Information System (INIS)

    Sowder, Andrew

    2013-01-01

    Following the March 2011 multi-unit accident at the Fukushima Daiichi plant, there has been increased interest in the development of breakthrough nuclear fuel designs that can reduce or eliminate many of the outcomes of a severe accident at a light water reactor (LWR) due to loss of core cooling following an extended station blackout or other initiating event. With this interest and attention comes a unique opportunity for the nuclear industry to fundamentally change the nature and impact of severe accidents. Clearly, this is no small feat. The challenges are many and the technical barriers are high. Early estimates for moving maturing R and D concepts to the threshold of commercialisation exceed one billion USD. Given the anticipated effort and resources required, no single entity or group can succeed alone. Accordingly, the Electric Power Research Institute (EPRI) sees the need for and promise of cooperation among many stakeholders on an international scale to bring about what could be transformation in LWR fuel performance and robustness. An important initial task in any R and D programme is to define the goals and metrics for measuring success. As starting points for accident-tolerant fuel development, the extension of core coolability under loss of coolant conditions and the elimination or reduction of hydrogen generation are widely recognised R and D endpoints for deployment. Furthermore, any new LWR fuel technology will, at a minimum, need to (1) be compatible with the safe, economic operation of existing plants and (2) maintain acceptable or improve nuclear fuel performance under normal operating conditions. While the primary focus of R and D to date has been on cladding and fuel improvements, there are a number of other potential paths to improve outcomes following a severe accident at an LWR that include modifications to other fuel hardware and core internals to fully address core coolability, criticality, and hydrogen generation concerns. The US

  1. Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Su Xia

    2013-01-01

    The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

  2. Study of behavior of cermet fuel elements on IGR reactor under RIA type accident condition

    International Nuclear Information System (INIS)

    Vasil'ev, Yu.S.; Vurim, A.D.; Koltyshev, S.M.; Pakhnits, V.A.; Tukhvatulin, Sh.T.; Popov, V.V.; Ryzhkov, A.N.

    1996-01-01

    In 1993 December in IGR reactor of Inst. of Atomic Energy of National Nuclear Center of Republic of Kazakstan the second batch of in-pile testing of perspective cermet fuel elements under the condition, simulating RIA type accident was conducted. In the second batch of testing during eight start-ups 10 cermet fuel elements were examined. Among which 8 of monolith type and 2 fuel elements with false jacket beside cladding (FJF), as well as, 6 standard fuel elements of WWER-1000 type reactor with dioxide fuel were tested. 2 fuel elements - cermet and standard were placed into capsule filled with water. To measure energy release for the each start-up two fission monitor and inside core control gauge were placed. In all the start-ups operation mode of IGR was neutron pulse. Power of fuel element kept changing from 151 to 336 k W; energy release was 38-93 kJ/gr m 235 U; maximum temperature of cermet fuel was 1943-2173 K, of dioxide fuel - 1923-2843 K. The testing has demonstrated that operability of cermet fuel elements under reactivity accident condition with pulse width of 0,2 s is, at least, not less that operability of dioxide fuel elements, through advantages of cermet fuel under these conditions are revealed to the least extent

  3. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    International Nuclear Information System (INIS)

    Watanabe, Norio; Tamaki, Hitoshi

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  4. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Norio [Planning and Analysis Division, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Tamaki, Hitoshi [Department of Safety Research Technical Support, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  5. Accident Tolerant Fuel Concepts for Light Water Reactors. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2016-06-01

    Nuclear fuel is a highly complex material that has been subject to continuous development over the past 40 years and has reached a stage where it can be safely and reliably irradiated up to 65 GWd/tU in commercial nuclear reactors. During this time, there have been many improvements to the original designs and materials used. However, the basic design of uranium oxide fuel pellets clad with zirconium alloy tubing has remained the fuel choice for the vast majority of commercial nuclear power plants. Severe accidents, such as those at the Three Mile Island and Fukushima Daiichi have shown that under such extreme conditions, nuclear fuel will fail and the high temperature reactions between zirconoi alloys and water will lead to the generation of hydrogen, with the potential for explosions to occur, daming the plant further. Recognizing that the current fuel designs are vulnerable to severe accident conditions, tehre is renewed interesst in alternative fuel designs that would be more resistant to fuel failure and hydrogen production. Such new fuel designs will need to be compatible with existing fuel and reactor systems if they are to be utilized in the current reactor fleet and in current new build designs, but there is also the possibility of new designs for new reactor systems. This publication provides a record of the Technical Meeting on Accident Tolerant Fuel Concepts for Light Water Reactors, held at Oak Ridge National Laboratories (ORNL), United States of America, 13-16 October 2014, to consider the early stages of research and development into accident tolerant fuel. There were 45 participants from 10 countries taking part in the meeting, with 32 papers organized into 7 sessions, of which 27 are included in this publication. This meeting is part of a wider investigation into such designs, and it is anticipated that further Technical Meetings and research programmes will be undertaken in this field

  6. Models of fuel masses transition during second stage of the accident on Chernobyl NPP

    International Nuclear Information System (INIS)

    Tarapon, A.

    2002-01-01

    In ISPE NASU of Ukraine are developed mathematical models and software, which allow to research the processes of fuel masses transition during the accident at ChNPP. We found out, that the main reason of accident on ChNPP is the happening in the reactor of crisis of heat exchange of the second sort, instead of the effect positive output of reactivity from displacers of rods of system of emergency protection, as is accepted in official version

  7. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    OpenAIRE

    Bo Cheng; Young-Jin Kim; Peter Chou

    2016-01-01

    In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident managem...

  8. Cumulative damage fraction design approach for LMFBR metallic fuel elements

    International Nuclear Information System (INIS)

    Johnson, D.L.; Einziger, R.E.; Huchman, G.D.

    1979-01-01

    The cumulative damage fraction (CDF) analytical technique is currently being used to analyze the performance of metallic fuel elements for proliferation-resistant LMFBRs. In this technique, the fraction of the total time to rupture of the cladding is calculated as a function of the thermal, stress, and neutronic history. Cladding breach or rupture is implied by CDF = 1. Cladding wastage, caused by interactions with both the fuel and sodium coolant, is assumed to uniformly thin the cladding wall. The irradiation experience of the EBR-II Mark-II driver fuel with solution-annealed Type 316 stainless steel cladding provides an excellent data base for testing the applicability of the CDF technique to metallic fuel. The advanced metal fuels being considered for use in LMFBRs are U-15-Pu-10Zr, Th-20Pu and Th-2OU (compositions are given in weight percent). The two cladding alloys being considered are Type 316 stainless steel and a titanium-stabilized Type 316 stainless steel. Both are in the cold-worked condition. The CDF technique was applied to these fuels and claddings under the assumed steady-state operating conditions

  9. Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

    2004-11-01

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design

  10. Phenomena occurring in the reactor coolant system during severe core damage accidents

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1989-01-01

    The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab

  11. Safety demonstration analyses on criticality for severe accident during overland transport of fresh nuclear fuel

    International Nuclear Information System (INIS)

    Takahashi, Satoshi; Okuno, Hiroshi; Yamada, Kenji; Watanabe, Kouji; Nomura, Yasushi; Miyoshi, Yoshinori

    2005-01-01

    Criticality safety analysis was performed for transport packages of uranium dioxide powder or of fresh PWR fuel involved in a severe accident during overland transportation, and as a result, sub-criticality was confirmed against impact accident conditions such as loaded by a drop from high position to a concrete or asphalt surface, and fire accident conditions such as caused by collisions with an oil tank trailer carrying lots of inflammable material in open air, or with a commonly used two-ton-truck inside an unventilated tunnel. (author)

  12. Computational analysis of the behaviour of nuclear fuel under steady state, transient and accident conditions

    International Nuclear Information System (INIS)

    2007-12-01

    Accident analysis is an important tool for ensuring the adequacy and efficiency of the provision in the defence in depth concept to cope with challenges to plant safety. Accident analysis is the milestone of the demonstration that the plant is capable of meeting any prescribed limits for radioactive releases and any other acceptable limits for the safe operation of the plant. It is used, by designers, utilities and regulators, in a number of applications such as: (a) licensing of new plants, (b) modification of existing plants, (c) analysis of operational events, (d) development, improvement or justification of the plant operational limits and conditions, and (e) safety cases. According to the defence in depth concept, the fuel rod cladding constitutes the first containment barrier of the fission products. Therefore, related safety objectives and associated criteria are defined, in order to ensure, at least for normal operation and anticipated transients, the integrity of the cladding, and for accident conditions, acceptable radiological consequences with regard to the postulated frequency of the accident, as usually identified in the safety analysis reports. Therefore, computational analysis of fuel behaviour under steady state, transient and accident conditions constitutes a major link of the safety case in order to justify the design and the safety of the fuel assemblies, as far as all relevant phenomena are correctly addressed and modelled. This publication complements the IAEA Safety Report on Accident Analysis for Nuclear Power Plants (Safety Report Series No. 23) that provides practical guidance for establishing a set of conceptual and formal methods and practices for performing accident analysis. Computational analysis of the behaviour of nuclear fuel under transient and accident conditions, including normal operation (e.g. power ramp rates) is developed in this publication. For design basis accidents, depending on the type of influence on a fuel element

  13. Mechanistic model for Sr and Ba release from severely damaged fuel

    International Nuclear Information System (INIS)

    Rest, J.; Cronenberg, A.W.

    1985-11-01

    Among radionuclides associated with fission product release during severe accidents, the primary ones with health consequences are the volatile species of I, Te, and Cs, and the next most important are Sr, Ba, and Ru. Considerable progress has been made in the mechanistic understanding of I, Cs, Te, and noble gas release; however, no capability presently exists for estimating the release of Sr, Ba, and Ru. This paper presents a description of the primary physical/chemical models recently incorporated into the FASTGRASS-VFP (volatile fission product) code for the estimation of Sr and Ba release. FASTGRASS-VFP release predictions are compared with two data sets: (1) data from out-of-reactor induction-heating experiments on declad low-burnup (1000 and 4000 MWd/t) pellets, and (2) data from the more recent in-reactor PBF Severe Fuel Damage Tests, in which one-meter-long, trace-irradiated (89 MWd/t) and normally irradiated (approx.35,000 MWd/t) fuel rods were tested under accident conditions. 10 refs

  14. Analysis of metal fuel transient overpower experiments with the SAS4A accident analysis code

    International Nuclear Information System (INIS)

    Tentner, A.M.; Kalimullah; Miles, K.J.

    1990-01-01

    The results of the SAS4A analysis of the M7 TREAT Metal fuel experiment are presented. New models incorporated in the metal fuel version of SAS4A are described. The computational results are compared with the experimental observations and this comparison is used in the interpretation of physical phenomena. This analysis was performed using the integrated metal fuel SAS4A version and covers a wide range of events, providing an increased degree of confidence in the SAS4A metal fuel accident analysis capabilities

  15. Review of the accident source terms for aluminide fuel: Application to the BR2 reactor

    International Nuclear Information System (INIS)

    Joppen, F.

    2005-01-01

    A major safety review of the BR2, a material test reactor, is to be conducted for the year 2006. One of the subjects selected for the safety review is the definition of source terms for emergency planning and in particular the development of accident scenarios. For nuclear power plants the behaviour of fuel under accident conditions is a well studied object. In case of non-power reactors this basic knowledge is rather scarce. The usefulness of information from power plant fuels is limited due to the differences in fuel type, power level and thermohydraulical conditions. First investigation indicates that using data from power plant fuel leads to an overestimation of the source terms. Further research on this subject could be very useful for the research reactor community, in order to define more realistic source terms and to improve the emergency preparedness. (author)

  16. Analytical criteria for fuel failure modes observed in reactivity initiated accidents

    International Nuclear Information System (INIS)

    Luxat, J.C.

    2005-01-01

    The behaviour of nuclear fuel subjected to a short duration power pulse is of relevance to LWR and CANDU reactor safety. A Reactivity Initiated Accident (RIA) in an LWR would subject fuel to a short duration power pulse of large amplitude, whereas in CANDU a large break Loss of Coolant Accident (LOCA) would subject fuel to a longer duration, lower amplitude power excursion. The energy generated in the fuel during the power pulse is a key parameter governing the fuel response. This paper reviews the various power pulse tests that have been conducted in research reactors over the past three decades and summarizes the fuel failure modes that that have been observed in these tests. A simple analytical model is developed to characterize fuel behaviour under power pulse conditions and the model is applied to assess the experimental data from the power pulse tests. It is shown that the simple model provides a good basis for establishing criteria that demarcate the observed fuel failure modes for the various fuel designs that have been used in these tests. (author)

  17. Behaviour of rock-like oxide fuels under reactivity-initiated accident conditions

    International Nuclear Information System (INIS)

    Kazuyuki, Kusagaya; Takehiko, Nakamura; Makio, Yoshinaga; Hiroshi, Akie; Toshiyuki, Yamashita; Hiroshi, Uetsuka

    2002-01-01

    Pulse irradiation tests of three types of un-irradiated rock-like oxide (ROX) fuel - yttria-stabilised zirconia (YSZ) single phase, YSZ and spinel (MgAl 2 O 4 ) homogeneous mixture and particle-dispersed YSZ/spinel - were conducted in the Nuclear Safety Research Reactor to investigate the fuel behaviour under reactivity-initiated accident conditions. The ROX fuels failed at fuel volumetric enthalpies above 10 GJ/m 3 , which was comparable to that of un-irradiated UO 2 fuel. The failure mode of the ROX fuels, however, was quite different from that of the UO 2 fuel. The ROX fuels failed with fuel pellet melting and a part of the molten fuel was released out to the surrounding coolant water. In spite of the release, no significant mechanical energy generation due to fuel/coolant thermal interaction was observed in the tested enthalpy range below∼12 GJ/m 3 . The YSZ type and homogenous YSZ/spinel type ROX fuels failed by cladding burst when their temperatures peaked, while the particle-dispersed YSZ/spinel type ROX fuel seemed to have failed by cladding local melting. (author)

  18. Post-accident fuel relocation and heat removal in the LMFBR

    International Nuclear Information System (INIS)

    Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

    1976-08-01

    Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

  19. Computational Aerodynamics of Shuttle Orbiter Damage Scenarios in Support of the Columbia Accident Investigation

    Science.gov (United States)

    Bibb, Karen L.; Prabhu, Ramadas K.

    2004-01-01

    In support of the Columbia Accident Investigation, inviscid computations of the aerodynamic characteristics for various Shuttle Orbiter damage scenarios were performed using the FELISA unstructured CFD solver. Computed delta aerodynamics were compared with the reconstructed delta aerodynamics in order to postulate a progression of damage through the flight trajectory. By performing computations at hypervelocity flight and CF4 tunnel conditions, a bridge was provided between wind tunnel testing in Langley's 20-Inch CF4 facility and the flight environment experienced by Columbia during re-entry. The rapid modeling capability of the unstructured methodology allowed the computational effort to keep pace with the wind tunnel and, at times, guide the wind tunnel efforts. These computations provided a detailed view of the flowfield characteristics and the contribution of orbiter components (such as the vertical tail and wing) to aerodynamic forces and moments that were unavailable from wind tunnel testing. The damage scenarios are grouped into three categories. Initially, single and multiple missing full RCC panels were analyzed to determine the effect of damage location and magnitude on the aerodynamics. Next is a series of cases with progressive damage, increasing in severity, in the region of RCC panel 9. The final group is a set of wing leading edge and windward surface deformations that model possible structural deformation of the wing skin due to internal heating of the wing structure. By matching the aerodynamics from selected damage scenarios to the reconstructed flight aerodynamics, a progression of damage that is consistent with the flight data, debris forensics, and wind tunnel data is postulated.

  20. Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Su, S.F.; Cahalan, J.E.; Sevy, R.H.

    1985-01-01

    This paper presents the results of a systematic study of the effectiveness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). The results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs, and safety margins are quantified in sensitivity studies. All analyses were carried out using the SASSYS LMFBR systems analysis code (1)

  1. Radiological health risks from accidents during transportation of spent nuclear fuels

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1988-01-01

    Potential radiological health risks from severe accident scenarios during the transportation of spent nuclear fuels are estimated. These extremely low probability, but potentially credible, scenarios are characterized by the U.S. Nuclear Regulatory Commission's Modal Study in terms of the maximum credible structural responses and/or the maximum credible cask temperature responses. In some accident scenarios, the spent nuclear fuel casks are assumed to be breached, resulting in the release of radioactivity to the atmosphere. Models have been developed to estimate radiological health consequences, including potential short-term exposures and health effects to individuals and potential long-term environmental dose commitments and health effects to the population. The population risks are calculated using state-level data, and the resulting overall health risks are compared for several levels of cleanup effort to determine the relative effects on long-term risks to the population in the event of an accident. 4 refs., 3 figs., 3 tabs

  2. Thermal hydraulic features of the TMI accident

    International Nuclear Information System (INIS)

    Tolman, B.

    1985-01-01

    The TMI-2 accident resulted in extensive core damage and recent data confirms that the reactor vessel was challenged from molten core materials. A hypothesized TMI accident sencario is presented that consistently explains the TMI data and is also consistent with research findings from independent severe fuel damage experiements. The TMI data will prove useful in confirming our understanding of severe core damage accidents under realistic reactor systems conditions. This understanding will aid in addressing safety and regulatory issues related to severe core damage accidents in light water reactors

  3. Questionnaire survey report about the criticality accident at a nuclear fuel processing facility

    International Nuclear Information System (INIS)

    2000-01-01

    The Radiation Protection Section of the Japanese Society of Radiological Technology conducted a questionnaire survey on the criticality accident at the nuclear fuel processing facility in Tokai village on September 30, 1999 in order to identify factors related to the accident and consider countermeasures to deal with such accidents. The questionnaire was distributed to 347 members (122 facilities) of the Japanese Society of Radiological Technology who were working or living in Ibaraki Prefecture, and replies were obtained from 104 members (75 facilities). Questions to elicit the opinions of individuals were as following: method of obtaining information about the accident, knowledge about radiation, opinions about the accident, and requests directed to the Society. Questions regarding facilities concerned the following: communication after the accident, requests for dispatch to the accident site, and possession of radiometry devices. In regard to acquisition of information, 91 of the 104 members (87.5%) answered 'television or radios' followed by newspapers. Forty-five of 101 members were questioned about radiation exposure and radiation effects by the public. There were many opinions that accurate news should be provided rapidly, by the mass media. Many members (75%) felt that they lacked knowledge about radiation, reconfirming the importance of education and instruction concerning radiation. Dispatch was requested of 36 of the 75 facilities (48%), and 44 of 83 facilities (53%) owned radiometry instruments. (K.H.)

  4. Comparison of the Transportation Risks Resulting from Accidents during the Transportation of the Spent Fuel

    International Nuclear Information System (INIS)

    Jeong Jong Tae; Cho, Dong Kuen; Choi, Heui Joo; Choi, Jong Won

    2007-01-01

    The safe, environmentally sound and publicly acceptable disposal of high level wastes and spent fuels is becoming a very important issue. The operational safety assessment of a repository including a transportation safety assessment is a fundamental part in order to achieve this goal. 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 (CISF) which is to start operation in 2016. Therefore, we have to determine the safe and economical logistics for the transportation of these spent fuels by considering their transportation risks and costs. In this study, we developed four transportation scenarios by considering the type of transportation casks and transport means in order to suggest safe and economical transportation logistics for spent fuels. Also, we estimated and compared the transportation risks resulting from the accidents during the transportation of spent fuels for these four transportation scenarios

  5. Spent nuclear fuel structural response when subject to an end impact accident

    Energy Technology Data Exchange (ETDEWEB)

    Tang, D.T.; Guttmann, J. [United States Nuclear Regulatory Commission, Rockville, MD (United States)]|[United States Nuclear Regulatory Commission, Washington, DC (United States); Koeppel, B.J.; Adkins, H.E.

    2004-07-01

    The US Nuclear Regulatory Commission (USNRC) is responsible for licensing spent fuel storage and transportation systems. A subset of this responsibility is to investigate and understand the structural performance of these systems. Studies have shown that the fuel rods of intact spent fuel assemblies with burn-ups up to 45 gigawatt days per metric ton of uranium (Gwd/MTU) are capable of resisting the normally expected impact loads subjected during drop accident conditions. However, effective cladding thickness for intact spent fuel assemblies with burn ups greater than 45 Gwd/MTU can be reduced due to corrosion. The capability of the fuel rod to withstand the expected loads encountered under normal and accident conditions may also be reduced, given degradation of the material properties under extended use, such as decrease in ductility. The USNRC and Pacific Northwest Laboratory (PNNL) performed computational studies to predict the structural response of spent nuclear fuel in a transport system that is subjected to a hypothetical regulatory impact accident, as defined in 10 CFR71.73. This study performs a structural analysis of a typical high burn up Pressurized Water Reactor (PWR) fuel assembly using the ANSYS {sup registered} ANSYS {sup registered} /LS- DYNA {sup registered} finite element analysis (FEA) code. The material properties used in the analyses were based on expert judgment and included uncertainties. Ongoing experimental programs will reduce the uncertainties. The current evaluations include the pins, spacer grids, and tie plates to assess possible cladding failure/rupture under hypothetical impact accident loading. This paper describes the USNRC and PNNL staff's analytical approach, provides details on the single pin model developed for this assessment, and presents the results.

  6. Spent fuel transport cask thermal evaluation under normal and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Pugliese, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Lo Frano, R., E-mail: rosa.lofrano@ing.unipi.i [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy); Forasassi, G. [Department of Mechanical, Nuclear and Production Engineering, University of Pisa, Via Diotisalvi, no 2-56126 Pisa (Italy)

    2010-06-15

    The casks used for transport of nuclear materials, especially the spent fuel element (SPE), must be designed according to rigorous acceptance criteria and standards requirements, e.g. the International Atomic Energy Agency ones, in order to provide protection to people and environment against radiation exposure particularly in a severe accident scenario. The aim of this work was the evaluation of the integrity of a spent fuel cask under both normal and accident scenarios transport conditions, such as impact and rigorous fire events, in according to the IAEA accident test requirements. The thermal behaviour and the temperatures distribution of a Light Water Reactor (LWR) spent fuel transport cask are presented in this paper, especially with reference to the Italian cask designed by AGN, which was characterized by a cylindrical body, with water or air inside the internal cavity, and two lateral shock absorbers. Using the finite element code ANSYS a series of thermal analyses (steady-state and transient thermal analyses) were carried out in order to obtain the maximum fuel temperature and the temperatures field in the body of the cask, both in normal and in accidents scenario, considering all the heat transfer modes between the cask and the external environment (fire in the test or air in the normal conditions) as well as inside the cask itself. In order to follow the standards requirements, the thermal analyses in accidents scenarios were also performed adopting a deformed shape of the shock absorbers to simulate the mechanical effects of a previous IAEA 9 m drop test event. Impact tests on scale models of the shock absorbers have already been conducted in the past at the Department of Mechanical, Nuclear and Production Engineering, University of Pisa, in the '80s. The obtained results, used for possible new licensing approval purposes by the Italian competent Authority of the cask for PWR spent fuel cask transport by the Italian competent Authority, are

  7. Severe fuel damage in steam and helium environments observed in in-reactor experiments

    International Nuclear Information System (INIS)

    Saito, S.; Shiozawa, S.

    1984-01-01

    The bahavior of severe fuel damages has been studied in gaseous environments simulating core uncovery accidents in the in-reactor experiments utilizing the NSRR. Two types of cladding relocation modes, azimuthal flow and melt-down, were revealed through the parametric experiments. The azimuthal flow was evident in an oxidizing environment in case of no oxide film break. The melt-down can be categorized into flow-down and move-down, according to the velocity of the melt-down. Cinematographies showed that the flow-down was very fast as water flows down while the move-down appeared to be much slower. The flow-down was possible in an unoxidizing environment, whereas the move-down of molten cladding occured through a crack induced in an oxide film in an oxidizing environment. The criterion of the relocation modes was developed as a function of peak cladding temperature and oxidation condition. It was also found that neither immediate quench nor fuel fracture occurred upon flooding when cladding temperature was about 1800 0 C at water injection. The external mechanical force is needed for fuel fracture. (orig.)

  8. Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

    2015-08-15

    Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

  9. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  10. Trial evaluation on criticality safety of the fuel assemblies at falling accident as spent fuel transport and storage cask

    International Nuclear Information System (INIS)

    Tadano, Tomoaki

    2016-01-01

    The authors conducted critical safety assessment on the supposed event at the time of a fall accident of cask, and examined the influence on criticality safety. If the spacer of fuel assembly is sound, it is assumed that the pitch of fuel rod interval changes, and if the spacer is broken, it is assumed that the fuel rod is unevenly distributed in the basket. For the critical calculation of fuel assembly basket system, they performed it using a calculation code. For both of the single cell and assembly, calculation results showed an increase in the effective multiplication factor of reactivity of 2-3%. When this reactivity is applied to the criticality analysis result of PWR fuel assembly, the value approaches to the limit 0.95 of the neutron effective multiplication factor keff. However, the keff when new fuel is loaded is sufficiently lower than 0.93. Therefore, it is unlikely that the criticality analysis result approaches to 0.95 at all burnups, and the possibility to become criticality is very low in actual spent fuel transport. When considering the reactivity of this research, it is possible that the design condition for the assumption of novel fuel loading becomes severer. Furthermore, criticality analysis under non-uniform pitch will become necessary, and criticality safety analysis for BWR fuel with heterogeneous enrichment degree and burnup degree will become also necessary. (A.O.)

  11. Analysis of reactivity accidents of the RSG-GAS core with silicide fuel

    International Nuclear Information System (INIS)

    Tukiran

    2002-01-01

    The fuels of RSG-GAS reactor is changed from uranium oxide to uranium silicide. For time being, the fuel of RSG-GAS core are mixed up between oxide and silicide fuels with 250 gr of loading and 2.96 g U/cm 3 of density, respectively. While, silicide fuel with 300 gr of loading is still under research. The advantages of silicide fuels are can be used in high density, so that, it can be stayed longer in the core at higher burn-up, therefore, the length of cycle is longer. The silicide fuel in RSG-GAS core is used in step-wise by using mixed up core. Firstly, it is used silicide fuel with 250 gr of loading and then, silicide fuel with 300 gr of loading (3.55 g U/cm 3 of density). In every step-wise of fuel loading must be analysed its safety margin. In this occasion, it is analysed the reactivity accident of RSG-GAS core with 300 gr of silicide fuel loading. The calculation was done by using POKDYN code which available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. From all cases which were have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 gr silicide fuel loading

  12. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Cahalan, J.; Wigeland, R.; Friedel, G.; Kussmaul, G.; Royl, P.; Moreau, J.; Perks, M.

    1990-01-01

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs

  13. Fuel assembly stress and deflection analysis for loss-of-coolant accident and seismic excitation

    International Nuclear Information System (INIS)

    DeMars, R.V.; Steinke, R.R.

    1975-01-01

    Babcock and Wilcox has evaluated the capability of the fuel assemblies to withstand the effects of a loss-of-coolant accident (LOCA) blowdown, the operational basis earthquake (OBE) and design basis earthquake (DBE), and the simultaneous occurrence of the DBE and LOCA. This method of analysis is applicable to all of B and W's nuclear steam system contracts that specify the skirt-supported pressure vessel. Loads during the saturated and subcooled phases of blowdown following a loss-of-coolant accident were calculated. The maximum loads on the fuel assemblies were found to be below allowable limits, and the maximum deflections of the fuel assemblies were found to be less than those that could prevent the insertion of control rods or the flow of coolant through the core. (U.S.)

  14. Loss of cooling accident simulation of nuclear power station spent-fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.; Liang, K-S., E-mail: mlee@ess.nthu.edu.tw, E-mail: ksliang_1@hotmail.com [National Tsing Hua Univ., Hsinchu, Taiwan (China); Lin, K-Y., E-mail: syrup760914@gmail.com [Taiwan Power Company, Taiwan (China)

    2014-07-01

    The core melt down accident of Fukushima Nuclear Power Station on March 11th, 2011 alerted nuclear industry that the long term loss of cooling of spent fuel pool may need some attention. The target plant analyzed is the Chinshan Nuclear Power Station of Taiwan Power Company. The 3-Dimensional RELAP5 input deck of the spent fuel pool of the station is built. The results indicate that spent fuel of Chinshan Nuclear Power Station is uncovered at 6.75 days after an accident of loss cooling takes place and cladding temperature rises above 2,200{sup o}F around 8 days. The time is about 13 hours earlier than the results predicted using simple energy balance method. The results also show that the impact of Counter Current Flow Limitation (CCFL) and radiation heat transfer model is marginal. (author)

  15. Mechanical energy release and fuel fragmentation in high energy deposition into fuel under a reactivity initiated accident condition

    International Nuclear Information System (INIS)

    Tsuruta, Takaharu; Saito, Shinzo; Ochiai, Masaaki

    1985-01-01

    The fuel fragmentation is one of important subjects to be studied, since it is one of basic processes of molten fuel-coolant interaction (MFCI) and it has not yet been made clear enough. Accordingly, UO 2 fuel fragmentation was studied in the NSRR experiments simulating a reactivity initiated accident (RIA). As results of the experiments, the distribution of the size of fuel fragments was obtained and the mechanism of fuel fragmentation was discussed as described below. It was revealed that the distribution was well displayed in the form of logarithmic Rosin-Rammler's distribution law. It was shown that the conversion ratio from thermal energy to mechanical in the experiment was in inverse propotion to the volume-surface mean diameter defined as a ratio of the total volume of fragments to the total surface. Consequently, it was confirmed that the mean diameter was proper as an index for the degree of the fuel fragmentation. It was also pointed out that the Weber-type hydraulic instability model for fragmentation was consistent with the experimental results. The mechanism of the fuel fragmentation is understood as follows. Cladding tube is ruptured due to the increase in rod pressure when fuel is molten, and then molten fuel spouts through the openings in the form of jet. As a result of molten fuel spouting, fuel is fragmented by the Weber-type of hydraulic instability. The model well explains the effects of experimental parameters as heat deposition, subcooling of cooling water and capsule diameter, on the fuel fragmentation. According to the model, fuel fragments have to be spherical. There were many spherical particles which had hollow and burst crack. This may be due to internal burst during solidification process. The items which should be studied further are also described in the end of this report. (author)

  16. Development of Cr cold spray–coated fuel cladding with enhanced accident tolerance

    Directory of Open Access Journals (Sweden)

    Martin Ševeček

    2018-03-01

    Full Text Available Accident-tolerant fuels (ATFs are currently of high interest to researchers in the nuclear industry and in governmental and international organizations. One widely studied accident-tolerant fuel concept is multilayer cladding (also known as coated cladding. This concept is based on a traditional Zr-based alloy (Zircaloy-4, M5, E110, ZIRLO etc. serving as a substrate. Different protective materials are applied to the substrate surface by various techniques, thus enhancing the accident tolerance of the fuel. This study focuses on the results of testing of Zircaloy-4 coated with pure chromium metal using the cold spray (CS technique. In comparison with other deposition methods, e.g., Physical vapor deposition (PVD, laser coating, or Chemical vapor deposition techniques (CVD, the CS technique is more cost efficient due to lower energy consumption and high deposition rates, making it more suitable for industry-scale production. The Cr-coated samples were tested at different conditions (500°C steam, 1200°C steam, and Pressurized water reactor (PWR pressurization test and were precharacterized and postcharacterized by various techniques, such as scanning electron microscopy, Energy-dispersive X-ray spectroscopy (EDX, or nanoindentation; results are discussed. Results of the steady-state fuel performance simulations using the Bison code predicted the concept's feasibility. It is concluded that CS Cr coating has high potential benefits but requires further optimization and out-of-pile and in-pile testing. Keywords: Accident-Tolerant Fuel, Chromium, Cladding, Coating, Cold Spray, Nuclear Fuel

  17. Development of likelihood estimation method for criticality accidents of mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tamaki, Hitoshi; Yoshida, Kazuo; Kimoto, Tatsuya; Hamaguchi, Yoshikane

    2010-01-01

    A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility. (author)

  18. Routine methods for post-transportation accident recovery of spent fuel casks

    International Nuclear Information System (INIS)

    Shappert, L.B.; Pope, R.B.; Best, R.E.; Jones, R.H.

    1991-01-01

    Spent fuel casks and other large radioactive material packages have been examined to determine whether the designs are adequate to allow the casks to be recovered using conventional recovery methods following a transportation accident. Casks and similar packages are typically designed with, and handled by, trunnions that support the package during transport. These trunnions are considered the best cask feature with which to grapple the cask once it is no longer in its usual shipping mode. Following a transport accident, the trunnions may be buried or entangled so that they are not readily accessible to initiate the recovery process. To evaluate the effectiveness of applying traditional recovery methods to spent fuel casks, a workshop was held in which a series of accidents involving casks were postulated; the modes of transportation considered included truck, rail, and barge. These participants knowledgeable in transport, handling, and, in some cases, recovery of large, heavy containers attended. Participants concluded that the physical recovery of a cask involved in an accident, irrespective of where the accident occurs, would be a straightforward rigging operation and that the addition of specific recovery features (e.g., additional trunnions) to the cask appears unnecessary

  19. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  20. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  1. Classification of the railway accident in accordance with the requirement of the safety analysis of transporting spent fuel

    International Nuclear Information System (INIS)

    Wu Tao

    1993-01-01

    Based on the analysis of the difference between the accident severity categorization used in the Ministry of Railway and that used in the safety analysis of the transporting spent fuel, a method used for the classification of the railway accident in accordance with the requirement of the safety analysis of transporting spent fuel is suggested. The method classifies the railway accidents into 10 scenarios and make it possible to scale the accident through directly using the data documented by the Ministry of Railway without any additional effort

  2. Dynamic modeling of physical phenomena for probabilistic assessment of spent fuel accidents

    International Nuclear Information System (INIS)

    Benjamin, A.S.

    1997-01-01

    If there should be an accident involving drainage of all the water from a spent fuel pool, the fuel elements will heat up until the heat produced by radioactive decay is balanced by that removed by natural convection to air, thermal radiation, and other means. If the temperatures become high enough for the cladding or other materials to ignite due to rapid oxidation, then some of the fuel might melt, leading to an undesirable release of radioactive materials. The amount of melting is dependent upon the fuel loading configuration and its age, the oxidation and melting characteristics of the materials, and the potential effectiveness of recovery actions. The authors have developed methods for modeling the pertinent physical phenomena and integrating the results with a probabilistic treatment of the uncertainty distributions. The net result is a set of complementary cumulative distribution functions for the amount of fuel melted

  3. Behavior of irradiated ATR/MOX fuel under reactivity initiated accident conditions (Joint research)

    International Nuclear Information System (INIS)

    Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi

    2000-03-01

    Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20 MWd/kgHM were conducted at the NSRR in JAERI to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Deformation of the fuel rods due to PCMI occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet. (author)

  4. Precursors to potential severe core damage accidents: 1996. A status report. Volume 25

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D.; Dolan, B.W.; Minarick, J.W.

    1997-12-01

    This report describes the 14 operational events in 1996 that affected 13 commercial light-water reactors and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by first computer-screening the 1996 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1995 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  5. Precursors to potential severe core damage accidents: 1997 - A status report. Volume 26

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Muhlheim, M.D.; Dolan, B.W.; Minarick, J.W.

    1998-11-01

    This report describes the five operational events in 1997 that affected five commercial light-water reactors (LWRs) and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by first computer-screening the 1997 licensee event reports from commercial LWRs to identify those events that could be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1996 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  6. Organizing irresponsibility? The (inter)national management of a nuclear accident damages as discursive regime

    International Nuclear Information System (INIS)

    Topcu, Sezin

    2014-01-01

    This article analyzes the historical process related to the international organization of responsibilities and the management of the damages in case of a nuclear disaster. The author shows that the political and legal settings on which the discourse of an 'international regime of civil responsibility' (that emerged in the 1960's) relies, have globally aimed at maintaining a 'historical and spectacular gap' between the damages the nuclear operators are taking responsibility for, and the real and extensive damages engendered by a major accident. She argues that the existence of such a 'gap' is inherent to the nuclear sector, that it is a form of government (both of economic affairs and of the public space) which was historically constructed, and that the existence of such a gap is crucial for the survival of the nuclear industry itself. Thus the notion of 'responsibility' in the nuclear sector appears to serve mainly as a discursive regime, as a means to organize not only responsibilities but also irresponsibilities, whatever the geographic scale (national or international) at which they should be managed

  7. Thermal analyses for the spend fuel pool of Taiwan BWR plants during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Chen, B-Y.; Yeh, C-L.; Wei, W-C.; Chen, Y-S., E-mail: onepicemine@iner.gov.tw, E-mail: clinyeh@iner.gov.tw, E-mail: hn150456@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research, Longtan Township, Taoyuan County, Taiwan (China)

    2014-07-01

    After the Fukushima nuclear accident, the safety of the spent fuel pool has become an important concern. In this study, thermal analysis of the spent fuel pool under a loss of cooling accident is performed. The BWR spent fuel pools in Taiwan are investigated, including the Chinshan, Kuosheng, and Lungmen plants. The transient pool temperature and level behaviors are calculated based on lumped energy balance. After the pool level drops below the top of the fuel, the peak cladding temperature is predicted by the Computational Fluid Dynamics (CFD) analysis. The influence to the cladding temperature of the uniform and checkboard fuel loading patterns is also investigated. (author)

  8. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

    1992-01-01

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  9. Accident Testing of High Temperature Reactor Fuel Elements with the KueFA Device

    International Nuclear Information System (INIS)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Ferreira-Teixeira, A.E.; Van Winckel, S.; Rondinella, V.V.; Allelein, H.J.

    2013-06-01

    The High Temperature Reactor (HTR) is characterised by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with Tri-Isotropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1600-1650 deg. C, remaining well below the melting point of the fuel. The Cold Finger Apparatus (KueFA) is used to observe the combined effects of Depressurization and Loss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in He atmosphere for several hundred hours, mimicking accident temperatures up to 1800 deg. C and realistic temperature transients. Non-gaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analysing plate deposits by means of HPGe gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. In order to achieve a good quantification of the release, a careful calibration of the setup is mandatory. An especially tailored collimator was designed to perform plate scanning with high spatial resolution, thus yielding information about the fission product distribution on the condensation plates. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Chemical dissolution has been performed for some condensation plates in order to assess beta nuclides of interest such as 90 Sr and possibly 129 I using an Inductively Coupled Plasma - Mass Spectrometer (ICP-MS) and to cross check the HPGe gamma spectroscopy measurements

  10. Potential biological indicators of multi-organ damage: Application to radiation accident victims

    International Nuclear Information System (INIS)

    Bertho, J.M.; Souidi, M.; Gourmelon, P.

    2009-01-01

    Accidental irradiations induce a complex pathological situation, difficult to assess and to treat. However, recent results describing new biological indicators of radiation-induced damages such as Flt3-ligand, citrulline and oxy-sterol concentration in the plasma, together with results obtained in large animal models of high dose irradiation, allowed a better understanding of pathophysiological mechanisms induced by uncontrolled irradiations. This conducted to leave the classical paradigm of the acute radiation syndrome, described as the association of three individual syndromes, the hematopoietic syndrome, the gastro-intestinal syndrome and the cerebrovascular syndrome, in favour of a multiple organ dysfunction syndrome, with the implication of other organs and systems. Follow-up of victims from two recent radiation accidents brings a confirmation of the usefulness of the newly described biological indicators, and also a partial confirmation of this new concept of a multiple organ dysfunction syndrome. (authors)

  11. Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

    International Nuclear Information System (INIS)

    Jernkvist, Lars Olof; Massih, Ali R.

    2004-12-01

    Core failure limits for high-burnup light water reactor UO 2 fuel rods, subjected to postulated reactivity initiated accidents (RIAs), are here assessed by use of best-estimate computational methods. The considered RIAs are the hot zero power rod ejection accident (HZP REA) in pressurized water reactors and the cold zero power control rod drop accident (CZP CRDA) in boiling water reactors. Burnup dependent core failure limits for these events are established by calculating the fuel radial average enthalpy connected with incipient fuel pellet melting for fuel burnups in the range of 30 to 70 MWd/kgU. The postulated HZP REA and CZP CRDA result in lower enthalpies for pellet melting than RIAs that take place at rated power. Consequently, the enthalpy thresholds presented here are lower bounds to RIAs at rated power. The calculations are performed with best-estimate models, which are applied in the FRAPCON-3.2 and SCANAIR-3.2 computer codes. Based on the results of three-dimensional core kinetics analyses, the considered power transients are simulated by a Gaussian pulse shape, with a fixed width of either 25 ms (REA) or 45 ms (CRDA). Notwithstanding the differences in postulated accident scenarios between the REA and the CRDA, the calculated core failure limits for these two events are similar. The calculated enthalpy thresholds for fuel pellet melting decrease gradually with fuel burnup, from approximately 960 J/gUO 2 at 30 MWd/kgU to 810 J/gUO 2 at 70 MWd/kgU. The decline is due to depression of the UO 2 melting temperature with increasing burnup, in combination with burnup related changes to the radial power distribution within the fuel pellets. The presented fuel enthalpy thresholds for incipient UO 2 melting provide best-estimate core failure limits for low- and intermediate-burnup fuel. However, pulse reactor tests on high-burnup fuel rods indicate that the accumulation of gaseous fission products within the pellets may lead to fuel dispersal into the coolant at

  12. Safety technical investigation activities for shipment of damaged spent fuels from Fukushima Daiichi Nuclear Power Station

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Japan Nuclear Energy Safety Organization(JNES) carries out the investigation for damaged fuel transportation from Fukushima Daiichi Nuclear Power Station(1F) under safety condition to support Nuclear Regulation Authority (NRA). In 2012 fiscal year, JNES carried out the investigation of spent fuel condition in unit 4 of 1F and actual result of leak fuel transport in domestic /other countries. From this result, Package containing damaged fuel from unit 4 in 1F were considered. (author)

  13. Dose calculation for accident situations at WWR-S type spent nuclear fuel repository

    International Nuclear Information System (INIS)

    Margeanu, S.; Florescu, G.

    2006-01-01

    Full text: The Spent Nuclear Fuel Repository at IFIN-HH Bucharest (SNFR IFIN-HH) consists in four pools, repository hall, radiological monitoring system, ventilation system and auxiliary systems. At the moment the remaining activity in the repository is about 3500 Ci. Despite of the small activity, for emergency preparedness purposes, several accident scenarios, with a non zero probability of occurrence during the repository lifetime, have been postulated. Evaluations of radiological consequences to personnel, general public and environment, for each accident scenario have been performed. The radioactive inventory was evaluated with ORIGEN code from SCALE computer code system and radiological consequences were evaluated with COSYMA computer code. Assumptions for the source term determination, meteorological conditions and release, are presented. The calculated values of doses and risk are also presented. The impact of these accident scenarios on population and environment is also discussed. (authors)

  14. Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

    International Nuclear Information System (INIS)

    Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

    2013-01-01

    The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

  15. Present status and needs of research on severe core damage

    International Nuclear Information System (INIS)

    1982-05-01

    The needs for research on severe core damage accident have been emphasized recently, in particular, since TMI-2 accident. The Severe Core Damage Research Task Force was established by the Divisions of Reactor Safety and Reactor Safety Evaluation to evaluate individual phenomenon, to survey the present status of research and to provide the recommended research subjects on severe accidents. This report describes the accident phenomena involving some analytical results, status of research and recommended research subjects on severe core damage accidents, divided into accident sequence, fuel damage, and molten material behavior, fission product behavior, hydrogen generation and combustion, steam explosion and containment integrity. (author)

  16. Fission product release from HTGR fuel under core heatup accident conditions - HTR2008-58160

    International Nuclear Information System (INIS)

    Verfondern, K.; Nabielek, H.

    2008-01-01

    Various countries engaged in the development and fabrication of modern fuel for the High Temperature Gas-Cooled Reactor (HTGR) have initiated activities of modeling the fuel and fission product release behavior with the aim of predicting the fuel performance under operating and accidental conditions of future HTGRs. Within the IAEA directed Coordinated Research Project CRP6 on 'Advances in HTGR Fuel Technology Development' active since 2002, the 13 participating Member States have agreed upon benchmark studies on fuel performance during normal operation and under accident conditions. While the former has been completed in the meantime, the focus is now on the extension of the national code developments to become applicable to core heatup accident conditions. These activities are supported by the fact that core heatup simulation experiments have been resumed recently providing new, highly valuable data. Work on accident performance will be - similar to the normal operation benchmark - consisting of three essential parts comprising both code verification that establishes the correspondence of code work with the underlying physical, chemical and mathematical laws, and code validation that establishes reasonable agreement with the existing experimental data base, but including also predictive calculations for future heating tests and/or reactor concepts. The paper will describe the cases to be studied and the calculational results obtained with the German computer model FRESCO. Among the benchmark cases in consideration are tests which were most recently conducted in the new heating facility KUEFA. Therefore this study will also re-open the discussion and analysis of both the validity of diffusion models and the transport data of the principal fission product species in the HTGR fuel materials as essential input data for the codes. (authors)

  17. MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

    International Nuclear Information System (INIS)

    Tautges, T.J.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data

  18. Accident considerations in LMFBR design

    International Nuclear Information System (INIS)

    Simpson, D.E.; Alter, H.; Fauske, H.K.; Hikido, K.; Keaten, R.W.; Stevenson, M.G.; Strawbridge, L.

    1975-12-01

    LMFBR safety design criteria are discussed from the standpoints of accident severity classification and damage criteria, and the following design events are considered: fuel failure propagation, reactivity addition faults, heat transport system events, steam generator faults, sodium spills, fuel handling and storage faults, and external events

  19. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-15

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and err000.

  20. A Deformation Analysis Code of CANDU Fuel under the Postulated Accident: ELOCA

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Jung, Jong Yeob

    2006-11-01

    Deformations of the fuel element or fuel channel might be the main cause of the fuel failure. Therefore, the accurate prediction of the deformation and the analysis capabilities are closely related to the increase of the safety margin of the reactor. In this report, among the performance analysis or the transient behavior prediction computer codes, the analysis codes for deformation such as the ELOCA, HOTSPOT, CONTACT-1, and PTDFORM are briefly introduced and each code's objectives, applicability, and relations are explained. Especially, the user manual for ELOCA code which is the analysis code for the fuel deformation and the release of fission product during the transient period after the postulated accidents is provided so that it can be the guidance to the potential users of the code and save the time and economic loss by reducing the trial and error

  1. Fuel pin behaviour under conditions of control rod withdrawal accident in CABRI-2 experiments

    International Nuclear Information System (INIS)

    Papin, Joelle; Lemoine, Francette; Sato, Ikken; Struwe, Dankward; Pfrang, Werner

    1994-01-01

    Simulation of the control rod withdrawal accident has been performed in the international CABRI-2 experimental programme. The tests realized with industrial pins led to clarification of the influence of the pellet design and have shown the important role of fission products on the solid fuel swelling which promotes early pin failure with solid fuel pellet. With annular pellet design, large fuel swelling combined to low smear density leads to degradation of fuel thermal conductivity and thus reduces power to melt. However, the high margin to deterministic failure is confirmed with hollow pellets. Improvements of the modelling were necessary to describe such behaviours in computer codes as SAS-4A, PAPAS-2S and PHYSURAC. (author)

  2. Accidents and troubles in nuclear fuel facilities in fiscal year 1987

    International Nuclear Information System (INIS)

    1988-01-01

    The number of the accidents and troubles reported in fiscal year 1987 in relation to nuclear fuel facilities based on the stipulation of the law on the regulation of nuclear raw materials, nuclear fuel materials and nuclear reactors was two. In Tokai Works, Power Reactor and Nuclear Fuel Development Corp., on September 17, 1987, the conveyor for transporting spent fuel in the separation and refining shop of the reprocessing plant broke down, consequently, the operation of the reprocessing plant was stopped for about five months. In Tokai Testing Works, Mitsubishi Heavy Industries Ltd., on February 7, 1988, a worker who was putting up posters in the control area of the uranium experiment facilities fell from a stepladder, and required treatment by entering a hospital for about one month, suffering bone fracture. (K.I.)

  3. Accidents, troubles and others in nuclear fuel facilities in fiscal year 1988

    International Nuclear Information System (INIS)

    1990-01-01

    The number of the accidents, troubles and others reported on the basis of the 'Law concerning the regulation of nuclear raw material substances, nuclear fuel substances and nuclear reactors' in fiscal year 1988 was one. On February 23, 1989, in the controlled area of the plutonium waste treatment development facilities in Tokai Works. Power Reactor and Nuclear Fuel Development Corp., when one worker entered from a corridor into the material store, he fell down by mistake and broke the left collarbone, which required the hospitalization for about one month. (K.I.)

  4. Conservatism in effective dose calculations for accident events involving fuel reprocessing waste tanks.

    Science.gov (United States)

    Bevelacqua, J J

    2011-07-01

    Conservatism in the calculation of the effective dose following an airborne release from an accident involving a fuel reprocessing waste tank is examined. Within the regulatory constraints at the Hanford Site, deterministic effective dose calculations are conservative by at least an order of magnitude. Deterministic calculations should be used with caution in reaching decisions associated with required safety systems and mitigation philosophy related to the accidental release of airborne radioactive material to the environment.

  5. Criticality safety assessment of a TRIGA reactor spent-fuel pool under accident conditions

    International Nuclear Information System (INIS)

    Glumac, B.; Ravnik, M.; Logar, M.

    1997-01-01

    Additional criticality safety analysis of a pool-type storage for TRIGA spent fuel at the Jozef Stefan Institute in Ljubljana, Slovenia, is presented. Previous results have shown that subcriticality is not guaranteed for some postulated accidents (earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch) under the assumption that the fuel rack is loaded with fresh 12 wt% standard fuel. To mitigate this deficiency, a study was done on replacing a certain number of fuel elements in the rack with cadmium-loaded absorber rods. The Monte Carlo computer code MCNP4A with an ENDF/B-V library and detailed three-dimensional geometrical model of the spent-fuel rack was used for this purpose. First, a minimum critical number of fuel elements was determined for contact pitch, and two possible geometries of rack disintegration were considered. Next, it was shown that subcriticality can be ensured when pitch is decreased from a rack design pitch of 8 cm to contact, if a certain number of fuel elements (8 to 20 out of 70) are replaced by absorber rods, which are uniformly mixed into the lattice. To account for the possibility that random mixing of fuel elements and absorber rods can occur during rack disintegration and result in a supercritical configuration, a probabilistic study was made to sample the probability density functions for random absorber rod lattice loadings. Results of the calculations show that reasonably low probabilities for supercriticality can be achieved (down to 10 -6 per severe earthquake, which would result in rack disintegration and subsequent maximum possible pitch decrease) even in the case where fresh 12 wt% standard TRIGA fuel would be stored in the spent-fuel pool

  6. In-pile observations of fuel and clad relocation during LMFBR initiation phase accident experiments - the STAR experiments

    International Nuclear Information System (INIS)

    Wright, S.A.; Schumacher, G.; Henkel, P.R.; Royl, P.

    1987-01-01

    A series of seven in-pile experiments (the STAR experiments) were performed in which clad motion and fuel dispersal were observed in small pin bundles with high-speed cinematography. The experimental heating conditions reproduced a range of Loss of Flow (LOF) accident scenarios for the lead subassemblies in LMFBRs. The experiments show strong tendencies for limited clad motion in multiple pin bundles, early fuel disruption and dispersal (prior to fuel melting) in moderate power transients having simultaneous clad melting and fuel disruption. The more recent experiments indicate a possibility of steel vapor driven fuel dispersal after fuel breakup and intimate fuel/steel mixing. (author)

  7. Identification of the security threshold by logistic regression applied to fuel under accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel de Souza; Baptista Filho, Benedito; Oliveira, Fabio Branco de, E-mail: dsgomes@ipen.br, E-mail: bdbfilho@ipen.br, E-mail: fabio@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco

    2015-07-01

    A reactivity-initiated Accident (RIA) is a disastrous failure, which occurs because of an unexpected rise in the fission rate and reactor power. This sudden increase in the reactor power may activate processes that might lead to the failure of fuel cladding. In severe accidents, a disruption of fuel and core melting can occur. The purpose of the present research is to study the patterns of such accidents using exploratory data analysis techniques. A study based on applied statistics was used for simulations. Then, we chose peak enthalpy, pulse width, burnup, fission gas release, and the oxidation of zirconium as input parameters and set the safety boundary conditions. This new approach includes the logistic regression. With this, the present research aims also to develop the ability to identify the conditions and the probability of failures. Zirconium-based alloys fabricating the cladding of the fuel rod elements with niobium 1% were analyzed for high burnup limits at 65 MWd/kgU. The data based on six decades of investigations from experimental programs. In test, perform in American reactors such as the transient reactor test (TREAT), and power Burst Facility (PBF). In experiments realized in Japanese program at nuclear in the safety research reactor (NSRR), and in Kazakhstan as impulse graphite reactor (IGR). The database obtained from the tests and served as a support for our study. (author)

  8. Identification of the security threshold by logistic regression applied to fuel under accident conditions

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza; Baptista Filho, Benedito; Oliveira, Fabio Branco de; Giovedi, Claudia

    2015-01-01

    A reactivity-initiated Accident (RIA) is a disastrous failure, which occurs because of an unexpected rise in the fission rate and reactor power. This sudden increase in the reactor power may activate processes that might lead to the failure of fuel cladding. In severe accidents, a disruption of fuel and core melting can occur. The purpose of the present research is to study the patterns of such accidents using exploratory data analysis techniques. A study based on applied statistics was used for simulations. Then, we chose peak enthalpy, pulse width, burnup, fission gas release, and the oxidation of zirconium as input parameters and set the safety boundary conditions. This new approach includes the logistic regression. With this, the present research aims also to develop the ability to identify the conditions and the probability of failures. Zirconium-based alloys fabricating the cladding of the fuel rod elements with niobium 1% were analyzed for high burnup limits at 65 MWd/kgU. The data based on six decades of investigations from experimental programs. In test, perform in American reactors such as the transient reactor test (TREAT), and power Burst Facility (PBF). In experiments realized in Japanese program at nuclear in the safety research reactor (NSRR), and in Kazakhstan as impulse graphite reactor (IGR). The database obtained from the tests and served as a support for our study. (author)

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

    International Nuclear Information System (INIS)

    Lee, Chan Bok; Lee, Chung Chan; Kim, Oh Hwan; Kim, Jong Jin

    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)

  10. Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

    International Nuclear Information System (INIS)

    Zhang Zhuohua; Yu Junchong; Peng Shinian

    2014-01-01

    According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

  11. Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions

    International Nuclear Information System (INIS)

    Pettersson, Kjell; Chung, Haijung; ); Billone, Michael; Fuketa, Toyoshi; Nagase, Fumihisa; Grandjean, Claude; Hache, George; Papin, Joelle; Heins, Lothar; Hozer, Zoltan; In de Betou, Jan; Kelppe, Seppo; Mayer, Ralph; Scott, Harold; Voglewede, John; Sonnenburg, Heinz; Sunder, Sham; Valach, Mojmir; Vrtilkova, Vera; Waeckel, Nicolas; Wiesenack, Wolfgang; Zimmermann, Martin

    2009-01-01

    The NEA Working Group on Fuel Safety (WGFS) is tasked with advancing the current understanding of fuel safety issues by assessing the technical basis for current safety criteria and their applicability to high burn-up and to new fuel designs and materials. The group aims at facilitating international convergence in this area, including as regards experimental approaches and interpretation and the use of experimental data relevant for safety. In 1986, a working group of the NEA Committee on the Safety of Nuclear Installations (CSNI) issued a state-of-the-art report on water reactor fuel behaviour in design-basis accident (DBA) conditions. The 1986 report was limited to the oxidation, embrittlement and deformation of pressurised water reactor (PWR) fuel in a loss-of-coolant accident (LOCA). Since then, considerable experimental and analytical work has been performed, which has led to a broader and deeper understanding of LOCA-related phenomena. Further, new cladding alloys have been produced, which might behave differently than the previously used Zircaloy-4, both under normal operating conditions and during transients. Compared with 20 years ago, fuel burn-up has been significantly increased, which requires extending the LOCA database in order to cover the high burnup range. There was also a clear need to address LOCA performance for reactor types other than PWRs. The present report has been prepared by the WGFS and covers the following technical aspects: - Description of different LOCA scenarios for major types of reactors: BWRs, PWRs, VVERs and to a lesser extent CANDUs. - LOCA phenomena: ballooning, burst, oxidation, fuel relocation and possible fracture at quench. - Details of high-temperature oxidation behaviour of various cladding materials. - Metallurgical phase change, effect of hydrogen and oxygen on residual cladding ductility. - Methods for LOCA testing, for example two-sided oxidation and ring compression for ductility, and integral quench test for

  12. Scoping studies of vapor behavior during a severe accident in a metal-fueled reactor

    International Nuclear Information System (INIS)

    Spencer, B.W.; Marchaterre, J.F.

    1985-01-01

    Scoping calculations have been performed examining the consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel. The principal gas and vapor species released are shown to be Xe, Cs,and bond sodium contained within the fuel porosity. Fuel vapor pressure is insignificant, and there is no energetic fuel-coolant interaction for the conditions considered. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core (although reactor-material experiments are needed to confirm these high condensation rates). If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the implication is that the ability of vapor expansion to perform appreciable work on the system is largely eliminated. Furthermore, the ability of an expanding vapor bubble to transport fuel and fission product species to the cover gas region where they may be released to the containment is also largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool

  13. Predicted HIFAR fuel element temperatures for postulated loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Green, W.J.

    1987-04-01

    A two-dimensional theoretical heat transfer model of a HIFAR Mark IV/Va fuel element has been developed and validated by comparing predicted thermal performances with experimental temperature responses obtained from irradiated fuel elements during simulated accident conditions. Full details of the model's development and its verification have been reported elsewhere. In this report, the model has been further used to ascertain acceptable limits of fuel element decay power for the start of two specific LOCAs which have been identified by the Regulatory Bureau of the AAEC. For a single fuel element which is positioned within a fuel load/unload flask and is not subjected to any forced convective air cooling, the model indicates that fission product decay powers must not exceed 1.86 kW if fuel surface temperatures are not to exceed 450 deg C. In the case of a HIFAR core LOCA in which the complete inventory of heavy water is lost, it is calculated that the maximum fission product decay power of a central element must not exceed 1.1 kW if fuel surface temperatures are not to exceed 450 deg C anywhere in the core

  14. The WWER fuel element safety research under the design and heavy accident imitation on the 'PARAMETR' stand

    International Nuclear Information System (INIS)

    Deniskin, V.P.; Nalivaev, V.I.; Parshin, N. Ya.; Fedik, I.I.

    2000-01-01

    Analysis of fuel element behavior in the course of the design and heavy accidents is the component of reactor facility safety prevention. Many tasks of fuel element behavior research may be solved with the help of thermophysical stands. One of such stands implemented in 1991 was thermophysical stand 'PARAMETER'.Several experiments on model assemblies chiefly imitating both heavy accident and design basic accident have already been conducted in 'PARAMETER' stand. There were obtained data about fuel claddings seal failure and deformation condition. In particular it was defined that seal failure of all fuel claddings occurs on stage of fuel element warming, in temperature range (770-900) degree celsius and almost does not depend on inner pressure level

  15. Effects of spent fuel types on offsite consequences of hypothetical accidents

    International Nuclear Information System (INIS)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-01-01

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor (χ/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only 134 Cs, 137 Cs- 137m Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents

  16. Proceedings of the Second OECD/NEA Organisation Meeting on Increased Accident Tolerance of Fuels for LWRs

    International Nuclear Information System (INIS)

    Massara, Simone; ); Bragg-Sitton, Shannon; Braase, Lori; Merrill, Brad; Teague, Melissa; Stanek, Chris R.; Montgomery, Robert H.; Ott, Larry J.; Robb, Kevin; Snead, Lance; Farmer, Mitch; Billone, Michael C.; Todosow, Michael; Brown, Nicholas; Brachet, J.C.; Le Flem, M.; Sauder, C.; Idarraga-Trujillo, I.; Michaux, A.; Lorrette, C.; Le Saux, M.; Blanpain, P.; Park, Jeong-Yong; Yang, Jae-Ho; Kim, Weon-Ju; Koo, Yang-Hyun; Liu, T.; Hallstadius, Lars; Lahoda, Ed; Waeckel, N.; Bonnet, J.M.; Vitanza, Carlo; Ohta, Hirokazu; Ogata, Takanari; Nakamura, Kinya; Dyck, Gary; Inozemtsev, Victor; )

    2013-01-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the 2. Meeting on Increased Accident Tolerance of Fuels for LWRs. Content: 1 - Overview of the exchanges after the December-2012 Workshop through the discussion forum established at the OECD-NEA (S. Massara, NEA); 2 - Metrics Development for Enhanced Accident Tolerant LWR Fuels (S. Bragg-Sitton, INL); 3 - Candidate ATF Clad Technologies and Key Feasibility Issues (L. Snead, ORNL); 4 - CEA studies on nuclear fuel claddings for LWRs enhanced accident tolerant fuel. Some recent results, pending issues and prospects (J.C. Brachet, CEA); 5 - Current status on the accident tolerant fuel development in the Republic of Korea (J.Y. Park, J.H. Chang, KAERI); 6 - The current status of fuel R and D in the P.R. of China (T. Liu, CGN). Session 2: Key elements for a work programme on ATF: 7 - Beneficial characteristics of ATF (metrics) (L. Hallstadius, Westinghouse); 8 - Reactor types of interest (applicability) (L. Ott, ORNL); 9 - Impact on normal operations

  17. Analysis of the loss of pool cooling accident in a PWR spent fuel pool with MAAP5

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2014-01-01

    Highlights: • A PWR spent fuel pool was modeled by using MAAP5. • Loss of pool cooling severe accident scenarios were studied. • Loss of pool cooling accidents with two mitigation measures were analyzed. - Abstract: The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents the analysis of loss of pool cooling accident scenarios and the discussion of mitigation measures for the SFP at a pressurized water reactor (PWR) NPP with the MAAP5 code. Analysis of uncompensated loss of water due to the loss of pool cooling with different initial pool water levels of 12.2 m (designated as a reference case) and 10.7 m have been performed based on a MAAP5 input model. Scenarios of the accident such as overheating of uncovered fuel assemblies, oxidation of claddings and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission products were predicted with the assumption that mitigation measures were unavailable. The results covered a broad spectrum of severe accident evaluations in the SFP. Furthermore, as important mitigation measures, the effects of recovering the SFP cooling system and makeup water in SFP on the accident progressions have also been investigated respectively based on the events of pool water boiling and spent fuels uncovery. Based upon the reference case, three cases with the recovery of SFP cooling system and three other cases with makeup water in SFP have been studied. The results showed that, severe accident might happen if SFP cooling system was not restored timely before the spent fuels started to become uncovered; spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate larger than the average evaporation rate defined as the division of pool water mass above the

  18. Assessment of fuel damage of pool type research reactor in the case of fuel plates blockage

    Energy Technology Data Exchange (ETDEWEB)

    Jalil, Jafari; Samad, Khakshournia [AEOI, Karegar Ave. School of R and D of Nuclear Reactors and Accelerators, Teheran (Iran, Islamic Republic of); D' Auria, F. [Pisa Univ., DIMNP (Italy)

    2007-07-01

    Tehran Research Reactor (TRR) is a pool type 5 MW research reactor. It is assumed that external objects or debris that may fall down to reactor core cause obstruction of coolant flow through one of the fuel assemblies. Thermal hydraulic analysis of this event, using the RELAP5 system code has been studied. The reported transient is related to the partial and total obstruction of a single Fuel Element (FE) cooling channel of 27 FE equilibrium core of TRR. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FE integrity. Two scenarios are analysed to emphasize the severity of the accident. The first one is a partial blockage of an average FE considering four different obstruction levels: 25%, 50%, 75% and 97% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FE. This study constitutes the first step of a larger work which consists of performing a 3-dimensional simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation included in the RELAP5 code. Main results obtained from the RELAP5 calculations are as following. First, in the case of flow blockage under 97% of the nominal flow area of an average FE, only an increase of the coolant and clad temperatures is observed without any consequences for the integrity of the FE. The mass flow rate remains sufficient to cool the clad safely. Secondly, in the case of total obstruction of the nominal flow area, it is seen that transient turns out to be a severe accident due to the dryout conditions are reached shortly and melting of the cladding occurs. Thirdly, the use of the point kinetic approach leads to conservative results. A best estimate simulation of such kind of transients requires the use of 3-dimensional kinetic calculations, which could be done using the current Coupled Codes

  19. High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

  20. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    International Nuclear Information System (INIS)

    Rebak, Raul B.

    2014-01-01

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  1. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  2. Status Report on Spent Fuel Pools under Loss-of-Cooling and Loss-of-Coolant Accident Conditions - Final Report

    International Nuclear Information System (INIS)

    Adorni, M.; Esmaili, H.; Grant, W.; Hollands, T.; Hozer, Z.; Jaeckel, B.; Munoz, M.; Nakajima, T.; Rocchi, F.; Strucic, M.; ); Tregoures, N.; Vokac, P.; Ahn, K.I.; Bourgue, L.; Dickson, R.; Douxchamps, P.A.; Herranz, L.E.; Jernkvist, L.O.; Amri, A.; Kissane, M.P.; )

    2015-01-01

    Following the 2011 accident at the Fukushima Daiichi Nuclear Power Station, the Nuclear Energy Agency Committee on the Safety of Nuclear Installations decided to launch several high-priority activities to address certain technical issues. Among other things, it was decided to prepare a status report on spent fuel pools (SFPs) under loss of cooling accident conditions. This activity was proposed jointly by the CSNI Working Group on Analysis and Management of Accidents (WGAMA) and the Working Group on Fuel Safety (WGFS). The main objectives, as defined by these working groups, were to: - Produce a brief summary of the status of SFP accident and mitigation strategies, to better contribute to the post-Fukushima accident decision making process; - Provide a brief assessment of current experimental and analytical knowledge about loss of cooling accidents in SFPs and their associated mitigation strategies; - Briefly describe the strengths and weaknesses of analytical methods used in codes to predict SFP accident evolution and assess the efficiency of different cooling mechanisms for mitigation of such accidents; - Identify and list additional research activities required to address gaps in the understanding of relevant phenomenological processes, to identify where analytical tool deficiencies exist, and to reduce the uncertainties in this understanding. The proposed activity was agreed and approved by CSNI in December 2012, and the first of four meetings of the appointed writing group was held in March 2013. The writing group consisted of members of the WGAMA and the WGFS, representing the European Commission and the following countries: Belgium, Canada, Czech Republic, France, Germany, Hungary, Italy, Japan, Korea, Spain, Sweden, Switzerland and the USA. This report mostly covers the information provided by these countries. The report is organised into 8 Chapters and 4 Appendices: Chapter 1: Introduction; Chapter 2: Spent fuel pools; Chapter 3: Possible accident

  3. FeCrAl/Zr dual layer fuel cladding for improved safety margin under accident scenario

    International Nuclear Information System (INIS)

    Park, D.J.; Park, J.H.; Jung, Y.I.; Kim, H.G.; Park, J.Y.; Koo, Y.H.

    2014-01-01

    For application of advanced steel as a cladding material in light water reactor (LWR), FeCrAl/Zr dual layer tube was manufactured by using a hot isostatic pressing (HIP) method. To optimize HIP condition for joining both FeCrAl and Zr alloys, HIP was carried out under various temperature conditions. Tensile test and 3-point bend test performed for measuring mechanical properties of HIPed sample. To better understand microstructural characteristics in interface region between two alloys, SEM and TEM study were conducted by using HIPed sample with different process conditions. Based on this optimization study and analyzed results, optimized HIP condition was determined and FeCrAl/Zr dual layer fuel cladding having same wall thickness with current LWR fuel cladding was manufactured. Simulated loss-of-coolant accident test was carried out using FeCrAl/Zr dual layer cladding sample and fuel integrity was measured by mechanical test. (authors)

  4. Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, D.J.; Brehm, J.R.

    1994-01-01

    The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable.

  5. Postulated accident scenarios for the on-site transport of spent nuclear fuel

    International Nuclear Information System (INIS)

    Morandin, G.; Sauve, R.

    2004-01-01

    Once a spent fuel container is loaded with spent fuel it typically travels on-site to a processing building for permanent lid attachment. During on-site transport a lid clamp is utilized to ensure the container lid remains in place. The safe on-site transport of spent nuclear fuel must rely on the structural integrity of the transport container and system of transport. Regard for on-site traffic and safe, efficient travel routes are important and manageable with well thought-out planning. Non-manageable incidences, such as flying debris from tornado force winds or postulated blasts in proximity to the transport container, that may result in high velocity impact and shock loading on the transport system must be considered. This paper consists of simulations that consider these types of postulated accident scenarios using detailed nonlinear finite element techniques

  6. Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility

    International Nuclear Information System (INIS)

    Johnson, D.J.; Brehm, J.R.

    1994-01-01

    The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable

  7. Fuel containment and damage tolerance for large composite primary aircraft structures. Phase 1: Testing

    Science.gov (United States)

    Sandifer, J. P.

    1983-01-01

    Technical problems associated with fuel containment and damage tolerance of composite material wings for transport aircraft were identified. The major tasks are the following: (1) the preliminary design of damage tolerant wing surface using composite materials; (2) the evaluation of fuel sealing and lightning protection methods for a composite material wing; and (3) an experimental investigation of the damage tolerant characteristics of toughened resin graphite/epoxy materials. The test results, the test techniques, and the test data are presented.

  8. Coupled fuel performance and thermal-hydraulics simulation with BISON and RELAP-7 at accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Martineau, R.C., E-mail: Richard.Martineau@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' RELAP-7 is expected to be the next in the RELAP nuclear reactor safety/systems analysis application series developed at the Idaho National Laboratory (INL). The development of RELAP-7 began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of Department of Energy's (DOE) Light Water Reactor Sustainability (LWRS) Program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support modern nuclear power safety analysis. RELAP-7 is built using the INL's modern scientific software development framework, MOOSE (Multi-physics Object Oriented Simulation Environment). MOOSE provides improved implicit numerical schemes, including higher-order integration in both space and time, and yielding converged second-order accuracy for RELAP-7. The code structure is based on multiple physical component models such as pipes, junctions, pumps, etc. This component-based software architecture allows RELAP-7 to quickly adopt different physical models for different applications. One of the main advantages of building RELAP-7 on the MOOSE framework is that tight coupling with other MOOSE-based applications solving physics not present in RELAP-7 requires little to no additional lines of code. For example, the RELAP-7 core channel component is based upon a one-dimensional flow channel and a three-zone two-dimensional heat structure designed to represent fuel, gap, and cladding conjugate heat transfer with the coolant. However, the RELAP-7 application does not carry the fuels performance physics to analyze irradiated fuel, especially for accident scenarios. Here, we demonstrate the tightly coupled capability of the BISON nuclear fuels performance application with RELAP-7 for the station black out (SBO) accident scenario (Fukushima type event) and

  9. Categorization of failed and damaged spent LWR [light-water reactor] fuel currently in storage

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1987-11-01

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

  10. Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding

    Directory of Open Access Journals (Sweden)

    Bo Cheng

    2016-02-01

    Full Text Available In severe loss of coolant accidents (LOCA, similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are

  11. Comparison of US/FRG accident condition models for HTGR fuel failure and radionuclide release

    International Nuclear Information System (INIS)

    Verfondern, K.

    1991-03-01

    The objective was to compare calculation models used in safety analyses in the US and FRG which describe fission product release behavior from TRISO coated fuel particles under core heatup accident conditions. The frist step performed is the qualitative comparison of both sides' fuel failure and release models in order to identify differences and similarities in modeling assumptions and inputs. Assumptions of possible particle failure mechanisms under accident conditions (SiC degradation, pressure vessel) are principally the same on both sides though they are used in different modeling approaches. The characterization of a standard (= intact) coated particle to be of non-releasing (GA) or possibly releasing (KFA/ISF) type is one of the major qualitative differences. Similar models are used regarding radionuclide release from exposed particle kernels. In a second step, a quantitative comparison of the calculation models was made by assessing a benchmark problem predicting particle failure and radionuclide release under MHTGR conduction cooldown accident conditions. Calculations with each side's reference method have come to almost the same failure fractions after 250 hours for the core region with maximum core heatup temperature despite the different modeling approaches of SORS and PANAMA-I. The comparison of the results of particle failure obtained with the Integrated Failure and Release Model for Standard Particles and its revision provides a 'verification' of these models in this sense that the codes (SORS and PANAMA-II, and -III, respectively) which were independently developed lead to very good agreement in the predictions. (orig./HP) [de

  12. Criticality accident in uranium fuel processing plant. Questionnaires from Research Committee of Nuclear Safety

    International Nuclear Information System (INIS)

    Kataoka, Isao; Sekimoto, Hiroshi

    2000-01-01

    The Research Committee of Nuclear Safety carried out a research on criticality accident at the JCO plant according to statement of president of the Japan Atomic Energy Society on October 8, 1999, of which results are planned to be summarized by the constitutions shown as follows, for a report on the 'Questionnaires of criticality accident in the Uranium Fuel Processing Plant of the JCO, Inc.': general criticality safety, fuel cycle and the JCO, Inc.; elucidation on progress and fact of accident; cause analysis and problem picking-up; proposals on improvement; and duty of the Society. Among them, on last two items, because of a conclusion to be required for members of the Society at discussions of the Committee, some questionnaires were send to more than 1800 of them on April 5, 2000 with name of chairman of the Committee. As results of the questionnaires contained proposals and opinions on a great numbers of fields, some key-words like words were found on a shape of repeating in most questionnaires. As they were thought to be very important nuclei in these two items, they were further largely classified to use for summarizing proposals and opinions on the questionnaires. This questionnaire had a big characteristic on the duty of the Society in comparison with those in the other organizations. (G.K.)

  13. Review of design criteria for Criticality Accident Alarm System (CAAS) used in Fuel Reprocessing Facility

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Basu, Pew; Sivasubramaniyan, K.; Venkatraman, B.

    2016-01-01

    Though fuel cycle facilities handling fissile materials are designed with careful criticality safety analysis, the criticality accident cannot be ruled out completely. Criticality Accident Alarm System (CAAS) is being installed as part of criticality safety management in fuel cycle facilities. CAAS system being used in India, is ECIL make, ionization chamber based gamma detector, which houses three identical detectors and works on 2/3 logic. As per ISO 7753 and ANSI/ANS-8.3, the CAAS must be designed to be capable of detecting any minimum accident occurs which could be of concern. Based on this, alarm limit used in CAAS is: 4 R/h (fast transient excursion) and 3 mR in 0.5 sec (slow excursion). In case of reprocessing facilities wherein process tanks located in heavy shielding, identification of CAAS installation locations require detailed radiation transport calculations. A study has been taken to estimate the gamma dose rate from thick concrete hot cells in order to determine the locations of CAAS to meet the present design criteria of alarm limit

  14. The coupling algorithm between fuel pin and coolant channel in the European Accident Code EAC-2

    International Nuclear Information System (INIS)

    Goethem, G. van; Lassmann, K.

    1989-01-01

    In the field of fast breeder reactors the Commission of the European Communities (CEC) is conducting coordination and harmonisation activities as well as its own research at the CEC's Joint Research Centre (JRC). The development of the modular European Accident Code (EAC) is a typical example of concerted action between EC Member States performed under the leadership of the JRC. This computer code analyzes the initiation phase of low-probability whole-core accidents in LMFBRs with the aim of predicting the rapidity of sodium voiding, the mode of pin failure, the subsequent fuel redistribution and the associated energy release. This paper gives a short overview on the development of the EAC-2 code with emphasis on the coupling mechanism between the fuel behaviour module TRANSURANUS and the thermohydraulics modules which can be either CFEM or BLOW3A. These modules are also briefly described. In conclusion some numerical results of EAC-2 are given: they are recalculations of an unprotected LOF accident for the fictitious EUROPE fast breeder reactor which was earlier analysed in the frame of a comparative exercise performed in the early 80s and organised by the CEC. (orig.)

  15. Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Suman, Siddharth [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Khan, Mohd Kaleem, E-mail: mkkhan@iitp.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Pathak, Manabendra [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Singh, R.N.; Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2016-10-15

    Highlights: • Modelling of nuclear fuel cladding during loss-of-coolant accident transient. • Phase transformation, corrosion, and creep combined to evaluate burst criterion. • Effect of oxygen concentration on burst stress and burst strain. • Effect of heating rate, internal pressure fluctuation, shear modulus incorporated. - Abstract: A burst criterion model accounting the simultaneous phenomena of corrosion, solute-strengthening effect of oxygen, oxygen concentration based non-isothermal phase transformation, and thermal creep has been developed to predict the rupture behaviour of zircaloy-4 nuclear fuel cladding during the loss-of-coolant accident transients. The present burst criterion model has been validated using experimental data obtained from single-rod transient burst tests performed in steam environment. The predictions are in good agreement with the experimental results. A detailed computational analysis has been performed to assess the role of different parameters in the rupture of zircaloy cladding during loss-of-coolant accidents. This model reveals that at low temperatures, lower heating rates produce higher burst strains as oxidation effect is nominal. For high temperatures, the lower heating rates produce less burst strains, whereas higher heating rates yield greater burst strains.

  16. Study of a criticality accident involving fuel rods and water outside a power reactor

    International Nuclear Information System (INIS)

    Beloeil, L.

    2000-01-01

    It is possible to imagine highly unlikely but numerous accidental situations where fuel rods come into contact with water under conditions close to atmospheric values. This work is devoted to modelling and simulation of first instants of the power excursion that may result from such configurations. We show that void effect is a preponderant feedback for most severe accidents. The formation of a vapour film around the rods is put forward and confirmed with the help of experimental transients using electrical heating. We propose then a vapour/liquid flow model able to reproduce void fraction evolution. The vapour film is treated as a compressible medium. Conservation balance equations are solved on a moving mesh with a two-dimensional scheme and boundary conditions taking notice of interfacial phenomena and axial escape possibility. Movements of the liquid phase are modelled through a non-stationary integral equation and a dissipative term suited to the particular geometry of this flow. The penetration of energy into the liquid is also calculated. Thus, the coupling of aerodynamic and hydrodynamic modules gives results in excellent agreement with experiments. Next, neutronic phenomena into the fuel pellet, their feedback effects and the distribution of power through the rod are numerically translated. For each developed module, validation tests are provided. Then, it is possible to simulate the first seconds of the whole criticality accident. Even if this calculation tool is only a way of study as a first approach, performed simulations are proving coherent with reported data on recorded accidents. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. Fission products and nuclear fuel behaviour under severe accident conditions part 2: Fuel behaviour in the VERDON-1 sample

    Science.gov (United States)

    Geiger, E.; Le Gall, C.; Gallais-During, A.; Pontillon, Y.; Lamontagne, J.; Hanus, E.; Ducros, G.

    2017-11-01

    Within the framework of the International Source Term Programme (ISTP), the VERDON programme aims at quantifying the source term of radioactive materials in case of a hypothetical severe accident in a light water reactor (LWR). Tests were performed in a new experimental laboratory (VERDON) built in the LECA-STAR facility (CEA Cadarache). The VERDON-1 test was devoted to the study of a high burn-up UO2 fuel and FP releases at very high temperature (≈2873 K) in a reducing atmosphere. Post-test qualitative and quantitative characterisations of the VERDON-1 sample led to the proposal of a scenario explaining the phenomena occurring during the experimental sequence. Hence, the fuel and the cladding may have interacted which led to the melting of UO2-ZrO2 alloy. Although no relocation was observed during the test, it may have been imminent.

  19. Phase 1A Final Report for the AREVA Team Enhanced Accident Tolerant Fuels Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Morrell, Mike E. [AREVA Federal Services LLC, Charlotte, NC (United States)

    2015-03-19

    In response to the Department of Energy (DOE) funded initiative to develop and deploy lead fuel assemblies (LFAs) of Enhanced Accident Tolerant Fuel (EATF) into a US reactor within 10 years, AREVA put together a team to develop promising technologies for improved fuel performance during off normal operations. This team consisted of the University of Florida (UF) and the University of Wisconsin (UW), Savannah River National Laboratory (SRNL), Duke Energy and Tennessee Valley Authority (TVA). This team brought broad experience and expertise to bear on EATF development. AREVA has been designing; manufacturing and testing nuclear fuel for over 50 years and is one of the 3 large international companies supplying fuel to the nuclear industry. The university and National Laboratory team members brought expertise in nuclear fuel concepts and materials development. Duke and TVA brought practical utility operating experience. This report documents the results from the initial “discovery phase” where the team explored options for EATF concepts that provide enhanced accident tolerance for both Design Basis (DB) and Beyond Design Basis Events (BDB). The main driver for the concepts under development were that they could be implemented in a 10 year time frame and be economically viable and acceptable to the nuclear fuel marketplace. The economics of fuel design make this DOE funded project very important to the nuclear industry. Even incremental changes to an existing fuel design can cost in the range of $100M to implement through to LFAs. If this money is invested evenly over 10 years then it can take the fuel vendor several decades after the start of the project to recover their initial investment and reach a breakeven point on the initial investment. Step or radical changes to a fuel assembly design can cost upwards of $500M and will take even longer for the fuel vendor to recover their investment. With the projected lifetimes of the current generation of nuclear power

  20. A study on gap heat transfer of LWR fuel rods under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Fujishiro, Toshio

    1984-03-01

    Gap heat transfer between fuel pellet and cladding have a large influence on the LWR fuel behaviors under reactivity initiated accident (RIA) conditions. The objective of the present study is to investigate the effects of gap heat transfer on RIA fuel behaviors based on the results of the gap-gas parameter tests in NSRR and on their analysis with NSR-77 code. Through this study, transient variations of gap heat transfer, the effects of the gap heat transfer on fuel thermal behaviors and on fuel failure, effects of pellet-cladding sticking by eutectic formation, and the effects of cladding collapse under high external pressure have been clearified. The studies have also been performed on the applicability and its limit of modified Ross and Stoute equation which is extensively utilized to evaluate the gap heat transfer coefficient in the present fuel behavior codes. The method to evaluate the gap conductance to the conditions beyond the applicability limit of the Ross and Stoute equation has also been proposed. (author)

  1. Simulation of accident-tolerant U3Si2 fuel using FRAPCON code

    International Nuclear Information System (INIS)

    Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia

    2017-01-01

    The research on accident-tolerant fuels (ATFs) increased after the Fukushima event. This benefited risk management in nuclear operations. In this investigation, the physical properties of the materials being developed for the ATF program were compared with those of the standard UO 2 - Zr fuel system. The research efforts in innovative fuel design include rigorous characterization of thermal, mechanical, and chemical assessment, with the objectives of making the burnup cycle longer, increasing power density, and improving safety performance. Fuels must reach a high uranium density - above that supported by UO 2 - and possess coating that exhibits better oxidation resistance than Zircaloy. The uranium density and thermal conductivity of ATFs, such as U 3 Si 2 , UN, and UC, is higher than that of UO 2 ; their combination with advanced cladding provides possible fuel - cladding options. An ideal combination of fuel and cladding must increase fuel performance in loss-of-coolant scenarios. The disadvantages of U 3 Si 2 , UN, and UC are their swelling rates, which are higher than that of UO 2 . The thermal conductivities of ATFs are approximately four times higher than that of UO2. To prevent the generation of hydrogen due to oxidation of zirconium-based alloys in contact with steam, cladding options, such as ferritic alloys, were studied. It was verified that FeCrAl alloys and SiC provide better response under severe conditions because of their thermophysical properties. The findings of this study indicate that U 3 Si 2 and the FeCrAl fuel cladding concept should replace UO 2 - Zr as the fuel system of choice. (author)

  2. Large scale experiments simulating hydrogen distribution in a spent fuel pool building during a hypothetical fuel uncovery accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Mignot, Guillaume; Paranjape, Sidharth; Paladino, Domenico; Jaeckel, Bernd; Rydl, Adolf [Paul Scherrer Institute, Villigen (Switzerland)

    2016-08-15

    Following the Fukushima accident and its extended station blackout, attention was brought to the importance of the spent fuel pools' (SFPs) behavior in case of a prolonged loss of the cooling system. Since then, many analytical works have been performed to estimate the timing of hypothetical fuel uncovery for various SFP types. Experimentally, however, little was done to investigate issues related to the formation of a flammable gas mixture, distribution, and stratification in the SFP building itself and to some extent assess the capability for the code to correctly predict it. This paper presents the main outcomes of the Experiments on Spent Fuel Pool (ESFP) project carried out under the auspices of Swissnuclear (Framework 2012–2013) in the PANDA facility at the Paul Scherrer Institut in Switzerland. It consists of an experimental investigation focused on hydrogen concentration build-up into a SFP building during a predefined scaled scenario for different venting positions. Tests follow a two-phase scenario. Initially steam is released to mimic the boiling of the pool followed by a helium/steam mixture release to simulate the deterioration of the oxidizing spent fuel. Results shows that while the SFP building would mainly be inerted by the presence of a high concentration of steam, the volume located below the level of the pool in adjacent rooms would maintain a high air content. The interface of the two-gas mixture presents the highest risk of flammability. Additionally, it was observed that the gas mixture could become stagnant leading locally to high hydrogen concentration while steam condenses. Overall, the experiments provide relevant information for the potentially hazardous gas distribution formed in the SFP building and hints on accident management and on eventual retrofitting measures to be implemented in the SFP building.

  3. Characterization of SiC–SiC composites for accident tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Deck, C.P., E-mail: Christian.Deck@ga.com; Jacobsen, G.M.; Sheeder, J.; Gutierrez, O.; Zhang, J.; Stone, J.; Khalifa, H.E.; Back, C.A.

    2015-11-15

    Silicon carbide (SiC) is being investigated for accident tolerant fuel cladding applications due to its high temperature strength, exceptional stability under irradiation, and reduced oxidation compared to Zircaloy under accident conditions. An engineered cladding design combining monolithic SiC and SiC–SiC composite layers could offer a tough, hermetic structure to provide improved performance and safety, with a failure rate comparable to current Zircaloy cladding. Modeling and design efforts require a thorough understanding of the properties and structure of SiC-based cladding. Furthermore, both fabrication and characterization of long, thin-walled SiC–SiC tubes to meet application requirements are challenging. In this work, mechanical and thermal properties of unirradiated, as-fabricated SiC-based cladding structures were measured, and permeability and dimensional control were assessed. In order to account for the tubular geometry of the cladding designs, development and modification of several characterization methods were required.

  4. Development of the model for the stress calculation of fuel assembly under accident load

    International Nuclear Information System (INIS)

    Kim, Il Kon

    1993-01-01

    The finite element model for the stress calculation in guide thimbles of a fuel assembly (FA) under seismic and loss-of-coolant-accident (LOCA) load is developed. For the stress calculation of FA under accident load, at first the program MAIN is developed to select the worst bending mode shaped FA from core model. And then the model for the stress calculation of FA is developed by means of the finite element code. The calculated results of program MAIN are used as the kinematic constraints of the finite element model of a FA. Compared the calculated results of the stiffness of the finite element model of FA with the test results they have good agreements. (Author)

  5. Calculation of health risks from spent-nuclear-fuel transportation accidents

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1987-01-01

    Models developed to analyze potential radiological health risks from various accident scenarios during transportation of spent nuclear fuels are described. The models are designed both for detailed route-specific risk analyses and for use in conducting overall risk analyses for route selection and related decision-making activities. The radiological risks calculated include individual dose commitments, collective dose commitments, and long-term (100-year) environmental dose commitments to a population following release of radioactivity. To facilitate route-specific analysis, a state-level database was developed and incorporated into the model. Route-specific analysis is demonstrated by the calculation of radiological risks resulting from various accident scenarios, as postulated by the recent US Nuclear Regulatory Commission Modal Study, for four representative states selected from various regions of the United States. 10 refs., 3 figs., 3 tabs

  6. Reassessment of the basis for NRC fuel damage criteria for reactivity transients

    International Nuclear Information System (INIS)

    McCardell, R.K.

    1994-01-01

    The present basis for NRC Fuel Damage Criteria was obtained from experiments performed in the Special Power Excursion Reactor Test (SPERT) IV Reactor Capsule Driver Core (CDC) at the Idaho National Engineering Laboratory (INEL) between 1967 and 1970. Most of the CDC test fuel rods were previously unirradiated and the failure threshold for these unirradiated fuel rods was measured to be about 200 calories per gram of UO 2 radially averaged fuel enthalpy at the axial peak

  7. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  8. Relevance of IAEA tests to severe accidents in nuclear fuel cycle transport

    International Nuclear Information System (INIS)

    Wilkinson, W.L.

    2004-01-01

    The design and performance standards for packages used for the transport of nuclear fuel cycle materials, are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials, TS-R-1, in order to ensure safety under both normal and accident conditions of transport. The underlying philosophy is that safety is vested principally in the package and the design and performance criteria are related to the potential hazard. Type B packages are high duty packages which are used for the transport of the more radioactive materials, notably spent fuel and vitrified high-level waste (VHLW). Tests are specified in the IAEA Regulations to ensure the integrity of these packages in potential transport accidents involving impacts, fires or immersion in water. The mechanical tests for Type B packages include drop tests onto an unyielding surface without giving rise to a significant release of radioactivity. The objects which a package could impact in real life transport accidents, such as concrete roads, bridge abutments and piers, will yield to some extent and absorb some of the energy of the moving package. Impact tests onto an unyielding surface are therefore relevant to impacts onto real-life objects at much higher speeds. The thermal test specifies that Type B packages should be able to withstand a fully engulfing fire of 8000 C for 30 minutes. Analytical studies backed up by experimental tests have shown that these packages can withstand such conditions without significant release of radioactivity. The Regulations also specify immersion tests for Type B packages; 15 metres for 8 hours without significant release of radioactivity and, in addition for spent fuel and VHLW packages, 200 metres for 1 hour without rupture of the containment. Studies have shown that spent fuel and VHLW casks would meet these conditions. Therefore, there is a large body of evidence to show that the current IAEA Type B test requirements are severe and cover all the situations which can

  9. Whole-Pin Furnace system: An experimental facility for studying irradiated fuel pin behavior under potential reactor accident conditions

    International Nuclear Information System (INIS)

    Liu, Y.Y.; Tsai, H.C.; Donahue, D.A.; Pushis, D.O.; Savoie, F.E.; Holland, J.W.; Wright, A.E.; August, C.; Bailey, J.L.; Patterson, D.R.

    1990-05-01

    The whole-pin furnace system is a new in-cell experimental facility constructed to investigate how irradiated fuel pins may fail under potential reactor accident conditions. Extensive checkouts have demonstrated excellent performance in remote operation, temperature control, pin breach detection, and fission gas handling. The system is currently being used in testing of EBIR-II-irradiated Integral Fast Reactor (IFR) metal fuel pins; future testing will include EBR-II-irradiated mixed-oxide fuel pins. 7 refs., 4 figs

  10. Analysis of fuel behaviour after loss-of-coolant accident with the TESPA-code

    International Nuclear Information System (INIS)

    Keusenhoff, J.

    1981-01-01

    After a loss-of-coolant accident fuel rods go through a phase of high temperature and differential pressure before quenching and initiation of long term cooling. For licensing purpose the highest cladding temperature and the coolability of the core is of interest. The highest temperature is evaluated by a hot channel calculation with conservative assumptions. It gives little information about the status of the entire core. Therefore more detailed information is necessary. TESPA is a fast running code, which uses best-estimate assumptions, considers statistical uncertainties in the input parameters and calculates clad ballooning and rupture. The code is a usefull tool for calculation of channel blockage and cladding rupture

  11. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    International Nuclear Information System (INIS)

    Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

    1984-01-01

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

  12. Damage and failure of unirradiated and irradiated fuel rods tested under film boiling conditions

    International Nuclear Information System (INIS)

    Mehner, A.S.; Hobbins, R.R.; Seiffert, S.L.; MacDonald, P.E.; McCardell, R.K.

    1979-01-01

    Power-cooling-mismatch experiments are being conducted as part of the Thermal Fuels Behavior Program in the Power Burst Facility at the Idaho National Engineering Laboratory to evaluate the behavior of unirradiated and previously irradiated light water reactor fuel rods tested under stable film boiling conditions. The observed damage that occurs to the fuel rod cladding and the fuel as a result of film boiling operation is reported. Analyses performed as a part of the study on the effects of operating failed fuel rods in film boiling, and rod failure mechanisms due to cladding embrittlement and cladding melting upon being contacted by molten fuel are summarized

  13. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

  14. NPP Krsko Severe Accident Management Guidelines Upgrade

    International Nuclear Information System (INIS)

    Mihalina, Mario; Spalj, Srdjan; Glaser, Bruno; Jalovec, Robi; Jankovic, Gordan

    2014-01-01

    Nuclear Power Plant Krsko (NEK) has decided to take steps for upgrade of safety measures to prevent severe accidents, and to improve the means to successfully mitigate their consequences. The content of the program for the NEK Safety Upgrade is consistent with the nuclear industry response to Fukushima accident, which revealed many new insights into severe accidents. Therefore, new strategies and usage of new systems and components should be integrated into current NEK Severe Accident Management Guidelines (SAMG's). SAMG's are developed to arrest the progression of a core damage accident and to limit the extent of resulting releases of fission products. NEK new SAMG's revision major changes are made due to: replacement of Electrical Recombiners by Passive Autocatalytic Recombiners (PARs) and the installation of Passive Containment Filtered Vent System (PCFV); to handle a fuel damage situation in Spent Fuel Pool (SFP) and to assess risk of core damage situation during shutdown operation. (authors)

  15. Activity release from the damaged spent VVER-fuel during long-term wet storage

    Energy Technology Data Exchange (ETDEWEB)

    Slonszki, E.; Hozer, Z. [Hungarian Academy of Sciences, KFKI Atomic Energy Research Inst., Budapest (Hungary); Pinter, T.; Baracska Varju, I. [Nuclear Power Plant Paks, Paks (Hungary)

    2010-07-01

    An ex-core fuel damage incident took place at Unit 2 of Paks Nuclear Power Plant in Hungary on the 10{sup th} April 2003. After this event the damaged fuel assemblies were stored under water for four years. During wet storage a continuous activity release was observed. The evaluation of the measured activity concentration showed that the UO{sub 2} mass released from the fuel into the coolant was {approx} 1.8% of the total fuel mass. Furthermore this paper contains the calculation methods and the calculated activity release of the main analysed isotopes. (orig.)

  16. Activity release from the damaged spent VVER-fuel during long-term wet storage

    International Nuclear Information System (INIS)

    Slonszki, E.; Hozer, Z.; Pinter, T.; Baracska Varju, I.

    2010-01-01

    An ex-core fuel damage incident took place at Unit 2 of Paks Nuclear Power Plant in Hungary on the 10 th April 2003. After this event the damaged fuel assemblies were stored under water for four years. During wet storage a continuous activity release was observed. The evaluation of the measured activity concentration showed that the UO 2 mass released from the fuel into the coolant was ∼ 1.8% of the total fuel mass. Furthermore this paper contains the calculation methods and the calculated activity release of the main analysed isotopes. (orig.)

  17. The differentiated assessment of damage to economy of subjects of the Siberian Federal District from road and transport accident rate

    Science.gov (United States)

    Petrov, Artur I.; Svistunova, Vera A.; Petrova, Daria A.

    2018-01-01

    The results of assessment of damage from the road accident rate in subjects of the Siberian Federal District (SFD) of the Russian Federation are presented in the article. The thesis about spatial differentiation of the Gross Regional Product (GRP) losses in different regions of the country because of people’s death and injuries in the road accidents (RA) and due to formations of property and ecological damage was chosen as a working hypothesis. The calculations, carried out for 12 subjects of the SFD, confirmed this idea. The range of calculated values of economic damage from road accident rate (in % of GRP) was from 1.3 (Tomsk region) to 12.6 (Republic of Tyva) in 2015. In article the attempt to explain the received result by heterogeneous development of economics in various Russian regions is made. The consequence of it is a heterogeneous quality of people’s life and quite various perception of life value by inhabitants of different regions that influences their life safety level.

  18. Study on fracture of fuel element cladding for naval reactor during typical accidents

    International Nuclear Information System (INIS)

    Zhang Fan; Shang Xueli; Zheng Zhongliang; Yu Lei

    2011-01-01

    Aiming at defining the grade of nuclear emergency response, the best estimate model has been adopted; the simulation of large break loss of coolant accident (LBLOCA) has been carried out by the radioactive analysis software coupled with relap5/mod 3.2 and core physics model. First, the peak clad temperature of the critical failure channel is calculated in relap5 code, and simultaneously its power factor is obtained. Second, pin power distribution of the fuel assemblies has been calculated in coarse-mesh nodal method. According to the pin power distribution in the whole core and the result gained above, the fraction of fuel element fracture is calculated. Finally, the radioactive analysis has been carried out and the reasonable source term is gotten, which can offer reference for the nuclear emergency decision making. (authors)

  19. Fission product release profiles from spherical HTR fuel elements at accident temperatures

    International Nuclear Information System (INIS)

    Schenk, W.; Pitzer, D.; Nabielek, H.

    1986-10-01

    A total of 22 fuel elements with modern TRISO particles has been tested in the temperature range 1500-2500 0 C. Additionally, release profiles of iodine and other isotopes have been obtained with seven UO 2 samples at 1400-1800 0 C. For heating times up to 100 hours at the maximum temperature, the following results are pertinent to HTR accident conditions: Ag 110 m is the only fission products to be released at 1200-1600 0 C by diffusion through intact SiC, but it is of low significance in accident assessments; cesium, iodine, strontium, and noble gas releases up to 1600 0 C are solely due to various forms of contamination; at 1700-1800 0 C, corrosion induced SiC defects cause the release of Cs, Sr, I/Xe/Kr; above 2000 0 C, thermal decomposition of the silicon carbide layer sets in while pyrocarbons still remain intact. Around 1600 0 C, the accident specific contribution of cesium, strontium, iodine, and noble gases is negligible. (orig./HP) [de

  20. Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-01-01

    Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

  1. Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

    Directory of Open Access Journals (Sweden)

    Hyun-Gil Kim

    2016-02-01

    Full Text Available For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density composite pellet concepts are being developed as ATF pellets. A microcell UO2 pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts—surface-modified Zr-based alloy and SiC composite material—are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.

  2. Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

    Science.gov (United States)

    Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

    2018-02-01

    The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

  3. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  4. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

  5. Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)

    2017-04-15

    The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

  6. Accidents at nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The accidents which accurred at Wuergassen, Browns Ferry and Three Mile Island are each briefly described and discussed. The last is naturally treated in much more detail than the first two. Damage to the fuel elements is briefly considered and the release of fission products, radiation doses to the population and their expected consequences are discussed. The accidents are evaluated and related to risk evaluations, especially in WASH-1400. (JIW)

  7. Discussion on the re-irradiated fuel assembly with damaged guide vanes

    International Nuclear Information System (INIS)

    Li Ligang

    2013-01-01

    In January 2011, during the second plant of CNNC Nuclear Power Operations Management Co., Ltd.(hereinafter referred to as the second plant) refueling outage, the visual inspection found the guide vanes of fuel assembly A had felling off. After the National Nuclear Safety Administration (NNSA) estimated and approved, the fuel assembly A was reloaded in the specified location of reactor core. During the refueling outage in March 2012, the fuel assembly A was removed again from the reactor core. Visual inspection confirmed that the fuel assembly A was complete and without abnormal changes. The practice provides reference for re-irradiated of fuel assembly with the same type of damaged guide vanes, and provides case support for standard development for the same type of re-irradiated fuel assembly with damaged guide vanes. (author)

  8. Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

    International Nuclear Information System (INIS)

    Chang, L.K.; Koenig, J.F.; Porter, D.L.

    1987-01-01

    In the Shutdown Heat Removal Testing (SHRT) program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elemnts, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date. (orig.)

  9. Transport of volatile fission products in the fuel-to-sheath gap of defective fuel elements during normal and reactor accident conditions

    International Nuclear Information System (INIS)

    Lewis, B.J.; Bonin, H.W.

    1995-01-01

    An analytical treatment has been used to model the vapour transport of radioactive fission products released into the fuel-to-sheath gap of defective nuclear fuel elements. The model accounts for both diffusive and bulk-convective transport. Convective transport becomes important as a result of a significant release of gaseous fission products into the gap during a high-temperature reactor accident. However, during normal reactor operation, diffusion is shown to be the dominant process of transport. The model is based on an analysis of several in-reactor tests with operating defective fuel elements, and high-temperature annealing experiments with irradiated fuel specimens. ((orig.))

  10. The OECD/NEA workshop on the indemnification of nuclear damage in the event of a nuclear accident

    International Nuclear Information System (INIS)

    Wagstaff, F.

    2002-01-01

    Since 1993, the OECD Nuclear Energy Agency (OECD/NEA) has run the International Nuclear Emergency Exercise (INEX) Program. The program serves to discuss an effective accident management approach on the basis of a simulated nuclear accident situation together with the states involved and their institutions, and also elaborate measures for its further improvement. At the present time, the INEX Program has reached Phase 3 in which, for the first time, also aspects of liability for the consequences of accidents were included. These aspects were made the subject of a workshop held after an emergency exercise. The scenario covered was based on an INES level-4 accident in the French Gravelines Nuclear Power Station situated close to the French-Belgian border. The workshop dealt with these topics, among others: the application of the Paris Convention on Third Party Liability, the Brussels Supplementary Convention, and the Vienna Convention on Civil Liability for Nuclear Damage as well as the Supplementary Compensation Convention of 1997. It was seen that there was a clear need for further discussion, especially to shed more light on the interrelationship of these treaties. (orig.) [de

  11. Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

    International Nuclear Information System (INIS)

    Hozer, Zoltan; Szabo, Emese; Pinter, Tamas; Varju, Ilona Baracska; Bujtas, Tibor; Farkas, Gabor; Vajda, Nora

    2009-01-01

    During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.

  12. Activity release from damaged fuel during the Paks-2 cleaning tank incident in the spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Zoltan, E-mail: hozer@aeki.kfki.h [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest 114, P.O. Box 49 (Hungary); Szabo, Emese [Hungarian Academy of Sciences KFKI Atomic Energy Research Institute, H-1525 Budapest 114, P.O. Box 49 (Hungary); Pinter, Tamas; Varju, Ilona Baracska; Bujtas, Tibor; Farkas, Gabor [Nuclear Power Plant Paks, H-7031 Paks, P.O. Box 71 (Hungary); Vajda, Nora [Institute of Nuclear Techniques, Budapest University of Technology and Economics, H-1521 Budapest, Muegyetem rakpart 9 (Hungary)

    2009-07-01

    During crud removal operations the integrity of 30 fuel assemblies was lost at high temperature at the unit No. 2 of the Paks NPP. Part of the fission products was released from the damaged fuel into the coolant of the spent fuel storage pool. The gaseous fission products escaped through the chimney from the reactor hall. The volatile and non-volatile materials remained mainly in the coolant and were collected on the filters of water purification system. The activity release from damaged fuel rods during the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The typical release rate of noble gases, iodine and caesium was 1-3%. The release of non-volatile fission products and actinides was also detected.

  13. Study on severe accident fuel dispersion behavior in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.; Xiang, J.Y.

    1995-01-01

    Core flow blockage events are a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel could result in core heatup and melting under full coolant flow condition. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Heat transfer between melt particle and coolant, which affects particle breakup, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to relative motion of particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. Results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also core debris tends to move together upon melting and entrainment

  14. Study on severe accident fuel dispersion behavior in the advanced neutron source reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S. [Oak Ridge National Lab., TN (United States)] [and others

    1995-09-01

    Core flow blockage events have been determined to represent a leading contributor to core damage initiation risk in the Advanced Neutron Source (ANS) reactor. During such an accident, insufficient cooling of the fuel in a few adjacent blocked coolant channels out of several hundred channels, could also result in core heatup and melting under full coolant flow condition in other coolant channels. Coolant inertia forces acting on the melt surface would likely break up the melt into small particles. Under thermal-hydraulic conditions of ANS coolant channel, micro-fine melt particles are expected. Hat transfer between melt particle and coolant, which affects the particle breakup characteristics, was studied. The study indicates that the thermal effect on melt fragmentation seems to be negligible because the time corresponding to the breakup due to hydrodynamic forces is much shorter than the time for the melt surface to solidify. The study included modeling and analyses to predict transient behavior and transport of debris particles throughout the coolant system. The transient model accounts for the surface forces acting on the particle that results from the pressure variation on the surface, inertia, virtual mass, viscous force due to the relative motion of the particle in the coolant, gravitation, and resistance due to inhomogenous coolant velocity radially across piping due to possible turbulent coolant motions. The results indicate that debris particles would reside longest in heat exchangers because of lower coolant velocity there. Also they are entrained and move together in a cloud.

  15. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, Charles E.; Mustin, Tracy P.; Massey, Charles D.

    1999-01-01

    Since initiating the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel causing a degradation of the fuel assembly exposing fuel meat and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues and implementation challenges associated with the transport of MTR type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, implementation status, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be of interest to foreign research reactor operators, shippers, and cask vendors in evaluating the condition of their fuel to ensure it can be transported in accordance with appropriate cask certificate requirements. (author)

  16. High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nelson, Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parkison, Adam [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-12-23

    The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H2 environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H2O with the alumina reaction tube at 1700°C resulted in Al(OH)3

  17. SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

    International Nuclear Information System (INIS)

    Gulden, W.

    1977-08-01

    The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

  18. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance

    Directory of Open Access Journals (Sweden)

    Cheng Bo

    2016-01-01

    Full Text Available Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1 fabricability of long, thin wall Mo-alloy tubes, (2 formability of a protective outer coating, (3 weldability of Mo tube to endcaps, (4 corrosion resistance in autoclaves with simulated LWR coolant, (5 oxidation resistance to steam at 1000–1500 °C, and (6 sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the

  19. Reactivity initiated accident test series Test RIA 1-4 fuel behavior report

    International Nuclear Information System (INIS)

    Cook, B.A.; Martinson, Z.R.

    1984-09-01

    This report presents and discusses results from the final test in the Reactivity Initiated Accident (RIA) Test Series, Test RIA 1-4, conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Nine preirradiated fuel rods in a 3 x 3 bundle configuration were subjected to a power burst while at boiling water reactor hot-startup system conditions. The test resulted in estimated axial peak, radial average fuel enthalpies of 234 cal/g UO 2 on the center rod, 255 cal/g UO 2 on the side rods, and 277 cal/g UO 2 on the corner rods. Test RIA 1-4 was conducted to investigate fuel coolability and channel blockage within a bundle of preirradiated rods near the present enthalpy limit of 280 cal/g UO 2 established by the US Nuclear Regulatory Commission. The test design and conduct are described, and the bundle and individual rod thermal and mechanical responses are evaluated. Conclusions from this final test and the entire PBF RIA Test Series are presented

  20. Modification of Ca2+ - exchange in blood cells of cattles with radioiodine damage of thyroid gland after Chernobyl accident

    International Nuclear Information System (INIS)

    Shevchenko, A.S.; Kobyalko, V.O.; Shevchenko, T.S.; Lazarev, N.M.; Astasheva, N.P.; Aleksakhin, R.M.

    1993-01-01

    Increasing of Ca 2+ concentration in cytoplasm and of the rate of 45 Ca iflux into cows erythrocytes in 19-24 month efter Chernobyl accident was revealed. Correlation between Ca 2+ concentration in cytoplasm of erythrocytes and thyroxin content in plasma of cows with radioiodine damage of thyroid gland was found. Reduction of the rate of 45 Ga influx into erythrocytes in cows with radiation doses of 20-60 By on thyroid gland was shown in later time after accident (3-5 years). Changes in Ca 2+ permeability through membranes of erythrocytes and neutrophiles after injection of 131 I into calfs in doses of 300 Gy and more on thyroid gland was found

  1. Computer simulation of fuel behavior during loss-of-flow accidents in a gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Wehner, T.R.

    1980-01-01

    The sequence of events in a loss-of-flow accident without reactor shutdown in a gas-cooled fast breeder reactor is strongly influenced by the manner in which the fuel deforms. In order to predict the mode of initial gross fuel deformation, welling, melting or cracking, a thermomechanical computer simulation program was developed. Methods and techniques used make the simulation an economical, efficient, and flexible engineering tool. An innovative application of the enthalpy model within a finite difference scheme is used to caculate temperatures in the fuel rod. The method of successive elastic solutions is used to calculate the thermoelastic-creep response. Calculated stresses are compared with a brittle-fracture stress criterion. An independent computer code is used to calculate fission-gas-induced fuel swelling. Results obtained with the computer simulation indicate that swelling is not a mode of initial fuel deformation. Faster transients result in fuel melting, while slower transients result in fuel cracking. For investigated faster coolant flow coastdowns with time constants of 1 second and 10 seconds, compressive stresses in the outer radial portion of the fuel limit fuel swelling and inhibit fuel cracking. For a slower coolant flow coastdown with a 300 second time constant, tensile stresses in the outer radial portion of the fuel induce early fuel cracking before any melting or significant fuel swelling has occurred. Suggestions for further research are discussed. A derived noniterative solution for mechanics calculations may offer an order of magnitude decrease in computational effort

  2. Analytical and experimental assessment of TVS-2006 fuel assembly thermal-mechanical shape deformation at temperature modeling of a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Afanasiev, A.; Semishkin, V.; Makarov, V.; Matvienko, I.; Puzanov, D.

    2015-01-01

    Full or partial core drying-out takes place in loss-of-coolant accidents, which leads to worsening of heat removal from the fuel rods. Depending on the accident scenario the fuel rod cladding temperature can be in a wide range from 350 to 1200°C. It is worth mentioning, that the length of the process can considerably affect the fuel rod cladding loadcarrying capacity and the FA structure as a whole, and in the long run it defines the radiation consequences of the accident and the possibility of postaccident core disassembly at low cost. Most experiments staged of late were devoted to a study of FA behaviour in the temperature range 800-900°C of α→β phase transition that is characterized by a sharp increase in the rate of zirconium alloy creep which leads to fuel rod cladding ballooning and loss of their tightness within a short period of time. The 600-700°C temperature range turned out to be less investigated whereas this is the range where the change of zirconium alloy mechanical properties is also observed but only with the retention of α-phase. The tests of a full-scale FA dummy with the skeleton of guide tubes and spacer grids connected by friction forces, carried out at the testing facility of JSC OKB “GIDROPRESS”, were devoted to a study of FA behaviour in this temperature range. The model was heated up with hot air to 650°C for 6 hours. The tests ended with fuel rod cladding ballooning due to gauge pressure and shape deformation. No loss of fuel rod cladding integrity was observed. Therefore, a conclusion can be made that a long-time core holdup at the parameters implemented at the test facility is permitted and the deformations of the FA structure do not lead to the damage that could considerably complicate the core disassembly. The test results were used for the verification of the calculational model of FA TVS-2006 structure with a welded skeleton by ANSYS code. On the basis of the verified calculational model a calculational model was

  3. Issues and decisions for nuclear power plant management after fuel damage events

    International Nuclear Information System (INIS)

    1997-04-01

    Experience has shown that the on-site activities following an incident that results in severely damaged fuel at a nuclear power plant required extraordinary effort. Even in cases that are not extreme but in which fuel damage is greater than mentioned in the specifications for operation, the recovery will require extensive work. This publication includes information from several projects at the IAEA since 1989 that have resulted in a Technical Report, a TECDOC and a Workshop. While the initial purpose of the projects was focused on providing technical information transfer to the experts engaged in recovery work at the damaged unit of Chernobyl NPP, the results have led to a general approach to managing events in which there is substantial fuel damage. This TECDOC summarizes the work to focus on management issues that may be encountered in any such event whether small or large. 11 refs, 2 figs, 5 tabs

  4. An analysis of reactor structural response to fuel sodium interaction in a hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

    1976-01-01

    This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A., calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. This work was supported by a grant from Power Reactor and Nuclear Fuel Development Corporation. (auth.)

  5. Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

    Science.gov (United States)

    Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

    2018-02-01

    FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

  6. Fuel rod thermal analysis of the Angra-1 reactor during a postulated loss of coolant accident

    International Nuclear Information System (INIS)

    Praes, J.G.L.

    1982-01-01

    A thermal analysis of a fuel element is performed, as subject to the most severe cooling conditions, such as those occurring during a postulated Loss of Coolant Accident in the Angra-I reactor. Our objective was to ascertain whether the cooling of the core is assured according to 10 CRF - 50. According to the stated purpose, sensitivity analyses are necessary, using the swelling and rupture models of the cladding, and at the same time, an updating of the FLECHT heat transfer correlations in the computing program used, which is TOODEE-2 e 1 Version(28), with the purpose of adequating it to the Angra-I core analysis. In addition, we did sensitivity studies on heat transfer coefficient calculations for the steam cooling model. From the results obtained we conclude that the maximum temperature values of the cladding and the oxidation rate due to the Z sub(r) H 2 O reaction were kept well below the maximum allowable limits. Thus, the cooling of the Angra-I core is assured for the assumed accident. (Author) [pt

  7. Probabilistic Risk Assessment of Cask Drop Accident during On-site Spent Nuclear Fuel Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Jae Hyun; Christian, Robby; Momani, Belal Al; Kang, Hyun Gook [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    There are two ways to transfer the SNF from a site to other site, one is land transportation and the other is maritime transportation. Maritime transportation might be used because this way uses more safe route which is far from populated area. The whole transportation process can be divided in two parts: transferring the SNF between SNP and wharf in-Nuclear Power Plant (NPP) site by truck, and transferring the SNF from the wharf to the other wharf by ship. In this research, on-site SNF transportation between SNP and wharf was considered. Two kinds of single accident can occur during this type of SNF transportation, impact and fire, caused by internal events and external events. In this research, PRA of cask drop accident during onsite SNF transportation was done, risk to a person (mSv/person) from a case with specific conditions was calculated. In every 11 FEM simulation drop cases, FDR is 1 even the fuel assemblies are located inside of the cask. It is a quite larger value for all cases than the results with similar drop condition from the reports which covers the PRA on cask storage system. Because different from previous reports, subsequent impact was considered. Like in figure 8, accelerations which are used to calculate the FDR has extremely higher values in subsequent impact than the first impact for all SNF assemblies.

  8. Experience with fuel damage caused by abnormal conditions in handling and transporting operations

    International Nuclear Information System (INIS)

    Bailey, W.J.

    1983-01-01

    Pacific Northwest Laboratory (PNL) conducted a study to determine the expected condition of spent USA light-water reactor (LWR) fuel upon arrival at interim storage or fuel reprocessing facilities or, if fuel is declared a waste, at disposal facilities. Initial findings were described in an earlier PNL paper at PATRAM '80 and in a report. Updated findings are described in this paper, which includes an evaluation of information obtained from the literature and a compilation of cases of known or suspected damage to fuel as a result of handling and/or transporting operations. To date, PNL has evaluated 123 actual cases (98 USA and 25 non-USA). Irradiated fuel was involved in all but 10 of the cases. From this study, it is calculated that the frequency of unusual occurrences involving fuel damage from handling and transporting operations has been low. The damage that did occur was generally minor. The current base of experience with fuel handling and transporting operations indicates that nearly all of these unusual occurrences had only a minor or negligible effect on spent fuel storage facility operations

  9. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Gyori, C.; Schubert, A.; Laar, J. van de; Hozer, Z.; Spykman, G.

    2008-01-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events

  10. 2nd Gen FeCrAl ODS Alloy Development For Accident-Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Massey, Caleb P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Edmondson, Philip D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-01

    Extensive research at ORNL aims at developing advanced low-Cr high strength FeCrAl alloys for accident tolerant fuel cladding. One task focuses on the fabrication of new low Cr oxide dispersion strengthened (ODS) FeCrAl alloys. The first Fe-12Cr-5Al+Y2O3 (+ ZrO2 or TiO2) ODS alloys exhibited excellent tensile strength up to 800 C and good oxidation resistance in steam up to 1400 C, but very limited plastic deformation at temperature ranging from room to 800 C. To improve alloy ductility, several fabrication parameters were considered. New Fe-10-12Cr-6Al gas-atomized powders containing 0.15 to 0.5wt% Zr were procured and ball milled for 10h, 20h or 40h with Y2O3. The resulting powder was then extruded at temperature ranging from 900 to 1050 C. Decreasing the ball milling time or increasing the extrusion temperature changed the alloy grain size leading to lower strength but enhanced ductility. Small variations of the Cr, Zr, O and N content did not seem to significantly impact the alloy tensile properties, and, overall, the 2nd gen ODS FeCrAl alloys showed significantly better ductility than the 1st gen alloys. Tube fabrication needed for fuel cladding will require cold or warm working associated with softening heat treatments, work was therefore initiated to assess the effect of these fabrications steps on the alloy microstructure and properties. This report has been submitted as fulfillment of milestone M3FT 16OR020202091 titled, Report on 2nd Gen FeCrAl ODS Alloy Development for the Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign of the Fuel Cycle R&D program.

  11. Malignant neoplasms on the territories of Russia damaged owing to the Chernobyl accident

    International Nuclear Information System (INIS)

    Remennik, L.V.; Starinsky, V.V.; Mokina, V.D.; Chissov, V.I.; Scheplyagina, L.A.; Petrova, G.V.; Rubtsova, M.M.

    1996-01-01

    The work presents the results of descriptive analysis of development of onco epidemiological situation in six of the most polluted regions owing to the Chernobyl accident in 1981-1994. The growth of malignancies incidence is marked in all territories as well as in the Russian Federation as a whole. The most adverse tendencies have been revealed in the Bryansk, Orel, Ryazan regions. It is marked that the formation of a structure, levels and trends of the malignancies incidence has been occurring under influence of a complex of factors usual up to the accident. The analysis of the data from the specialized cancer-register evidences that the incidence of thyroid malignancies is actively growing in the population of the Bryansk region. The probability of connection of growth of the thyroid cancer incidence in children of the Bryansk region with the Chernobyl accident is reasonably high, but should be confirmed through the application of methods of analytical epidemiology

  12. The role of grain boundary fission gases in high burn-up fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Papin, J.; Frizonnet, J.M.; Cazalis, B.; Rigat, H.

    2002-01-01

    In the frame of reactivity-initiated accidents (RIA) studies, the CABRI REP-Na programme is currently performed, focused on high burn-up UO 2 and MOX fuel behaviour. From 1993 to 1998, seven tests were performed with UO 2 fuel and three with MOX fuel. In all these tests, particular attention has been devoted to the role of fission gases in transient fuel behaviour and in clad loading mechanisms. From the analysis of experimental results, some basic phenomena were identified and a better understanding of the transient fission gas behaviour was obtained in relation to the fuel and clad thermo-mechanical evolution in RIA, but also to the initial state of the fuel before the transient. A high burn-up effect linked to the increasing part of grain boundary gases is clearly evidenced in the final gas release, which would also significantly contribute to the clad loading mechanisms. (authors)

  13. DEFORM-4: fuel pin characterization and transient response in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Miles, K.J.; Hill, D.J.

    1986-01-01

    The DEFORM-4 module is the segment of the SAS4A Accident Analysis Code System that calculates the fuel pin characterization in response to a steady state irradiation history, thereby providing the initial conditions for the transient calculation. The various phenomena considered include fuel porosity migration, fission gas bubble induced swelling, fuel cracking and healing, fission gas release, cladding swelling, and the thermal-mechanical state of the fuel and cladding. In the transient state, the module continues the thermal-mechanical response calculation, including fuel melting and central cavity pressurization, until cladding failure is predicted and one of the failed fuel modules is initiated. Comparisons with experimental data have demonstrated the validity of the modeling approach

  14. Computer code for the analysis of destructive pressure generation process during a fuel failure accident, PULSE-2

    International Nuclear Information System (INIS)

    Fujishiro, Toshio

    1978-03-01

    The computer code PULSE-2 has been developed for the analysis of pressure pulse generation process when hot fuel particles come into contact with the coolant in a fuel rod failure accident. In the program, it is assumed that hot fuel fragments mix with the coolant instantly and homogeneously in the failure region. Then, the rapid vaporization of the coolant and transient pressure rise in failure region, and the movement of ejected coolant slugs are calculated. The effect of a fuel-particle size distribution is taken into consideration. Heat conduction in the fuel particles and heat transfer at fuel-coolant interface are calculated. Temperature, pressure and void fraction in the mixed region are calculated from the average enthalpy. With physical property subroutines for liquid sodium and water, the model is usable for both LMFBR and LWR conditions. (auth.)

  15. Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

    International Nuclear Information System (INIS)

    Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo

    2016-01-01

    The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated

  16. Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated.

  17. An Analysis of Reactor Structural Response to Fuel Sodium Interaction in a Hypothetical Core Disruptive Accident

    International Nuclear Information System (INIS)

    Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

    1976-01-01

    This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A, calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. In conclusion: FSI phenomena depend highly on constraints around FSI zone, so that the constraints must be dealt with realistically in analytical models. Although a two-dimensional model is superior to a quasi-two-dimensional model. The former needs long calculation time, so it is very expensive using in parametric study. Therefore, it is desirable that the two-dimensional model is used in the final study of reactor design and the quasi-two-dimensional model is used in parametric study. The blanket affects on the acoustic pressure and the deformations of radial structures, but affects scarcely on the upper vessel deformation. The blanket also affects on the mechanical work largely. The core barrel gives scarcely the effects on pressure in single phase but gives highly the effects on pressure in two-phase and deformation of reactor structures in this study. For studying the more realistic phenomena of FSI in the reactor design, the following works should be needed. (i) Spatial Distribution of FSI Region Spatial and time-dependent distribution of fuel temperature and molten fuel fraction must be taken in realistic simulation of accident condition. To this purpose, the code will

  18. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    International Nuclear Information System (INIS)

    Lindley, B.A.; Lillington, J.N.; Kotlyar, D.; Parks, G.T.; Petrovic, B.

    2016-01-01

    The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO_2/PuO_2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs)-clad systems, particularly for current and near-term build LWRs. R and D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN) and uranium silicide (U_3Si_2). Candidate cladding materials include advanced stainless steel (FeCrAl) and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R and D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I"2S-LWR), a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I"2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I"2S-LWR design are U_3Si_2 and advanced stainless steel respectively. In addition, the I"2S-LWR design

  19. Severe fuel damage investigations of KFK/PNS

    International Nuclear Information System (INIS)

    Fiege, A.

    1983-01-01

    This report is a comprehensive review of the objectives, the program planning, the status and the further procedure of the investigations of KfK/PNS on severe core damage. The investigations were started in 1981 and will be finished in 1985/86. (orig.) [de

  20. Theoretical analysis of the temperature changes and resultant loss of fuel integrity in the IEA-R1 research reactor fuel elements following a loss of coalant accident

    International Nuclear Information System (INIS)

    Garone, J.G.M.

    1983-01-01

    The IEA-R1 core following a loss of coolant accident (LOCA) is analysed. THe AIRLOCA code was used to calculate fuel temperatures, heat generation due to fission product decay and convective and radiative heat transfer from the fuel elements to the surrounding air both during and following the loss of coolant. The influence of certain critical parameters, such as log time, specific power was studied in detail. Representative results are presented and suggestions made to ensure that fuel integrity is maintained following a LOCA. (Author) [pt

  1. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    International Nuclear Information System (INIS)

    Zhang, Yongfeng

    2016-01-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  2. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Laboratory

    2016-09-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  3. Melt Fragmentation Characteristics of Metal Fuel with Melt Injection Mass during Initiating Phase of SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Lee, Min Ho; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of)

    2016-05-15

    The PGSFR has adopted the metal fuel for its inherent safety under severe accident conditions. However, this fuel type is not demonstrated clearly yet under the such severe accident conditions. Additional experiments for examining these issues should be performed to support its licensing activities. Under initiating phase of hypothetic core disruptive accident (HCDA) conditions, the molten metal could be better dispersed and fragmented into the coolant channel than in the case of using oxide fuel. This safety strategy provides negative reactivity driven by a good dispersion of melt. If the coolant channel does not sufficient coolability, the severe recriticality would occur within the core region. Thus, it is important to examine the extent of melt fragmentation. The fragmentation behaviors of melt are closely related to a formation of debris shape. Once the debris shape is formed through the fragmentation process, its coolability is determined by the porosity or thermal conductivity of the melt. There were very limited studies for transient irradiation experiments of the metal fuel. These studies were performed by Transient Reactor Test Facility (TREAT) M series tests in U.S. The TREAT M series tests provided basic information of metal fuel performance under transient conditions. The effect of melt injection mass was evaluated in terms of the fragmentation behaviors of melt. These behaviors seemed to be similar between single-pin and multi-pins failure condition. However, the more melt was agglomerated in case of multi-pins failure.

  4. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    International Nuclear Information System (INIS)

    Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-01-01

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  5. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-04-15

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  6. In-core fuel disruption experiments simulating LOF accidents for homogeneous and heterogeneous core LMFBRs: FD2/4 series

    International Nuclear Information System (INIS)

    Wright, S.A.; Mast, P.K.; Schumacher, Gustav; Fischer, E.A.

    1982-01-01

    A series of Fuel Disruption (FD) experiments simulating LOF accidents transients for homogeneous- and heterogeneous-core LMFBRs is currently being performed in the Annular Core Research Reactor at SNL. The test fuel is observed with high-speed cinematography to determine the timing and the mode of the fuel disruption. The five experiments performed to date show that the timing and mode of fuel disruption depend on the power level, fuel temperature (after preheat and at disruption), and the fuel temperature gradient. Two basic modes of fuel disruption were observed; solid-state disruption and liquid-state swelling followed by slumping. Solid-state dispersive fuel behavior (several hundred degrees prior to fuel melting) is only observed at high power levels (6P 0 ), low preheat temperatures (2000 K), and high thermal gradients (2800 K/mm). The swelling/slumping behavior was observed in all cases near the time of fuel melting. Computational models have been developed that predict the fuel disruption modes and timing observed in the experiments

  7. Probability analysis of WWER-1000 fuel elements behavior under steady-state, transient and accident conditions of reactor operation

    International Nuclear Information System (INIS)

    Tutnov, A.; Alexeev, E.

    2001-01-01

    'PULSAR-2' and 'PULSAR+' codes make it possible to simulate thermo-mechanical and thermo-physical parameters of WWER fuel elements. The probabilistic approach is used instead of traditional deterministic one to carry out a sensitive study of fuel element behavior under steady-state operation mode. Fuel elements initial parameters are given as a density of the probability distributions. Calculations are provided for all possible combinations of initial data as fuel-cladding gap, fuel density and gas pressure. Dividing values of these parameters to intervals final variants for calculations are obtained . Intervals of permissible fuel-cladding gap size have been divided to 10 equal parts, fuel density and gas pressure - to 5 parts. Probability of each variant realization is determined by multiplying the probabilities of separate parameters, because the tolerances of these parameters are distributed independently. Simulation results are turn out in the probabilistic bar charts. The charts present probability distribution of the changes in fuel outer diameter, hoop stress kinetics and fuel temperature versus irradiation time. A normative safety factor is introduced for control of any criterion realization and for determination of a reserve to the criteria failure. A probabilistic analysis of fuel element behavior under Reactivity Initiating Accident (RIA) is also performed and probability fuel element depressurization under hypothetical RIA is presented

  8. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, C.E.; Mustin, T.P.; Massey, C.D.

    1998-01-01

    Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF

  9. Visualization Study of Melt Dispersion Behavior for SFR with a Metallic Fuel under Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo Heo; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Jungang Univ., Seoul (Korea, Republic of)

    2015-05-15

    The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition.

  10. SSYST. A code system to analyze LWR fuel rod behavior under accident conditions

    International Nuclear Information System (INIS)

    Gulden, W.; Meyder, R.; Borgwaldt, H.

    1982-01-01

    SSYST (Safety SYSTem) is a modular system to analyze the behavior of light water reactor fuel rods and fuel rod simulators under accident conditions. It has been developed in close cooperation between Kernforschungszentrum Karlsruhe (KfK) and the Institut fuer Kerntechnik und Energiewandlung (IKE), University Stuttgart, under contract of Projekt Nukleare Sicherheit (PNS) at KfK. Although originally aimed at single rod analysis, features are available to calculate effects such as blockage ratios of bundles and wholes cores. A number of inpile and out-of-pile experiments were used to assess the system. Main differences versus codes like FRAP-T with similar applications are (1) an open-ended modular code organisation, (2) availability of modules of different sophistication levels for the same physical processes, and (3) a preference for simple models, wherever possible. The first feature makes SSYST a very flexible tool, easily adapted to changing requirements; the second enables the user to select computational models adequate to the significance of the physical process. This leads together with the third feature to short execution times. The analysis of transient rod behavior under LOCA boundary conditions e.g. takes 2 mins cpu-time (IBM-3033), so that extensive parametric studies become possible

  11. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  12. Study on recriticality of fuel debris during hypothetical severe accidents in the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.; Shin, S.T.

    1995-09-01

    A study has been performed to measure the potential of recriticality during hypothetical severe accident in Advanced Neutron Source (ANS). For the lumped debris configuration in the Reactor Coolant System (RCS), as found in the previous study, recriticality potential may be very low. However, if fuel debris is dispersed and mixed with heavy water in RCS, recriticality potential has been predicted to be substantial depending on thermal-hydraulic conditions surrounding fuel debris mixture. The recriticality potential in RCS is substantially reduced for the three element core design with 50% enrichment. Also, as observed in the previous study, strong dependencies of k eff on key thermal hydraulic parameters are shown. Light water contamination is shown to provide a positive reactivity, and void formation due to boiling of mixed water provides enough negative reactivity and to bring the system down to subcritical. For criticality potential in the subpile room, the lumped debris configuration does not pose a concern. Dispersed configuration in light water pool of the subpile room is also unlikely to result in criticality. However, if the debris is dispersed in the pool that is mixed with heavy water, the results indicate that a substantial potential exists for the debris to reach the criticality. However, if prompt recriticality disperses the debris completely in the subpile room pool, subsequent recriticality may be prevented since neutron leakage effects become large enough

  13. Study of source term evaluation from fuel solution under simulated nuclear criticality accident in TRACY

    International Nuclear Information System (INIS)

    Abe, Hitoshi; Tashiro, Shinsuke; Nagai, Hitoshi; Koike, Tadao; Okagawa, Seigo; Murata, Mikio

    1999-01-01

    In a accident at the dissolver in a reprocessing plant, various fission products and radiolysis gases will be produced in the fuel solution and volatile radioactive nuclides and radiolysis gases and nitrogen oxide will be released into vent-gas spontaneously. Moreover other on-volatile nuclide will be releases as radioactive aerosol (mist) with bursting bubbles at surface of the solution. Therefore quantitative estimation of release and transport behavior of the radioactive material from solution as source term is very important. TRACY is a transient criticality experimental facility for studying the transient criticality characteristics of low enriched uranium. In this paper, experiment methods and results about the release behavior of the hydrogen, radioactive aerosol and iodine species from the fuel solutions are reported. As the results of the experiments, release patterns of H 2 , 140 Ba and 131 I could be grasped. Concentrations of H 2 in the vent-gas and 140 Ba in the gas phase in the core tank attained to the peak just after the transient criticality and decreased exponentially with time. On the other hand, concentrations of 131 I in the gas phase of the tank began to increase with a time lag of several minutes from the transient criticality and attained approximately constant values. (J.P.N.)

  14. Influence of the Chernobyl accident on radioactivity of fuel peat and peat ash in Finland

    International Nuclear Information System (INIS)

    Mustonen, R.; Salonen, S.; Itkonen, A.

    1988-04-01

    The accident at the Chernobyl nuclear power plant in April 1986 caused very uneven deposition of radionuclides in Finland. The deposited radionuclides were measured in relative high concentrations in fuel peat and especially in peat ash. The radionuclide concentrations were measured at six peat-fired power plants in different parts of Finland throughout the heating season 1986-87. Also evaporation of different radionuclides in peat combustion and their condensation on fly ash particles were studied at four power plants. The 137 Cs-concentrations in compiled peat samples varied between 30 and 3600 Bq kg -1 dry weight and in ash samples between 600 and 68000 Bq kg -1 . Differences in radionuclide concentrations between the power plants were great and also the radionuclide composition in fuel peat varied regionally. The 137 Cs-concentrations of the fly ash after the ash precipitators varied between 12000 and 120000 Bq kg -1 and fly ash emissions varied from 17 to 1100 mg m -3 , depending on the power plant and the load of the boiler. High radioactivity concentrations in precipitator ash caused some restrictions to the utilization of peat ash for various purposes

  15. Study of fission products (Cs, Ba, Mo, Ru) behaviour in irradiated and simulated nuclear fuels during severe accidents using X-ray absorption Spectroscopy, SIMS and EPMA

    International Nuclear Information System (INIS)

    Geiger, Ernesto

    2016-01-01

    The identification of Fission Products (FP) release mechanism from irradiated nuclear fuels during a severe accident is of main importance for the development of codes for the estimation of the source-term (nature and quantity of radionuclides released into the environment). among the many FP Ba, Cs, Mo and Ru present a particular interest, since they may interact with each other or other elements and thus affect their release. In the framework of this thesis, two work axes have been set up in order to identify, firstly, the chemical phases initially present before the accident and, secondly, their evolution during the accident itself. The experimental approach consisted in reproducing nuclear severe accidents conditions at laboratory scale using both irradiated fuels and model materials (natural UO_2 doped with 12 FP). The advantage of these latter is the possibility of using characterization methods such as X-ray absorption Spectroscopy which are not available for irradiated fuels. Three irradiated fuel samples have been studied, representative to an initial state (before the accident), to an intermediate stage (1773 K) and to an advanced stage (2873 K) of a nuclear severe accident. Regarding to model materials, many accident sequences have been carried out, from 573 to 1973 K. Experimental results have allowed to establish a new release mechanism, considering both reducing and oxidizing conditions during an accident. These results have also demonstrated the importance of model materials as a complement to irradiated nuclear fuels in the study of nuclear severe accidents. (author) [fr

  16. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    International Nuclear Information System (INIS)

    1996-01-01

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

  17. International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

  18. Fuel containment, lightning protection and damage tolerance in large composite primary aircraft structures

    Science.gov (United States)

    Griffin, Charles F.; James, Arthur M.

    1985-01-01

    The damage-tolerance characteristics of high strain-to-failure graphite fibers and toughened resins were evaluated. Test results show that conventional fuel tank sealing techniques are applicable to composite structures. Techniques were developed to prevent fuel leaks due to low-energy impact damage. For wing panels subjected to swept stroke lightning strikes, a surface protection of graphite/aluminum wire fabric and a fastener treatment proved effective in eliminating internal sparking and reducing structural damage. The technology features developed were incorporated and demonstrated in a test panel designed to meet the strength, stiffness, and damage tolerance requirements of a large commercial transport aircraft. The panel test results exceeded design requirements for all test conditions. Wing surfaces constructed with composites offer large weight savings if design allowable strains for compression can be increased from current levels.

  19. Japan's compensation system for nuclear damage - As related to the TEPCO Fukushima Daiichi nuclear accidents

    International Nuclear Information System (INIS)

    Nomura, Toyohiro; Matsuura, Shigekazu; Takahashi, Yasufumi; Takenaka, Chihiro; Hokugo, Taro; Kamada, Toshihiko; Kamai, Hiroyuki

    2012-01-01

    Following the TEPCO Fukushima Daiichi nuclear power plant accident, extraordinary efforts were undertaken in Japan to implement a compensation scheme for the proper and efficient indemnification of the affected victims. This publication provides English translations of key Japanese legislative and administrative texts and other implementing guidance, as well as several commentaries by Japanese experts in the field of third party nuclear liability. The OECD Nuclear Energy Agency (NEA) has prepared this publication in co-operation with the government of Japan to share Japan's recent experience in implementing its nuclear liability and compensation regime. The material presented in the publication should provide valuable insights for those wishing to better understand the regime applied to compensate the victims of the accident and for those working on potential improvements in national regimes and the international framework for third party nuclear liability

  20. Full-length high-temperature severe fuel damage test No. 2

    International Nuclear Information System (INIS)

    Hesson, G.M.; Lombardo, N.J.; Pilger, J.P.; Rausch, W.N.; King, L.L.; Hurley, D.E.; Parchen, L.J.; Panisko, F.E.

    1993-09-01

    Hazardous conditions associated with performing the Full-Length High- Temperature (FLHT). Severe Fuel Damage Test No. 2 experiment have been analyzed. Major hazards that could cause harm or damage are (1) radioactive fission products, (2) radiation fields, (3) reactivity changes, (4) hydrogen generation, (5) materials at high temperature, (6) steam explosion, and (7) steam pressure pulse. As a result of this analysis, it is concluded that with proper precautions the FLHT- 2 test can be safely conducted

  1. Accidents in the operation of nuclear power stations. Action for damages against foreign operators

    International Nuclear Information System (INIS)

    Willner, K.

    1986-01-01

    On the occasion of a lecture evening of the Leo Goodman Library in Munich questions of civil liability of foreign reactor operators in cases of nuclear accidents were discussed by participants of various universities. Special subjects were i.a. problems of civil procedural and insurance law, absolute liability according to sec. 25 Atomic Energy Act as well as questions of applicable law. (WG) [de

  2. Precursors to potential severe core damage accidents: 1992, a status report

    International Nuclear Information System (INIS)

    1993-12-01

    This document is part of a report which documents 1992 operational events selected as accident sequence precursors. This report describes the 27 precursors identified from the 1992 licensee event reports. It also describe containment-related events; open-quote interesting close-quote events; potentially significant events that were considered impractical to analyze; copies of the licensee event reports which were cited in the cases above; and comments from the licensee and NRC in response to the preliminary reports

  3. Determination of possible damage/degradation of the Sandia National Laboratories Personal Nuclear Accident Dosimeter (PNAD)

    International Nuclear Information System (INIS)

    Potter, Charles Augustus; Ward, Dann C.

    2008-01-01

    This report describes the results of an inspection performed on the existing stock of SNL Personal Nuclear Accident Dosimeters (PNADs). The current stock is approximately 20 years old, and has not been examined since their initial acceptance. A small random sample of PNADs were opened (a destructive process) and the contents visually examined. Sample contents were not degraded and indicate that the existing stock of SNL PNADs is acceptable for continued use

  4. Precursors to potential severe core damage accidents: 1992, A status report

    International Nuclear Information System (INIS)

    Cox, D.F.; Cletcher, J.W.; Copinger, D.A.; Cross-Dial, A.E.; Morris, R.H.; Vanden Heuvel, L.N.; Dolan, B.W.; Jansen, J.M.; Minarick, J.W.; Lau, W.; Salyer, W.D.

    1993-12-01

    Twenty-seven operational events with conditional probabilities of subsequent severe core damage of 1.0 x 10E-06 or higher occurring at commercial light-water reactors during 1992 are considered to be precursors to potential core damage. These are described along with associated significance estimates, categorization, and subsequent analyses. The report discusses (1) the general rationale for this study, (2) the selection and documentation of events as precursors, (3) the estimation and use of conditional probabilities of subsequent severe core damage to rank precursor events, and (4) the plant models used in the analysis process

  5. Pressure and Temperature of the Room 1 for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-08-15

    This report deals with the prediction of the pressure and temperature of the room 1 for the pipe break accidents of the 3-pin fuel test loop. The 3-pin fuel test loop is an experimental facility for nuclear fuel tests at the operation conditions similar to those of PWR and CANDU power plants. Because the most processing systems of the 3-pin fuel test loop are placed in the room 1. The structural integrity of the room 1 should be evaluated for the postulated accident conditions. Therefore the pressures and temperatures of the room 1 needed for the structural integrity evaluation have been calculated by using MARS code. The pressures and temperatures of the room 1 have been calculated in various conditions such as the thermal hydraulic operation parameters, the locations of pipe break, and the thermal properties of the room 1 wall. It is assumed that the pipe break accident occurs in the letdown operation without regeneration, because the mass and energy release to the room 1 is expected to be the largest. As a result of the calculations the maximum pressure and temperature are predicted to be 208kPa and 369.2K(96.0 .deg. C) in case the heat transfer is considered in the room 1 wall. However the pressure and temperature are asymptotically 243kPa and 378.1K(104.9 .deg. C) assuming that the heat transfer does not occur in the room 1 wall.

  6. A study on items necessary to develop the requirements for the management of serious accidents postulated in fuel fabrication, enrichment and reprocessing facilities

    International Nuclear Information System (INIS)

    Takanashi, Mitsuhiro; Yamate, Kazuki; Asada, Kazuo; Yamada, Takashi; Endo, Shigeki

    2013-05-01

    The purpose of this study is to supply the points to discuss on new rules of fuel fabrication, enrichment and reprocessing facilities (hereinafter referred to as 'fuel cycle facilities') conducted by Nuclear Regulation Authority. Requirements for management of serious accidents in the fuel cycle facilities were summarized in this study. Taking into account the lessons learned from the accident of TEPCO Fukushima Daiichi Nuclear Power Plant in Mar. 2011, Act for the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors was amended in June 2012. The main items of the amendment were as follows: Preparation for the management of serious accidents, Introduction of evaluation system for safety improvement, Application of new standards to existing nuclear facilities (back-fitting). Japan Nuclear Energy Safety organization (JNES) conducted a fundamental study on serious accidents and their management in the fuel cycle facilities and made a report. In the report, the concept of Defense in Depth and the definition of serious accidents for the fuel cycle facilities were discussed. Those discussions were conducted by reference to new regulation rules (draft) for power reactors and from the view of features of the fuel cycle facilities. However, further detailed studies are necessary in order to clarify some issues in it. It was also reflected opinions from experts in JNES technical meetings on accident management of the fuel cycle facilities to brush up this report. (author)

  7. Precursors to potential severe core damage accidents. A status report, 1982--1983

    Energy Technology Data Exchange (ETDEWEB)

    Forester, J.A.; Mitchell, D.B.; Whitehead, D.W. [and others

    1997-04-01

    This study is a continuation of earlier work that evaluated 1969-1981 and 1984-1994 events affecting commercial light-water reactors. One-hundred nine operational events that affected 51 reactors during 1982 and 1983 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10{sup {minus}6}. These events were identified by first computer screening the 1982-83 licensee event reports from commercial light-water reactors to select events that could be precursors to core damage. Candidates underwent engineering evaluation that identified, analyzed, and documented the precursors. This report discusses the general rationale for the study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  8. Precursors to potential severe core damage accidents. A status report, 1982--1983

    International Nuclear Information System (INIS)

    Forester, J.A.; Mitchell, D.B.; Whitehead, D.W.

    1997-04-01

    This study is a continuation of earlier work that evaluated 1969-1981 and 1984-1994 events affecting commercial light-water reactors. One-hundred nine operational events that affected 51 reactors during 1982 and 1983 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by first computer screening the 1982-83 licensee event reports from commercial light-water reactors to select events that could be precursors to core damage. Candidates underwent engineering evaluation that identified, analyzed, and documented the precursors. This report discusses the general rationale for the study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  9. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

  10. Development of Innovative Accident Tolerant High Thermal Conductivity UO2-Diamond Composite Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Tulenko, James [Univ. of Florida, Gainesville, FL (United States); Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States)

    2016-01-01

    The University of Florida (UF) evaluated a composite fuel consisting of UO2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO2 – SiC and UO2 – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO2 may greatly enhanced the thermal conductivity of the UO2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

  11. Analysis of iodine chemical form noted from severe fuel damage experiments

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Osetek, D.J.

    1986-01-01

    Data from the TMI-2 accident has shown that only small amounts of iodine (I) escaped the plant. The postulated reason for such limited release is the formation of CsI (a salt) within fuel, which remains stable in a reducing high-temperature steam-H 2 environment. Upon cooldown CsI would dissolve in water condensate to form an ionic solution. However, recent data from fuel destruction experiments indicate different iodine release behavior that is tied to fuel burnup and oxidation conditions, as well as fission product concentration levels in the steam/H 2 effluent. Analysis of the data indicate that at low-burnup conditions, atomic I release from fuel is favored. Likewise, at low fission product concentration conditions HI is the favored chemical form in the steam/H 2 environment, not CsI. Results of thermochemical equilibria and chemical kinetics analysis support the data trends noted from the PBF-SFD tests. An a priori assumption of CsI for risk analysis of all accident sequences may therefore be inappropriate

  12. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-25

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  13. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    International Nuclear Information System (INIS)

    Heuser, Brent; Stubbins, James; Kozlowski, Tomasz; Uddin, Rizwan; Trinkle, Dallas; Downar, Thoms; Was, Gary; Ang, Yong; Phillpot, Simon; Sabharwall, Piyush

    2017-01-01

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  14. Experimental data report for Test TS-2 reactivity initiated accident test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Kobayashi, Shinsho; Yamahara, Takeshi; Sukegawa, Tomohide; Kikuchi, Teruo

    1993-02-01

    This report presents experimental data for Test TS-2 which was the second test in a series of Reactivity Initiated Accident (RIA) condition test using pre-irradiated BWR fuel rods, performed at the Nuclear Safety Research Reactor (NSRR) in February, 1990. Test fuel rod used in the Test TS-2 was a short sized BWR (7x7) type rod which was fabricated from a commercial rod irradiated at Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79% and a burnup of 21.3Gwd/tU (bundle average). A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 72±5cal/g·fuel (66±5cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and, results of pre and post pulse irradiation examinations are described in this report. (author)

  15. SIFAIL: a subprogram to calculate cladding deformation and damage for fast reactor fuel pins

    International Nuclear Information System (INIS)

    Wilson, D.R.; Dutt, D.S.

    1979-05-01

    SIFAIL is a series of subroutines used in conjunction with the thermal performance models of SIEX to assist in the evaluation of mechanical performance of mixed uranium plutonium oxide fuel pins. Cladding deformations due to swelling and creep are calculated. These have been compared to post-irradiation data from fuel pin tests in EBR-II. Several fuel pin cladding failure criteria (cumulative damage, total strain, and thermal creep strain) are evaluated to provide the fuel pin designer with a basis to select design parameters. SIFAIL allows the user many property options for cladding material. Code input is limited to geometric and environmental parameters, with a consistent set of material properties provided by the code. The simplified, yet adequate, thin wall stress--strain calculations provide a reliable estimate of fuel pin mechanical performance, while requiring a small amount of core storage and computer running time

  16. Fission product release measured during fuel damage tests at the Power Burst Facility

    International Nuclear Information System (INIS)

    Osetek, D.J.; Hartwell, J.K.; Vinjamuri, K.; Cronenberg, A.W.

    1985-01-01

    Results are presented of fission product release behavior observed during four severe fuel damage tests on bundles of UO 2 fuel rods. Transient temperatures up to fuel melting were obtained in the tests that included both rapid quench and slow cooldown, low and high (36 GWd/t) burnup fuel and the addition of Ag-In-Cd control rods. Release fractions of major fission product species and release rates of noble gas species are reported. Significant differences in release behavior are discussed between heatup and cooldown periods, low and high burnup fuel and long- and short-lived fission products. Explanations are offered for the probable reasons for the observed differences and recommendations for further studies are given

  17. PCDP [Prototypical Spent Fuel Consolidation Equipment Demonstration Project] design basis accident report 9315-P-103, Rev. A

    International Nuclear Information System (INIS)

    1987-12-01

    The Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) has identified a requirement to integrate the spent fuel rod consolidation design activities of each of several proposed geological repository facilities and the Monitored Retrievable Storage (MRS) facility, and to develop efficient and cost-effective equipment for the consolidation process. The equipment to be developed for the rod consolidation system will be required to operate in a dry environment at rates which can be appropriately scaled to approximate the waste management system acceptance rates, irrespective of repository geologic characteristics or the existence of an MRS facility in the waste management system. The purpose of this report is to identify and analyze the range of facility credible events and accident occurrences (from minor to the design basis accidents) and their causes and consequences. For each situation, the considerations to prevent or mitigate the event or accident is addressed

  18. The nuclear accident of Fukushima Daiichi - Soil contamination between the damaged reactors and the Pacific Ocean

    International Nuclear Information System (INIS)

    2013-01-01

    This document more particularly addresses the issue of management of contaminated waters present on the Fukushima site. It comments the assessment of contaminated water volumes, the presence of contaminated waters under the reactor buildings and under the turbine buildings which are a major risk of pollution for underground waters. It evokes the results of measurements of a high activity of Tritium and Strontium between the plant and the ocean, and discusses the possible origins of this increased radioactivity, and possible actions to create a barrier between the ocean and underground water. Some lessons learned from the Chernobyl accident are evoked

  19. Report on the preliminary fact finding mission following the accident at the nuclear fuel processing facility in Tokaimura, Japan

    International Nuclear Information System (INIS)

    1999-01-01

    Following the accident on 30 September 1999 at the nuclear fuel processing facility at Tokaimura, Japan, the IAEA Emergency Response Centre received numerous requests for information about the event's causes and consequences from Contact Points under the Conventions on Early Notification of a Nuclear Accident and on Assistance in the Case of a Nuclear Accident or Radiological Emergency. Although the lack of transboundary consequences of the accident meant that action under the Early Notification Convention was not triggered, the Emergency Response Centre issued several advisories to Member States which drew on official reports received from Japan. After discussions with the Government of Japan, the IAEA dispatched a team of three experts from the Secretariat on a fact finding mission to Tokaimura from 13 to 17 October 1999. The present preliminary report by that team documents key technical information obtained during the mission. At this stage, the report can in no way provide conclusive judgements on the causes and consequences of the accident. Investigations are proceeding in Japan and more information is expected to be made available after access has been gained to the building where the accident occurred. Moreover, much of the information already made available will be revised as more accurate assessments are made, for example of the radiation doses to the three individuals who received the highest exposures. Notwithstanding the preliminary nature of this report, it is clear that the accident was not one involving widespread contamination of the environment as in the 1986 Chernobyl accident. Although there was little risk off the site once the accident had been brought under control, the authorities evacuated the population living within a few hundred metres and advised people within about 10 km of the facility to take shelter for a period of about one day. The event at Tokaimura was nevertheless a serious industrial accident. The results of the detailed

  20. Resumption of pulsing the NSCR following the discovery of damaged fuel

    International Nuclear Information System (INIS)

    Feltz, D.E.; Rogers, R.D.

    1984-01-01

    Pulsing operations of the Nuclear Science Center Reactor (NSCR) at Texas A and M University were terminated in 1976 following the discovery of three damaged fuel elements during a routine inspection. A commitment was then made to the U.S. Nuclear Regulatory Commission to terminate pulsing of the NSCR until a thorough study of the damaged fuel had been completed. A report describing that study and discussing the possible mechanism of damage was issued in 1981. Based on a recommendation in the report to establish a limiting temperature to protect against damage, the USNRC issued a letter authorizing the reinitiation of pulsing the NSCR but limiting pulsing parameters 'to those in the current technical specifications or to a maximum calculated fuel temperature of 830 deg. C. It is felt based on the data obtained and fuel inspection results that the requirements of Phase I and Phase III of the Pulse Test Program for Core VIII have been met. Phase II of the test program will not be implemented unless there is a requirement for higher pulse energy and flux. The reproducibility of pulse data was very satisfactory

  1. A view of treatment process of melted nuclear fuel on a severe accident plant using a molten salt system

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R.; Takahashi, Y.; Nakamura, H.; Mizuguchi, K. [Power and Industrial Research and Development Center, Toshiba Corporation Power Systems Company, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan); Oomori, T. [Chemical System Design and Engineering Department, Toshiba Corporation Power Systems Company, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan)

    2013-07-01

    At severe accident such as Fukushima Daiichi Nuclear Power Plant Accident, the nuclear fuels in the reactor would melt and form debris which contains stable UO2-ZrO2 mixture corium and parts of vessel such as zircaloy and iron component. The requirements for solution of issues are below; -) the reasonable treatment process of the debris should be simple and in-situ in Fukushima Daiichi power plant, -) the desirable treatment process is to take out UO{sub 2} and PuO{sub 2} or metallic U and TRU metal, and dispose other fission products as high level radioactive waste; and -) the candidate of treatment process should generate the smallest secondary waste. Pyro-process has advantages to treat the debris because of the high solubility of the debris and its total process feasibility. Toshiba proposes a new pyro-process in molten salts using electrolysing Zr before debris fuel being treated.

  2. Water reactor fuel behaviour and fission products release in off-normal and accident conditions

    International Nuclear Information System (INIS)

    1987-09-01

    The present meeting was scheduled by the International Atomic Energy Agency upon the proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology and held at the IAEA Headquarters in Vienna from 10 to 13 November 1986. Thirty participants from 17 countries and an international organization attended the meeting. Eighteen papers were presented from 13 countries and one international organization. The meeting was composed of four sessions and covered subjects related to: physico-chemical properties of core materials under off-normal conditions, and their interactions up to and after melt-down (5 papers); core materials deformation, relocation and core coolability under (severe) accident conditions (4 papers); fission products release: including experience, mechanisms and modelling (5 papers); power plant experience (4 papers). A separate abstract was prepared for each of these 18 papers. Four working groups covering the above-mentioned topics were held to discuss the present status of the knowledge and to develop recommendations for future activities in this field. Refs, figs and tabs

  3. Phebus program main results and status for severe fuel damage studies

    International Nuclear Information System (INIS)

    Duco, J.; Reocreux, M.; Tattegrain, A.

    1986-06-01

    A large experimental in-pile program has been set up at the PHEBUS facility to investigate the actual behavior of .8 m active height, 25-rod PWR-type pressurized fresh fuel bundles under typical accident conditions. The program consists of four stages. Stage 1 was devoted to the adjustment of the operational procedure for stage 2. Stage 2 refers to the simulation of conservatively calculated L.B. LOCA 2 - peak transients. Stages 3/4 refer to four PWR severe accident scenarios retained for in-pile simulation at PHEBUS: a) a large break LOCA with injection failure; b) a small break LOCA associated with an injection failure; c) a prolonged total loss of the steam generator feedwater; and, d) a prolonged core uncovery a few days after reactor shutdown. The main PHEBUS stage 2 results are presented and finally interpreted

  4. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    Nakamura, Takehiko; Yoshinaga, Makio; Sobajima, Makoto; Fujishiro, Toshio; Horiki, Ohichiro; Yamahara, Takeshi; Ichihashi, Yoshinori; Kikuchi, Teruo

    1992-01-01

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  5. Theoretical analysis and numerical modelling of heat transfer and fuel migration in underlying soils and constructive elements of nuclear plants during an accident release from the core

    International Nuclear Information System (INIS)

    Arutunjan, R.V.; Bolshov, L.A.; Vitukov, V.V.; Goloviznin, V.M.; Dykhne, A.M.; Kiselev, V.P.; Klementova, S.V.; Krayushkin, I.E.; Moskovchenko, A.V.; Pismennii, V.D.; Popkov, A.G.; Chernov, S.Y.; Chudanov, V.V.; Khoruzhii, O.V.; Yudin, A.I.

    1990-01-01

    Migration of fuel fragments and core fission products during severe accidents on nuclear plants is studied analytically and numerically. The problems of heat transfer and migration of volume heat sources in construction materials and underlying soils are considered

  6. Full-length high-temperature severe fuel damage test No. 5

    International Nuclear Information System (INIS)

    Lanning, D.D.; Lombardo, N.J.; Hensley, W.K.; Fitzsimmons, D.E.; Panisko, F.E.; Hartwell, J.K.

    1993-09-01

    This report describes and presents data from a severe fuel damage test that was conducted in the National Research Universal (NRU) reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test, designated FLHT-5, was the fourth in a series of full-length high-temperature (FLHT) tests on light-water reactor fuel. The tests were designed and performed by staff from the US Department of Energy's Pacific Northwest Laboratory (PNL), operated by Battelle Memorial Institute. The test operation and test results are described in this report. The fuel bundle in the FLHT-5 experiment included 10 unirradiated full-length pressurized-water reactor (PWR) rods, 1 irradiated PWR rod and 1 dummy gamma thermometer. The fuel rods were subjected to a very low coolant flow while operating at low fission power. This caused coolant boilaway, rod dryout and overheating to temperatures above 2600 K, severe fuel rod damage, hydrogen generation, and fission product release. The test assembly and its effluent path were extensively instrumented to record temperatures, pressures, flow rates, hydrogen evolution, and fission product release during the boilaway/heatup transient. Post-test gamma scanning of the upper plenum indicated significant iodine and cesium release and deposition. Both stack gas activity and on-line gamma spectrometer data indicated significant (∼50%) release of noble fission gases. Post-test visual examination of one side of the fuel bundle revealed no massive relocation and flow blockage; however, rundown of molten cladding was evident

  7. Ex-core fuel damage event at paks causes, consequences and lessons learned

    International Nuclear Information System (INIS)

    Bajsz, J.; Gado, J.

    2004-01-01

    On April 10, 2003 Paks NPP experienced a loss of decay-heat removal to 30 irradiated fuel assemblies undergoing a cleaning process in a fuel service pit near the unit 2 spent fuel pool. Following chemical cleaning of high decay-heat fuel, a delay in removing the cleaning vessel's lid left the cleaning system in such a condition that did not provide adequate cooling to the fuel. After several hours of the fuel being under-cooled, a steam bubble developed in the vessel, essentially uncovering the fuel. When the lid of the vessel was removed, the sudden introduction of cool water thermally shocked the fuel causing significant structural damage and a release of fission product gases to the reactor building. The paper will discuss the causes of the event as well as the contributing factors to it. Detailed information will be given about the planning and preparation of the recovery actions. The in-depth analyses of the consequences and lessons learned complete the lecture. (author)

  8. Fixture and method for rectifying damaged guide thimble insert sleeves in a reconstitutable fuel assembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Ferlan, S.J.

    1987-01-01

    A guide thimble damage-rectifying method is described for use on a reconstitutable fuel assembly being held in a work station with its top nozzle removed to expose a plurality of guide thimbles having one of several different types of damage. The method consists of: (a) providing a base having a plurality of tool positioning openings defined therein in a pattern matched with that of the guide thimbles of the fuel assembly; (b) mounting the base on the work station with its tool positioning openings in alignment with the guide thimbles of the fuel assembly and such that the base is movable toward the guide thimbles; (c) providing a plurality of different tools each operable to rectify one of the different types of guide thimble damage; (d) mounting selected ones of the different tools in respective ones of the openings of the base in alignment with ones of the thimbles having the respective types of guide thimble damage capable of being rectified by the selected tools such that upon movement of the base toward the guide thimbles the respective types of guide thimble damage will be rectified by the selected tools; (e) providing a group of positioning elements; (f) mounting the positioning elements in selected ones of the base openings corresponding to undamaged ones of the guide thimbles such that upon movement of the base toward the guide thimbles the positioning elements become mounted on upper end portions of the corresponding undamaged ones of the guide thimbles for precisely locating the fixture relative to the guide thimble upper end portions for accurate performance of the repairable damage rectifying operation by the tools as the base is moved toward the guide thimbles; and (g) moving the base toward the guide thimbles so as to mount the positioning elements on the corresponding ones of the undamaged guide thimbles and effect rectification of the damaged guide thimbles by the selected tools

  9. Precursors to potential severe core damage accidents: 1995 A status report

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.

    1997-04-01

    Ten operational events that affected 10 commercial light-water reactors during 1995 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by first computer-screening the 1995 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969-1981 and 1984-1994 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  10. Precursors to potential severe core damage accidents: 1994, a status report. Volume 22: Appendix I

    International Nuclear Information System (INIS)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N.; Dolan, B.W.; Minarick, J.W.

    1995-12-01

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10 -6 . These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1

  11. Precursors to potential severe core damage accidents: 1995 A status report

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A. [and others

    1997-04-01

    Ten operational events that affected 10 commercial light-water reactors during 1995 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10{sup {minus}6}. These events were identified by first computer-screening the 1995 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969-1981 and 1984-1994 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events.

  12. Precursors to potential severe core damage accidents: 1994, a status report. Volume 22: Appendix I

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R.J.; Cletcher, J.W.; Copinger, D.A.; Vanden Heuvel, L.N. [Oak Ridge National Lab., TN (United States); Dolan, B.W.; Minarick, J.W. [Oak Ridge National Lab., TN (United States)]|[Science Applications International Corp., Oak Ridge, TN (United States)

    1995-12-01

    Nine operational events that affected eleven commercial light-water reactors (LWRs) during 1994 and that are considered to be precursors to potential severe core damage are described. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 {times} 10{sup {minus}6}. These events were identified by computer-screening the 1994 licensee event reports from commercial LWRs to identify those that could be potential precursors. Candidate precursors were then selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure that the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1981 and 1984--1993 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for events. This document is bound in two volumes: Vol. 21 contains the main report and Appendices A--H; Vol. 22 contains Appendix 1.

  13. Drying damaged K West fuel elements (Summary of whole element furnace runs 1 through 8); TOPICAL

    International Nuclear Information System (INIS)

    LAWRENCE, L.A.

    1998-01-01

    N Reactor fuel elements stored in the Hanford K Basins were subjected to high temperatures and vacuum conditions to remove water. Results of the first series of whole element furnace tests i.e., Runs 1 through 8 were collected in this summary report. The report focuses on the six tests with breached fuel from the K West Basin which ranged from a simple fracture at the approximate mid-point to severe damage with cladding breaches at the top and bottom ends with axial breaches and fuel loss. Results of the tests are summarized and compared for moisture released during cold vacuum drying, moisture remaining after drying, effects of drying on the fuel element condition, and hydrogen and fission product release

  14. Radiation doses in accidents at sea-transportation of spent fuel

    International Nuclear Information System (INIS)

    Appelgren, A.; Bergstroem, U.; Devell, L.

    1978-01-01

    In order to investigate the consequences of shipping accidents, a release of activity is assumed. This report presents the calculations of individual and collective doses from the two most severe postulated accidents which are given in a special accident analysis. One of the accidents is a ship collision together with fire on-board, the ship is floating after the collision and a certain quantity volatile fission products gives airborne activity. In the other case, it is a fire on-board, the ship will sink and cause a certain leakage to the sea

  15. Framatome-ANP France UO2 fuel fabrication. Criticality safety analysis in the light of the JCO accident

    International Nuclear Information System (INIS)

    Doucet, M.; Zheng, S.; Mouton, J.; Porte, R.

    2003-01-01

    In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The FRAMATOME-ANP production of its French low enriched (5 w/o) UO2 fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO2 blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach FRAMATOME-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (author)

  16. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-02-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

  17. Economic models of compensation for damages caused by nuclear accidents: some lessons for the revision of the Paris and Vienna Conventions

    International Nuclear Information System (INIS)

    Faure, Michael G.

    1995-01-01

    Alternative systems of compensation for damages caused by nuclear accidents have been proposed. In respect, the question merits attention to whether these alternative models of compensation discussed in the economic literature could be implemented when discussing the revision of the Paris and Vienna Conventions. 55 refs., 1 tab

  18. The loadings and strength of nuclear power plant structures in core damage accidents

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1994-01-01

    The reactor cavity of VVER-91 NPP is a thick-walled, cylindrical reinforced concrete structure. In case of molten core-water reaction during the severe reactor accident the load carrying capacity of the cavity structure is of interest against the short impulse type loading caused by the steam explosion phenomenon. The assumed size of the impulse was 20 kPa-s and the duration was 10 ms. This investigation was divided in several phases. First, the elastic response of the cavity was determined using the ABAQUS code. Next, the static response of the cavity was evaluated using elasto-plastic properties of reinforcement and concrete and also taking into account the cracking of the concrete. This analysis was done with the aid of ABAQUS/STANDARD and ANSYS codes and the obtained results agreed reasonably with each other. In order to obtain a qualitative picture of the behaviour of the structure under the impulse load a simplified single degree of freedorn model was developed. The hoop reinforcement of the cavity was taken as an elasto-plastic spring and the wall concrete acted as a mass. Using this model the suitable amount of hoop reinforcement was determined. In next phase, the dynamic analysis of the structure was attempted using elasto-plastic material properties and concrete cracking. (13 refs., 57 figs.)

  19. Phenomena occuring in the reactor coolant system during severe core damage accidents

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1990-01-01

    The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. This paper discusses, how in the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. The physical and chemical processes occurring within the RCS during normal operation of the reactor are relatively uncomplicated and are reasonably well understood. When the flow of coolant is properly adjusted, the thermal energy resulting from nuclear fission (or, in the shutdown mode, from radioactive decay processes) and secondary inputs, such as pumps, are exactly balanced by thermal losses through the RCS boundaries and to the various heat sinks that are employed to effect the conversion of heat to electrical energy. Because all of the heat and mass fluxes remain sensibly constant with time, mathematical descriptions of the thermophysical processes are relatively straightforward, even for boiling water reactor (BWR) systems. Although the coolant in a BWR does undergo phase changes, the phase boundaries remain well-defined and time-invariant

  20. Accident prevention and security measures to prevent intentional harm and damage through nuclear energy

    International Nuclear Information System (INIS)

    Rauschning, D.

    1984-01-01

    The author explains the authorities' duty to provide for protection against intentional damage or physical harm through the use of nuclear energy. It belongs to the competence of the various authorities to define ways and means to afford protection, and to establish an appropriate network of provisions. There are provisions belonging to criminal law, those concerning liability and indemnity, and security regulations incorporated in the law on licensing and supervision. The author presents a detailed account of the law applicable and discusses the conflict of interests between governmental duties and intentions and the civil rights of the individual affected by the provisions. (HSCH) [de

  1. Damaged Spent Nuclear Fuel at U.S. DOE Facilities Experience and Lessons Learned

    International Nuclear Information System (INIS)

    Brett W. Carlsen; Eric Woolstenhulme; Roger McCormack

    2005-01-01

    From a handling perspective, any spent nuclear fuel (SNF) that has lost its original technical and functional design capabilities with regard to handling and confinement can be considered as damaged. Some SNF was damaged as a result of experimental activities and destructive examinations; incidents during packaging, handling, and transportation; or degradation that has occurred during storage. Some SNF was mechanically destroyed to protect proprietary SNF designs. Examples of damage to the SNF include failed cladding, failed fuel meat, sectioned test specimens, partially reprocessed SNFs, over-heated elements, dismantled assemblies, and assemblies with lifting fixtures removed. In spite of the challenges involved with handling and storage of damaged SNF, the SNF has been safely handled and stored for many years at DOE storage facilities. This report summarizes a variety of challenges encountered at DOE facilities during interim storage and handling operations along with strategies and solutions that are planned or were implemented to ameliorate those challenges. A discussion of proposed paths forward for moving damaged and nondamaged SNF from interim storage to final disposition in the geologic repository is also presented

  2. Equipment for testing a group of nuclear reactor fuel elements for damage to the cans

    International Nuclear Information System (INIS)

    Mohm, F.

    1977-01-01

    Equipment is described for use in sodium cooled nuclear reactors, with which the fuel elements consisting of bundles of fuel and fertile rods can be examined for damage to the cans. Fission poducts occurring in the liquid coolant act as indicators. The coolant is sucked via pipelines which penetrate into the elements into a collecting container, and a special pipeline is available for every element of a group, where the highest points of individual pipelines at different hydrostatic heads are taken to the collecting container. This permits the checking of one line at a time due to pressure changes. (UWI) [de

  3. Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

    International Nuclear Information System (INIS)

    Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

    1978-09-01

    In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

  4. Behavior of U3Si2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

    International Nuclear Information System (INIS)

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso; Pizzocri, Davide; Pastore, Giovanni

    2016-01-01

    As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.

  5. Theoretical Derivation of Simplified Evaluation Models for the First Peak of a Criticality Accident in Nuclear Fuel Solution

    International Nuclear Information System (INIS)

    Nomura, Yasushi

    2000-01-01

    In a reprocessing facility where nuclear fuel solutions are processed, one could observe a series of power peaks, with the highest peak right after a criticality accident. The criticality alarm system (CAS) is designed to detect the first power peak and warn workers near the reacting material by sounding alarms immediately. Consequently, exposure of the workers would be minimized by an immediate and effective evacuation. Therefore, in the design and installation of a CAS, it is necessary to estimate the magnitude of the first power peak and to set up the threshold point where the CAS initiates the alarm. Furthermore, it is necessary to estimate the level of potential exposure of workers in the case of accidents so as to decide the appropriateness of installing a CAS for a given compartment.A simplified evaluation model to estimate the minimum scale of the first power peak during a criticality accident is derived by theoretical considerations only for use in the design of a CAS to set up the threshold point triggering the alarm signal. Another simplified evaluation model is derived in the same way to estimate the maximum scale of the first power peak for use in judging the appropriateness for installing a CAS. Both models are shown to have adequate margin in predicting the minimum and maximum scale of criticality accidents by comparing their results with French CRiticality occurring ACcidentally (CRAC) experimental data

  6. Criticality accident in uranium fuel processing plant. The estimation of the total number of fissions with related reactor physics parameters

    International Nuclear Information System (INIS)

    Nishina, Kojiro; Oyamatsu, Kazuhiro; Kondo, Shunsuke; Sekimoto, Hiroshi; Ishitani, Kazuki; Yamane, Yoshihiro; Miyoshi, Yoshinori

    2000-01-01

    This accident occurred when workers were pouring a uranium solution into a precipitation tank with handy operation against the established procedure and both the cylindrical diameter and the total mass exceeded the limited values. As a result, nuclear fission chain reactor in the solution reached not only a 'criticality' state continuing it independently but also an instantly forming criticality state exceed the criticality and increasing further nuclear fission number. The place occurring the accident at this time was not reactor but a place having not to form 'criticality' called by a processing process of uranium fuel. In such place, as because of relating to mechanism of chain reaction, it is required naturally for knowledge on the reactor physics, it is also necessary to understand chemical reaction in chemical process, and functions of tanks, valves and pumps mounted at the processes. For this purpose, some information on uranium concentration ratio, atomic density of nuclides largely affecting to chain reaction such as uranium, hydrogen, and so forth in the solution, shape, inner structure and size of container for the solution, and its temperature and total volume, were necessary for determining criticality volume of the accident uranium solution by using nuclear physics procedures. Here were described on estimation of energy emission in the JCO accident, estimation from analytical results on neutron and solution, calculation of various nuclear physics property estimation on the JCO precipitation tank at JAERI. (G.K.)

  7. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source reactor at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

    1995-01-01

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at ORNL. Damage propagation is postulated to occur from thermal conduction between dmaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur beause of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A parametric study was done for several uncertain variables. The study included investigating effects of plate contact area, convective heat transfer coefficient, thermal conductivity on fuel swelling, and initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects of damage propagation. Results provide useful insights into how variouss uncertain parameters affect damage propagation

  8. Proceedings of the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 23-25 September 2014, OECD-NEA HQ

    International Nuclear Information System (INIS)

    Massara, S.; ); Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Yang, Jae Ho; Dolley, Evan J.; Rebak, Raul B.; Sowder, Andrew; Cheng, Bo; Kurata, Masaki; Van Nieuwenhove, Rudi; Li, R.; McClellan, Ken; Nelson, Andy; Carmack, Jon; Harp, Jason; Finck, Phillip; ); Kakicuhi, K.

    2014-09-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Proposed Agenda; 2 - Expert Group meeting - 23 September 2014: - Introduction and background (S. Massara, OECD-NEA) - Expected outcomes from the TFs meetings scheduled on 24-25 September (K. Pasamehmetoglu, EG Chair, INL); 3 - Task Force 1 (Systems assessment) meeting - 24 September 2014: - Metrics for the Evaluation of LWR Accident Tolerant Fuel (S. Bragg-Sitton, INL); 4 - Task Force 2 (Cladding/core materials) meeting - 24 September 2014: - Summary on SiC Task Force 2 (Clad) meeting (J.H. Yang, KAERI); - Accident Tolerant Advanced Steels Cladding for Commercial Light Water Reactors (E. Dolley, GE); - Molybdenum-Alloy Fuel Cladding Development and Testing - Update from April 2014 NEA ATF Meeting (A. Sowder, EPRI); - Accident Tolerant Control Rod Development in Japan (M. Kurata, JAEA); - IFA-774: The first in-pile test with coated fuel rods (R. Van

  9. Materials properties utilization in a cumulative mechanical damage function for LMFBR fuel pin failure analysis

    International Nuclear Information System (INIS)

    Jacobs, D.C.

    1977-01-01

    An overview is presented of one of the fuel-pin analysis techniques used in the CRBRP program, the cumulative mechanical damage function. This technique, as applied to LMFBR's, was developed along with the majority of models used to describe the mechanical properties and environmental behavior of the cladding (i.e., 20 percent cold-worked, 316 stainless steel). As it relates to fuel-pin analyses the Cumulative Mechanical Damage Function (CDF) continually monitors cladding integrity through steady state and transient operation; it is a time dependent function of temperature and stress which reflects the effects of both the prior mechanical history and the variations in mechanical properties caused by exposure to the reactor environment

  10. Efficient prevention and compensation of catastrophic risks. The example of damage by nuclear accidents

    International Nuclear Information System (INIS)

    Vanden Borre, T.

    2001-01-01

    This book deals with the liability for damage due to catastrophic risks. The nuclear liability law serves as an example of such a catastrophic risk. The question that we tried to answer is what an efficient compensation scheme for catastrophic risks should look like. This question is dealt with both from a law and an economic point of view and from a comparative point of view. The main element in comparing the laws in different countries is the comparison between Belgian and Dutch civil (nuclear) liability law. But also American nuclear liability law is part of the analysis (the Price-Anderson Act). The book consists of four parts: (nuclear) civil liability law, legal and economic approach, analysis of other compensation systems and conclusions. The big themes in this book are therefore civil (nuclear) liability law, insurance law and environmental liability law [nl

  11. Simulation of accident-tolerant U{sub 3}Si{sub 2} fuel using FRAPCON code

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R., E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Universidade de São Paulo (USP), São Paulo, SP (Brazil). Departamento de Engenharia Naval e Oceânica

    2017-07-01

    The research on accident-tolerant fuels (ATFs) increased after the Fukushima event. This benefited risk management in nuclear operations. In this investigation, the physical properties of the materials being developed for the ATF program were compared with those of the standard UO{sub 2} - Zr fuel system. The research efforts in innovative fuel design include rigorous characterization of thermal, mechanical, and chemical assessment, with the objectives of making the burnup cycle longer, increasing power density, and improving safety performance. Fuels must reach a high uranium density - above that supported by UO{sub 2} - and possess coating that exhibits better oxidation resistance than Zircaloy. The uranium density and thermal conductivity of ATFs, such as U{sub 3}Si{sub 2}, UN, and UC, is higher than that of UO{sub 2}; their combination with advanced cladding provides possible fuel - cladding options. An ideal combination of fuel and cladding must increase fuel performance in loss-of-coolant scenarios. The disadvantages of U{sub 3}Si{sub 2}, UN, and UC are their swelling rates, which are higher than that of UO{sub 2}. The thermal conductivities of ATFs are approximately four times higher than that of UO2. To prevent the generation of hydrogen due to oxidation of zirconium-based alloys in contact with steam, cladding options, such as ferritic alloys, were studied. It was verified that FeCrAl alloys and SiC provide better response under severe conditions because of their thermophysical properties. The findings of this study indicate that U{sub 3}Si{sub 2} and the FeCrAl fuel cladding concept should replace UO{sub 2} - Zr as the fuel system of choice. (author)

  12. Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

    Directory of Open Access Journals (Sweden)

    Isabelle Guénot-Delahaie

    2018-03-01

    Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

  13. Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

    International Nuclear Information System (INIS)

    Fukano, Yoshitaka; Kurisaka, Kenichi; Nishimura, Masahiro; Naruto, Kenichi

    2015-01-01

    Experimental studies, deterministic approaches and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from, AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was, therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated transient without scram or protected loss of heat sink events from both the viewpoint of occurrence probability and consequences. (author)

  14. Diagnosing of car engine fuel injectors damage using DWT analysis and PNN neural networks

    Directory of Open Access Journals (Sweden)

    Piotr CZECH

    2013-01-01

    Full Text Available In many research centers all over the world nowadays works are being carried out aimed at compiling method for diagnosis machines technical condition. Special meaning have non-invasive methods including methods using vibroacoustic phenomena. In this article is proposed using DWT analysis and energy or entropy, which are a base for diagnostic system of fuel injectors damage in car combustion engine. There were conducted researches aimed at building of diagnostic system using PNN neural networks.

  15. Drilling-induced borehole-wall damage at spent fuel test-climax

    International Nuclear Information System (INIS)

    Weed, H.C.; Durham, W.B.

    1982-12-01

    Microcracks in a sample of quartz monzonite from the Spent Fuel Test-Climax were measured by means of a scanning electron microscope in order to estimate the background level of damage near the borehole-wall. It appears that the hammer-drilling operation used to create the borehole has caused some microfracturing in a region 10 to 30 mm wide around the borehole. Beyond 30 mm, the level of microfracturing cannot be distinguished from background

  16. Program requirements to determine and relate fuel damage and failure thresholds to anticipated conditions in pressurized water reactors

    International Nuclear Information System (INIS)

    Loyd, R.F.; Croucher, D.W.

    1980-03-01

    Anticipated transients, licensing criteria, and damage mechanisms for PWR fuel rods are reviewed. Potential mechanistic fuel rod damage limits for PWRs are discussed. An expermental program to be conducted out-of-pile and in the Engineering Test Reactor (ETR) to generate a safety data base to define mechanistic fuel damage and failure thresholds and to relate these thresholds to the thermal-hydraulic and power conditions in a PWR is proposed. The requirements for performing the tests are outlined. Analytical support requirements are defined

  17. Estimation of irradiation-induced material damage measure of FCM fuel in LWR core

    International Nuclear Information System (INIS)

    Lee, Kyung-Hoon; Lee, Chungchan; Park, Sang-Yoon; Cho, Jin-Young; Chang, Jonghwa; Lee, Won Jae

    2014-01-01

    An irradiation-induced material damage measure on tri-isotropic (TRISO) multi-coating layers of fully ceramic micro-encapsulated (FCM) fuel to replace conventional uranium dioxide (UO 2 ) fuel for existing light water reactors (LWRs) has been estimated using a displacement per atom (DPA) cross section for a FCM fuel performance analysis. The DPA cross sections in 47 and 190 energy groups for both silicon carbide (SiC) and graphite are generated based on the molecular dynamics simulation by SRIM/TRIM. For the selected FCM fuel assembly design with FeCrAl cladding, a core depletion analysis was carried out using the DeCART2D/MASTER code system with the prepared DPA cross sections to evaluate the irradiation effect in the Korean OPR-1000. The DPA of the SiC and IPyC coating layers is estimated by comparing the discharge burnup obtained from the MASTER calculation with the burnup-dependent DPA for each coating layer calculated using DeCART2D. The results show that low uranium loading and hardened neutron spectrum compared to that of high temperature gas-cooled reactor (HTGR) result in high discharge burnup and high fast neutron fluence. In conclusion, it can be seen that the irradiation-induced material damage measure is noticeably increased under LWR operating conditions compared to HTGRs. (author)

  18. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  19. ORAL MUCOSA DAMAGE BECAUSE OF HYPOCHLORITE ACCIDENT – A CASE REPORT AND LITERATURE REVIEW

    Directory of Open Access Journals (Sweden)

    Elitsa Deliverska

    2016-08-01

    Full Text Available Background Hypochlorite solution is widely used in dental practice during root canal treatment. Although it is generally regarded as being very safe, potentially severe complications can occur when it comes into contact with soft tissue especially due to its cytotoxic features. Objective The aim of our paper is to present a case of damage of oral mucosa because of leakage of 3% hypochlorite through rubber dam during endodontic treatment. Material and methods We present a 31 years old female with necrosis of buccal mucosa during the endodontic treatment of 46. Results Three days after the procedure the patient was referred to our department for consultation and treatment. Antiseptic lavage was performed and oral antibiotic was administrated. After 5 days intraoral examination showed signs of almost full recovery. Conclusion The need for proper tooth isolation during restorative procedures is obvious. Anything that obscures the operative field negatively impacts operator efficiency and effectiveness. Visibility, patient/operator safety, infection control and the physical properties of dental materials are all compromised when proper isolation is lacking.

  20. Prediction of failure enthalpy and reliability of irradiated fuel rod under reactivity-initiated accidents by means of statistical approach

    International Nuclear Information System (INIS)

    Nam, Cheol; Choi, Byeong Kwon; Jeong, Yong Hwan; Jung, Youn Ho

    2001-01-01

    During the last decade, the failure behavior of high-burnup fuel rods under RIA has been an extensive concern since observations of fuel rod failures at low enthalpy. Of great importance is placed on failure prediction of fuel rod in the point of licensing criteria and safety in extending burnup achievement. To address the issue, a statistics-based methodology is introduced to predict failure probability of irradiated fuel rods. Based on RIA simulation results in literature, a failure enthalpy correlation for irradiated fuel rod is constructed as a function of oxide thickness, fuel burnup, and pulse width. From the failure enthalpy correlation, a single damage parameter, equivalent enthalpy, is defined to reflect the effects of the three primary factors as well as peak fuel enthalpy. Moreover, the failure distribution function with equivalent enthalpy is derived, applying a two-parameter Weibull statistical model. Using these equations, the sensitivity analysis is carried out to estimate the effects of burnup, corrosion, peak fuel enthalpy, pulse width and cladding materials used

  1. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    , Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO 2 ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO 2 and MOX ceramics (Chromium oxide-doped UO 2 fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The nature of spent nuclear

  2. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    irradiation, Bubbles and precipitates, Modeling fuel behavior); Modeling defects and fission products in UO{sub 2} ceramic by ab initio computation (Ab initio computation, Point defects in uranium dioxide, Fission products in uranium dioxide, The indispensable coupling of modeling and experiment); Cladding and assembly materials (What is the purpose of cladding?, Zirconium alloys, Claddings: required to exhibit good mechanical strength, Mechanical behavior of irradiated Zr alloys, Claddings: required to prove corrosion resistant); Pellet-cladding interaction (The phenomena involved in pellet-cladding interaction (PCI), Experimental simulation of PCI and the lessons to be drawn from it, The requirement for an experimental basis, Numerical simulation of PCI, Towards a lifting of PCI-related operating constraints); Advanced UO{sub 2} and MOX ceramics (Chromium oxide-doped UO{sub 2} fuel, Novel MOX microstructures); Mechanical behavior of fuel assemblies (Assembly mechanical behavior in normal operating conditions, Assembly mechanical behavior in accident situations, Fuel in a loss of primary coolant accident (LOCA)); Introduction to LOCA-type accident transients (Overview of thermal-hydraulic and fuel-related aspects, Incidence of LOCA transients on the thermal-metallurgical-mechanical behavior of zirconium-base alloy cladding); Fuel in a reactivity insertion accident (RIA) (Safety criteria); Fuel in a severe accident (The VERCORS analytical program, The Phebus-FP global tests, Control of severe accidents in the EPR reactor); In-core fuel management (Relationships between cycle length, maximum burnup, and batch fraction Enrichment and burnable poisons, The impact of the nature of the fuel used, and its evolution, on the major parameters of core physics, and management Prospects for future trends in core management); Fuel cycle material balances (In-core evolution of materials, Decay heat and potential radiotoxicity, Plutonium management); Long-term behavior of spent fuel (The

  3. Criticality accident in uranium fuel processing plant. Emergency medical care and dose estimation for the severely overexposed patients

    Energy Technology Data Exchange (ETDEWEB)

    Akashi, Makoto; Ishigure, Nobuhito [National Inst. of Radiological Sciences, Chiba (Japan)

    2000-08-01

    A criticality accident occurred in JCO, a plant for nuclear fuel production in 1999 and three workers were exposed to extremely high-level radiation (neutron and {gamma}-ray). This report describes outlines of the clinical courses and the medical cares for the patients of this accident and the emergent medical system for radiation accident in Japan. One (A) of the three workers of JCO had vomiting and diarrhea within several minutes after the accident and another one (B) had also vomiting within one hour after. Based on these evidences, the exposure dose of A and B were estimated to be more than 8 and 4 GyEq, respectively. Generally, acute radiation syndrome (ARS) is assigned into three phases; prodromal phase, critical or manifestation phase and recovery phase or death. In the prodromal phase, anorexia, nausea, vomiting and diarrhea often develop, whereas the second phase is asymptotic. In the third phase, various syndromes including infection, hemorrhage, dehydration shock and neurotic syndromes are apt to occur. It is known that radiation exposure at 1 Gy or more might induce such acute radiation syndromes. Based on the clinical findings of Chernobyl accident, it has been thought that exposure at 0.5 Gy or more causes a lowering of lymphocyte level and a decrease in immunological activities within 48 hours. Lymphocyte count is available as an indicator for the evaluation of exposure dose in early phase, but not in later phase The three workers of JCO underwent chemical analysis of blood components, chromosomal analysis and analysis of blood {sup 24}Na immediately after the arrival at National Institute of Radiological Sciences via National Mito Hospital specified as the third and the second facility for the emergency medical care system in Japan, respectively. (M.N.)

  4. Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO2 or ZrO2. The new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.

  5. A summary of the assessment of fuel behaviour, fission product release and pressure tube integrity following a postulated large loss-of-coolant accident

    International Nuclear Information System (INIS)

    Langman, V.J.; Weaver, K.R.

    1984-05-01

    The Ontario Hydro analyses of fuel and pressure tube temperatures, fuel behaviour, fission product release and pressure tube integrity for large break loss-of-coolant accidents in Bruce A or Pickering A have been critically reviewed. The determinations of maximum fuel temperatures and fission product release are very uncertain, and pressure tube integrity cannot be assured where low steam flows are predicted to persist for times on the order of minutes

  6. Workbench experiments on interaction of nuclear fuel with channel reactor materials: the LFCM congestions criticality and accident scenario in both re-examinations

    International Nuclear Information System (INIS)

    Baryakhtar, V.; Gonchar, V.; Zhidkov, A.

    2002-01-01

    The heavy radioecological consequences of 1986 accident were mainly stipulated by destruction of both a significant part of fuel envelopes and fuel matrix due to high-temperature interaction with silicates, when nuclear fuel has lost an important property to keep the fission products inside its volume. In this respect the silicate material application in thermal-insulating filling is a crucial fault had been made when Chornobyl NPP channel reactor designing

  7. Engineered zircaloy cladding modifications for improved accident tolerance of LWR fuel: US DOE NEUP Integrated Research Project

    International Nuclear Information System (INIS)

    Heuser, Brent

    2013-01-01

    An integrated research project (IRP) to fabricate and evaluate modified zircaloy LWR cladding under normal BWR/PWR operation and off-normal events has been funded by the US DOE. The IRP involves three US academic institutions, a US national laboratory, an intermediate stock industrial cladding supplier, and an international academic institution. A combination of computational and experimental protocols will be employed to design and test modified zircaloy cladding with respect to corrosion and accelerated oxide growth, the former associated with normal operation, the latter associated with steam exposure during loss of coolant accidents (LOCAs) and low-pressure core re-floods. Efforts will be made to go beyond design-base accident (DBA) scenarios (cladding temperature equal to or less than 1204 deg. C) during the experimental phase of modified zircaloy performance characterisation. The project anticipates the use of the facilities at ORNL to achieve steam exposure beyond DBA scenarios. In addition, irradiation of down-selected modified cladding candidates in the ATR may be performed. Cladding performance evaluation will be incorporated into a reactor system modelling effort of fuel performance, neutronics, and thermal hydraulics, thereby providing a holistic approach to accident-tolerant nuclear fuel. The proposed IRP brings together personnel, facilities, and capabilities across a wide range of technical areas relevant to the study of modified nuclear fuel and LWR performance during normal operation and off-normal scenarios. Two pathways towards accident-tolerant LWR fuel are envisioned, both based on the modification of existing zircaloy cladding. The first is the modification of the cladding surface by the application of a coating layer designed to shift the M + O→MO reaction away from oxide growth during steam exposure at elevated temperatures. This pathway is referred to as the 'surface coating' solution. The second is the modification of the bulk

  8. Methods and data for HTGR fuel performance and radionuclide release modeling during normal operation and accidents for safety analysis

    International Nuclear Information System (INIS)

    Verfondern, K.; Martin, R.C.; Moormann, R.

    1993-01-01

    The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of knowledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying database for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed. (orig.) [de

  9. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    International Nuclear Information System (INIS)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000 degree F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion (''bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled

  10. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N.J.; Marseille, T.J.; White, M.D.; Lowery, P.S.

    1990-06-01

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000{degree}F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion ( bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled.

  11. Frost induced damages within porous materials - from concrete technology to fuel cells technique

    Science.gov (United States)

    Palecki, Susanne; Gorelkov, Stanislav; Wartmann, Jens; Heinzel, Angelika

    2017-12-01

    Porous media like concrete or layers of membrane electrode assemblies (MEA) within fuel cells are affected by a cyclic frost exposure due to different damage mechanisms which could lead to essential degradation of the material. In general, frost damages can only occur in case of a specific material moisture content. In fuel cells, residual water is generally available after shut down inside the membrane i.e. the gas diffusion layer (GDL). During subsequent freezing, this could cause various damage phenomena such as frost heaves and delamination effects of the membrane electrode assembly, which depends on the location of pore water and on the pore structure itself. Porous materials possess a pore structure that could range over several orders of magnitudes with different properties and freezing behaviour of the pore water. Latter can be divided into macroscopic, structured and pre-structured water, influenced by surface interactions. Therefore below 0 °C different water modifications can coexist in a wide temperature range, so that during frost exposure a high amount of unfrozen and moveable water inside the pore system is still available. This induces transport mechanisms and shrinkage effects. The physical basics are similar for porous media. While the freezing behaviour of concrete has been studied over decades of years, in order to enhance the durability, the know-how about the influence of a frost attack on fuel cell systems is not fully understood to date. On the basis of frost damage models for concrete structures, an approach to describe the impact of cyclic freezing and thawing on membrane electrode assemblies has been developed within this research work. Major aim is beyond a better understanding of the frost induced mechanisms, the standardization of a suitable test procedure for the assessment of different MEA materials under such kind of attack. Within this contribution first results will be introduced.

  12. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  13. Fuel containment and damage tolerance in large composite primary aircraft structures. Phase 2: Testing

    Science.gov (United States)

    Sandifer, J. P.; Denny, A.; Wood, M. A.

    1985-01-01

    Technical issues associated with fuel containment and damage tolerance of composite wing structures for transport aircraft were investigated. Material evaluation tests were conducted on two toughened resin composites: Celion/HX1504 and Celion/5245. These consisted of impact, tension, compression, edge delamination, and double cantilever beam tests. Another test series was conducted on graphite/epoxy box beams simulating a wing cover to spar cap joint configuration of a pressurized fuel tank. These tests evaluated the effectiveness of sealing methods with various fastener types and spacings under fatigue loading and with pressurized fuel. Another test series evaluated the ability of the selected coatings, film, and materials to prevent fuel leakage through 32-ply AS4/2220-1 laminates at various impact energy levels. To verify the structural integrity of the technology demonstration article structural details, tests were conducted on blade stiffened panels and sections. Compression tests were performed on undamaged and impacted stiffened AS4/2220-1 panels and smaller element tests to evaluate stiffener pull-off, side load and failsafe properties. Compression tests were also performed on panels subjected to Zone 2 lightning strikes. All of these data were integrated into a demonstration article representing a moderately loaded area of a transport wing. This test combined lightning strike, pressurized fuel, impact, impact repair, fatigue and residual strength.

  14. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    International Nuclear Information System (INIS)

    Schanz, G.; Hagen, S.; Hofmann, P.; Sepold, L.; Schumacher, G.

    1992-01-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400deg C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4 C), which are leading to extensive low-temperature melt formation around 1200deg C. Interrelations between those basic phenomena, resulting for example in cladding deformation ('flowering') and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ('quenching') are determining the evolution paths of fuel element destruction, which are to be identified. (orig.)

  15. Fission product behavior during the first two PBF severe fuel damage tests

    International Nuclear Information System (INIS)

    Osetek, D.J.; Cronenberg, A.W.; Hobbins, R.R.; Vinjamuri, K.

    1984-01-01

    The results of the first two severe fuel damage tests performed in the Power Burst Facility are assessed in terms of fission product release and chemical behavior. On-line gamma spectroscopy and grab sample data indicate limited release during solid-phase fuel heatup. Analysis indicates that the fuel morphology conditions for the trace-irradiated fuel employed in these two tests limit initial release. Only upon high temperature fuel restructuring and liquefaction is significant release indicated. Chemical equilibrium predictions, based on steam oxidation or reduction conditions, indicate I to be the primary iodine species during trnsport in the steam environment of the first test and CsI to be the primary species during transport in the hydrogen environment of the second test. However, the higher steam flow rate conditions of the first test transported the released iodine through the sample system; whereas, low-hydrogen flow rate of the second test apparently allowed the vast majority of iodine-bearing compounds to plateout during transport

  16. Proceedings of the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 3-5 March 2015, OECD-NEA HQ

    International Nuclear Information System (INIS)

    Bischoff, Jeremy; Gandrille, Pascal; Forgeron, Thierry; Brachet, Jean-Christophe; Lorrette, Christophe; Valot, C.; Freyss, M.; Braun, J.; Sauder, C.; Moatti, Marie; Waeckel, Nicolas; Ambard, Antoine; Pasamehmetoglu, Kemal; Johnston, Emma; Bragg-Sitton, Shannon M.; Kurata, M.; Hallstadius, Lars; Ohta, H.; Ogata, T.; Besmann, T.; Chauvin, Nathalie; Cornet, Stephanie; Massara, S.; Kohyama, Akira; Kishimoto, Hirotatsu; Park, Joon Soo; Nakazato, Naofumi; Hayasaka, Daisuke; Asakura, Yuuki; Kanda, Chisato; Kohyama, Akira; Terrani, Kurt; Katoh, Yutai; Yamamoto, Yukinori; Field, Kevin; Snead, Lance; Hu, Xunxiang; Dryepondt, Sebastien; Unocic, Kinga A.; Hoelzer, David T.; Pint, Bruce A.; Besmann, T.; Steinbrueck, M.; Grosse, M.; Jianu, A.; Weisenburger, A.; Avincola, V.; Ahmad, S.; Tang, C.; Heuser, Brent J.; Sickafus, Kurt; Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Park, Jeong-Yong; Koo, Yang-Hyun; Lee, B.O.; Van Nieuwenhove, Rudi; Kim, Young; Rebak, Raul; Dolly, Evan; Dolley, E.J.; Rebak, R.B.; Maloy, Stu; Yang, Jae-Ho; Kim, Dong-Joo; Kim, Keon-Sik; Koo, Yang-Hyun; Lee, Won Jae; Tulenko, James S.; Puide, Mattias; Liu, T.; Gueneau, C.; Gosse, S.; Dupin, N.; Barber, D.; Corcoran, E.; Dumas, J.C.; Hania, R.; Kaye, M.; Turchi, P.

    2015-03-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Task Force 1 (Systems assessment) meeting, 3-4 March 2015: - French evaluation of ATF Concepts (J. Bischoff, AREVA); - Technology Readiness Levels - TRL - for Fuels (K. Pasamehmetoglu, INL); - TRL-definition for advanced fuel concept applied for commercial LWRs in Japan (M. Kurata, JAEA); - Application of TRLs in NNL (E. Johnston, NNL); - Technology Readiness Levels for Advanced Nuclear Fuel and Materials (S. Bragg-Sitton, INL); 1a - Definition of the illustrative scenarios: - AREVA's proposal concerning scenario for Accident Tolerant Fuel studies (P. Gandrille); - A Simplified Accident Scenario (L. Hallstadius); - Accident Scenarios for ATF Performance Evaluation of BWR and PWR in Japan (H. Ohta, CRIEPI); 1b - Related NEA activities: - Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear Systems - WPMM, Expert

  17. The buckling of fuel rods in transportation casks under hypothetical accident conditions

    International Nuclear Information System (INIS)

    Bjorkman, G.S.

    2004-01-01

    The buckling analysis of fuel rods during an end drop impact of a spent fuel transportation cask has traditionally been performed to demonstrate the structural integrity of the fuel rod cladding or the integrity of the fuel geometry in criticality evaluations following a cask drop event. The actual calculation of the fuel rod buckling load, however, has been the subject of some controversy, with estimates of the critical buckling load differing by as much as a factor of 5. Typically, in the buckling analysis of a fuel rod, assumptions are made regarding the percentage of fuel mass that is bonded to or participates with the cladding during the buckling process, with estimates ranging from 0 to 100%. The greater the percentage of fuel mass that is assumed to be bonded to the cladding the higher the inertia loads on the cladding, and, therefore, the lower the ''g'' value at which buckling occurs. Current published solutions do not consider displacement compatibility between the fuel and the cladding. By invoking displacement compatibility between the fuel column and the cladding column, this paper presents an exact solution for the buckling of fuel rods under inertia loading. The results show that the critical inertia load magnitude for the buckling of a fuel rod depends on the weight of the cladding and the total weight of the fuel, regardless of the percentage of fuel mass that is assumed to be attached to or participate with the cladding in the buckling process. Therefore, 100% of the fuel always participates in the buckling of a fuel rod under inertia loading

  18. Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 3000C

    International Nuclear Information System (INIS)

    Pandarinathan, P.R.; Vasudevan, P.

    1980-01-01

    Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300 0 C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300 0 C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300 0 C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor. (orig.)

  19. Climate change adaptation, damages and fossil fuel dependence. An RETD position paper on the costs of inaction

    Energy Technology Data Exchange (ETDEWEB)

    Katofsky, Ryan; Stanberry, Matt; Hagenstad, Marca; Frantzis, Lisa

    2011-07-15

    The Renewable Energy Technology Deployment (RETD) agreement initiated this project to advance the understanding of the ''Costs of Inaction'', i.e. the costs of climate change adaptation, damages and fossil fuel dependence. A quantitative estimate was developed as well as a better understanding of the knowledge gaps and research needs. The project also included some conceptual work on how to better integrate the analyses of mitigation, adaptation, damages and fossil fuel dependence in energy scenario modelling.

  20. NARCISS critical stand experiments for studying the nuclear safety in accident water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Glushkov, E.S.; Bubelev, V.G.

    2005-01-01

    A brief description of the Topaz-2 SNPS designed under scientific supervision of RRC KI in Russia, and of the NARCISS critical facility, is given. At the NARCISS critical facility, neutronic peculiarities and nuclear safety issues of the Topaz-2 system reactor were studied experimentally. This work is devoted to a detailed description of experiments on investigation of criticality safety in accident water immersion og highly enriched uranium dioxide fuel elements, performed at the NARCISS facility. The experiments were carried out at water-moderated critical assemblies with varying height, number, and spacing of fuel elements. The results obtained in the critical experiments, computational models of the investigated critical configurations, and comparison of the computational and experimental results are given [ru

  1. An assessment of fuel freezing and drainage phenomena in a reactor shield plug following a core disruptive accident

    International Nuclear Information System (INIS)

    El-Genk, M.; Cronenberg, A.W.

    1978-01-01

    An important problem related to the assessment of the recriticality potential for an LMFBR following a core disruptive accident is an understanding of the freezing phenomena of molten fuel on a cold structure which may prevent fuel dispersal and sunsequent shutdown. Transient analytical freezing and drainage calculations have been applied to molten UO 2 travel through the rather cold lower shield plug of the Clinch River Breeder Reactor (CRBR). The successive approximation technique is used to obtain a solution of the non-linear freezing problem, where such effects as heat generation, viscous heat dissipation, temperature dependent thermophysical properties and a convective boundary condition at the solidification front have been incorporated into the present analytical formulation. Results indicate that previous steady-state analysis overestimate the rate of frozen layer build-up by about a factor of two. However, of primary importance is the driving force for drainage and the diameter of the shield plug flow channel. (Auth.)

  2. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-01-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

  3. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-12-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

  4. Experience in construction of new safe confinement over ChNPP-4 damages in beyond design-basis accident: view in 30 year

    International Nuclear Information System (INIS)

    Nosovskij, A.V.

    2016-01-01

    Many organizations and institutions participated in the elimination of ChNPP-4 accident. However, the main efforts on conservation of the damaged unit and ChNPP-3 commissioning were performed in 1986-1987 by experts of the enterprises and organizations of the Ministry of Medium Machine Building of the Soviet Union, who have been sent to the staff of the specially created Construction Administration US-605. The paper presents the activity of US-605 experts, describes administrative and technical measures of radiation safety during construction of the Shelter in difficult radiation conditions. Such efforts enabled accumulation of significant and unique experience in mitigation of severe accident consequences, which shall be used to prevent any nuclear accidents and eliminate their consequences

  5. Thermal analyses for the rack design with spent fuel pool during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Yeh, C-L.; Chen, Y-S.; Chen, B-Y., E-mail: clinyeh@iner.gov.tw, E-mail: yschen@iner.gov.tw, E-mail: onepicemine@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China); Tseng, Y-S., E-mail: ystseng@mx.nthu.edu.tw [National Tsing Hua Univ., Engineering and System Science, Hsinchu, Taiwan (China); Wei, W-C., E-mail: hn150456@iner.gov.tw [Inst. of Nuclear Energy Research, Taoyuan County, Taiwan (China)

    2014-07-01

    Alternative fuel arrangements separating the latest fuels discharge from the reactor core are proposed, such as the 1x4 configuration in which the hot assembly is surrounded by 4 assemblies with much lower decay heat. For the rack design in the BWR spent fuel pool design, the lateral flow is eliminated by solid walls. In this study, cooling enhancement of splitting fuel rack is investigated using Computational Fluid Dynamics (CFD). The fuels in the pool are modeled by porous medium. Separating the fuel rack by a distance of 10 cm can lower the peak cladding temperature and the natural convection between the fuels and then earns more response time for the site people to implement necessary mitigation actions. (author)

  6. Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

    International Nuclear Information System (INIS)

    Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

    1991-09-01

    The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

  7. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun, E-mail: pdj@kaeri.re.kr; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-15

    This study investigates protective coatings for improving the high temperature oxidation resistance of Zr fuel claddings for light water nuclear reactors. FeCrAl alloy and Cr layers were deposited onto Zr plates and tubes using cold spraying. For the FeCrAl/Zr system, a Mo layer was introduced between the FeCrAl coating and the Zr matrix to prevent inter-diffusion at high temperatures. Both the FeCrAl and Cr coatings improved the oxidation resistance compared to that of the uncoated Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated cladding samples were studied under simulated loss-of-coolant accident conditions. The coated samples showed higher burst temperatures, lower circumferential strain, and smaller rupture openings compared to the uncoated Zr. Although 4-point bend tests of the coated samples showed a small increase in the maximum load, ring compression tests of a sectioned sample showed increased ductility. - Highlights: • Cr and FeCrAl were coated onto Zr fuel cladding for light water nuclear reactors. • Mo layer between FeCrAl and Zr prevented inter-diffusion at high temperatures. • Coated claddings were tested under loss-of-cooling accident conditions. • Coating improved high-temperature oxidation resistance and mechanical properties.

  8. Sensitivity study for accident tolerant fuels: Property comparisons and behavior simulations in a simplified PWR to enable ATF development and design

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Kristina Yancey, E-mail: kristina.yancey@gmail.com; Sudderth, Laura; Brito, Ryan A.; Evans, Jordan A.; Hart, Clifford S.; Hu, Anbang; Jati, Andi; Stern, Karyn; McDeavitt, Sean M., E-mail: mcdeavitt@tamu.edu

    2016-12-01

    Highlights: • This study compared four accident tolerant fuels against uranium dioxide. • Material property correlations were developed to evaluate fuel performance. • The fuels’ neutronic and thermal hydraulic behaviors were studied in the AP1000. • No fuel type performed better in all areas, but each has strengths and weaknesses. • More research is needed to build a complete model of the fuel performances. - Abstract: Since the events at the Fukushima-Daiichi nuclear power plant, there has been increased interest in developing fuels to better withstand accidents for current light water reactors. Four accident tolerant fuel candidates are uranium oxide with beryllium oxide additives, uranium oxide with silicon carbide matrix additives, uranium nitride, and uranium nitride with uranium silicide composite. The first two candidates represent near-term high performance uranium oxide with high thermal conductivity and neutron transparency, and the second two represent mid-term high-density fuels with highly beneficial thermal properties. This study seeks to understand the benefits and drawbacks of each option in place of uranium dioxide. To assess the material properties for each of the fuel types, an extensive literature review was performed for material property data. Correlations were then made to evaluate the properties during reactor operation. Neutronics and thermal hydraulics studies were also completed to determine the impact of the use of each candidate in an AP1000 reactor. In most cases, the candidate fuels performed more desirably than uranium dioxide, but no fuel type performed better in all aspects. Much more research needs to be performed to build a complete model of the fuel performances, primarily experimental data for uranium silicide. Each of the fuels studied has its own benefits and drawbacks, and the comparisons discussed in this report can be used to aid in determining the most appropriate fuel depending on the desired specifications.

  9. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

    International Nuclear Information System (INIS)

    Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

    1987-11-01

    An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

  10. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  11. Study on the behavior of waterside corroded PWR fuel rods under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Sasajima, Hideo

    1989-06-01

    One of the highlighted problems from the fuel reliability point of view is a waterside corrosion of fuel cladding which becomes more significant at extended burnup stages. To date, at highly burned fuel, waterside corrosion was recognized as important because cladding oxidation increased with increasing burn-up. In experiments, as the basic research for the study of high burn-up fuel, the test fuel rods were prepressurized to ranges from 3.47 to 3.55 MPa, oxidized artificially to both 10 and 20 μm in thickness. Regarding fabricated oxide thickness of 10 μm, it is corresponded to be transition point from cubic law to linear law as a function of burn-up. Pulse irradiation experiments by NSRR were carried out to study the behavior of waterside corroded PWR type fuels under RIA conditions. Obtained results are: (1) The failure threshold of tested fuels was 110 cal/g·fuel (0.46 KJ/g·fuel) in enthalpy. This showed that the failure threshold of tested fuels was same as that of the past NSRR experimental data. (2) The failure mechanisms of the tested fuel rods was cladding rupture induced by ballooning. No differences in failure mechanisms existed between the past NSRR prepressurized standard fuel and the tested fuels. (3) Cracks were existed without propagating into cladding matrix, so that it was judged that these were not initiation of failure. (4) Whithin this experimental condition, reduction of cladding thickness being attributed to the increase of oxidation did not failure threshold. (author)

  12. Metallographic examination of damaged N reactor spent nuclear fuel element SFEC5,4378

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, S.C.; Pyecha, T.D.; Abrefah, J.

    1997-08-01

    N-Reactor spent nuclear fuel (SNF) is currently residing underwater in the K Basins at the Hanford site, in Richland, Washington. This report presents results of the metallographic examination of specimens cut from an SNF element (Mark IV-E) with breached cladding. The element had resided in the K-West (KW) Storage Basin for at least 10 years after it was discharged from the N-Reactor. The storage containers in the KW Basin were nominally closed, isolating the SNF elements from the open pool environment. Seven specimens from this Mark IV-E outer fuel element were examined using an optical metallograph. Included were two specimens that had been subjected to a conditioning process recommended by the Independent Technical Assessment Team, two specimens that had been subjected to a conditioning process recommended in the Integrated Process Strategy Report, and three that were in the as-received, as-cut condition. One of the as-received specimens had been cut from the damaged (or breached) end of the element. All other specimens were cut from the undamaged mid-region of the fuel element. The specimens were visually examined to (1) identify uranium hydride inclusions present in the uranium metal fuel, (2) measure the thickness of the oxide layer formed on the uranium edges and assess the apparent integrity and adhesion of the oxide layer, and (3) look for features in the microstructure that might provide an insight into the various corrosion processes that occurred during underwater storage in the KW Basin. These features included, but were not limited to, the integrity of the cladding and the fuel-to-cladding bond, obvious anomalies in the microstructure, excessive pitting or friability of the fuel matrix, and obvious anomalies in the distribution of uranium hydride or uranium carbide inclusions. Also, the observed metallographic features of the conditioned specimens were compared with those of the as-received (unconditioned) specimens. 11 refs., 93 figs., 2 tabs.

  13. Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Niels Gronbech; Asta, Mark; Ozolins, Nigel Browning' Vidvuds; de Walle, Axel van; Wolverton, Christopher

    2011-12-29

    The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

  14. Climate Science and the Responsibilities of Fossil Fuel Companies for Climate Damages and Adaptation

    Science.gov (United States)

    Frumhoff, P. C.; Ekwurzel, B.

    2017-12-01

    Policymakers in several jurisdictions are now considering whether fossil fuel companies might bear some legal responsibility for climate damages and the costs of adaptation to climate change potentially traceable to the emissions from their marketed products. Here, we explore how scientific research, outreach and direct engagement with industry leaders and shareholders have informed and may continue to inform such developments. We present the results of new climate model research quantifying the contribution of carbon dioxide and methane emissions traced to individual fossil fuel companies to changes in global temperature and sea level; explore the impact of such research and outreach on both legal and broader societal consideration of company responsibility; and discuss the opportunities and challenges for scientists to engage in further work in this area.

  15. Plutonium rock-like fuel LWR nuclear characteristics and transient behavior in accidents

    Energy Technology Data Exchange (ETDEWEB)

    Akie, Hiroshi; Anoda, Yoshinari; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yamaguchi, Chouichi; Sugo, Yukihiro

    1998-03-01

    For the disposition of excess plutonium, rock-like oxide (ROX) fuel systems based on zirconia (ZrO{sub 2}) or thoria (ThO{sub 2}) have been studied. Safety analysis of ROX fueled PWR showed it is necessary to increase Doppler reactivity coefficient and to reduce power peaking factor of zirconia type ROX (Zr-ROX) fueled core. For these improvements, Zr-ROX fuel composition was modified by considering additives of ThO{sub 2}, UO{sub 2} or Er{sub 2}O{sub 3}, and reducing Gd{sub 2}O{sub 3} content. As a result of the modification, comparable, transient behavior to UO{sub 2} fuel PWR was obtained with UO{sub 2}-Er{sub 2}O{sub 3} added Zr-ROX fuel, while the plutonium transmutation capability is slightly reduced. (author)

  16. MELCOR modeling of the PBF [Power Burst Facility] Severe Fuel Damage Test 1-4

    International Nuclear Information System (INIS)

    Madni, I.K.

    1990-01-01

    This paper describes a MELCOR Version 1.8 simulation of the Power Burst Facility (PBF) Severe Fuel Damage (SFD) Test 1--4. The input data for the analysis were obtained from the Test Results Report and from SCDAP/RELAP5 input. Results are presented for the transient liquid level in the test bundle, clad temperatures, shroud temperatures, clad oxidation and hydrogen generation, bundle geometry changes, fission product release, and heat transfer to the bypass flow. Comparisons are made with experimental data and with SCDAP/RELAP5 calculations. 10 refs., 7 figs

  17. Full-Length High-Temperature Severe Fuel Damage Test No. 5: Final safety analysis

    International Nuclear Information System (INIS)

    Lanning, D.D.; Lombardo, N.J.; Panisko, F.E.

    1993-09-01

    This report presents the final safety analysis for the preparation, conduct, and post-test discharge operation for the Full-Length High Temperature Experiment-5 (FLHT-5) to be conducted in the L-24 position of the National Research Universal (NRU) Reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test is sponsored by an international group organized by the US Nuclear Regulatory Commission. The test is designed and conducted by staff from Pacific Northwest Laboratory with CRNL staff support. The test will study the consequences of loss-of-coolant and the progression of severe fuel damage

  18. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Capolungo, L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wirth, B. D. [Univ. of Tennessee, Knoxville, TN (United States)

    2017-07-26

    Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. This allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to

  19. Computer code TRANS-ACE predicting for fire and explosion accidents in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Abe, Hitoshi; Nishio; Gunji; Naito, Yoshitaka

    1993-11-01

    The accident analysis code TRANS-ACE was developed to evaluate the safety of a ventilation system in a reprocessing plant in the event of fire and explosion accidents. TRANS-ACE can evaluate not only the integrity of a ventilation system containing HEPA filters but also the source term of radioactive materials for release out of a plant. It calculates the temperature, pressure, flow rate, transport of combustion materials and confinement of radioactive materials in the network of a ventilation system that might experience a fire or explosion accident. TRANS-ACE is based on the one-dimensional compressible thermo-fluid analysis code EVENT developed by Los Alamos National Laboratory (LANL). Calculational functions are added for the radioactive source term, heat transfer and radiation to cell and duct walls and HEPA filter integrity. For the second edition in the report, TRANS-ACE has been improved incorporating functions for the initial steady-state calculation to determine the flow rates, pressure drops and temperature in the network before an accident mode analysis. It is also improved to include flow resistance calculations of the filters and blowers in the network and to have an easy to use code by simplifying the input formats. This report is to prepare an explanation of the mathematical model for TRANS-ACE code and to be the user's manual. (author)

  20. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the advanced neutron source reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effect of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

  1. Radiation damage to the thyroid and metabolic changes in cattle in the initial and remote period after the Chernobyl accident

    International Nuclear Information System (INIS)

    Iljazov, R.G.; Yunousova, R.M.

    1997-01-01

    The initial period after the Chernobyl accident was the most dangerous for animals kept in the zone of radioactive contamination. Dose burdens from I-isotopes on the thyroid gland of cattle in the initial period after the accident contributed significantly into the alteration of the hormonal status, physiological state and productive, qualities of cattle on farms of the Gomel area of Belarus

  2. Fission product release in conditions of a spent fuel pool severe accident

    International Nuclear Information System (INIS)

    Ohai, Dumitru

    2007-01-01

    Full text: Depending on the residence time, fuel burnup, and fuel rack configuration, there may be sufficient decay heat for the fuel clad to heat up, swell, and burst in case of a loss of pool water. Initiating event categories can be: loss of offsite power from events initiated by severe weather, internal fire, loss of pool cooling, loss of coolant inventory, seismic event, aircraft impact, tornado, missile attack. The breach in the clad releases the radioactive gases present in the gap between the fuel and clad, what is called 'gap release'. If the fuel continues to heat up, the zirconium clad will reach the point of rapid oxidation in air. This reaction of zirconium and air, or zirconium and steam is exothermic. The energy released from the reaction, combined with the fuel's decay energy, can cause the reaction to become self-sustaining and ignite the zirconium. The increase in heat from the oxidation reaction can also raise the temperature in adjacent fuel assemblies and propagate the oxidation reaction. Simultaneously, the sintered UO 2 pellets resulting from pins destroying are oxidized. Due to the self-disintegration of pellets by oxidation, fission gases and low volatile fission products are released. The release rate, the chemical nature and the amount of fission products depend on powder granulation distribution and environmental conditions. The zirconium burning and pellets self-disintegration will result in a significant release of spent fuel fission products that will be dispersed from the reactor site. (author)

  3. Severe accident phenomena

    International Nuclear Information System (INIS)

    Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

    1995-02-01

    Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

  4. Increased Accident Tolerance of Fuels for Light Water Reactors - Workshop Proceedings, OECD/NEA Headquarters, Issy-les-Moulineaux, France, 10-12 December 2012

    International Nuclear Information System (INIS)

    2013-01-01

    The Fukushima accident in March 2011 raised concerns about the safety of current and future nuclear power plants both inside and outside the international nuclear energy community. With a view to learning lessons from this accident a large consensus emerged on the need to strengthen each level of Defence-In-Depth, reinforcing both prevention and mitigation. The fuel performance characteristics identified as being central to increased accident tolerance for long-term loss of coolant include reduced clad-steam reactions, reduced hydrogen production and improved fission product retention. New fuel designs which offered the potential to incorporate these characteristics, while retaining the operational performance of existing designs, would therefore be considered as suitable candidates for further investigation. Under the auspices of the NEA Nuclear Science Committee, a workshop has been organised to bring together international experts from the modelling, safety, operations and regulatory technical disciplines to discuss the various issues related to increased accident tolerance of fuels for Light Water Reactors and to help establish a co-ordinated international approach in this field. The organisation of this workshop was also supported by the NEA Committee on the Safety of Nuclear Installations. These proceedings include all the abstract papers presented at this workshop. The programme was comprised of 4 sessions: - Session 1: Lessons learned from the Fukushima accident; - Session 2: Accident-tolerant fuel design; - Session 3: Reactor operation, safety, fuel cycle constraints, economics and licensing; - Session 4: Synthesis and future programmes. A total of 55 participants from 16 countries attended the workshop, with 26 technical presentations and 2 breakout parallel sessions (one on safety issues, the other on reactor performance, R and D and technological issues). The attendees represented a broad spectrum of stakeholders involved in different nuclear energy

  5. Accident Analysis of High Density Storage Rack for Fresh Fuel Assemblies

    International Nuclear Information System (INIS)

    Jang, K. J.; Lee, M. J.; Jin, H. U.; Park, J. H.; Shin, S. Y.

    2009-01-01

    Recently KONES and KNF have developed the so called suspension-type High Density Storage Rack (HDSR) for fresh fuel assemblies. The USNRC OT position paper specifies that the design of the rack must ensure the functional integrity of the fuel racks under all credible fuel assembly drop events. In this context the functional integrity means the criticality safety. That is to say, the drop events must not bring any danger to the criticality safety of HDSR. This paper shows the results of the analysis carried out to demonstrate the regulatory compliance of the proposed racks under postulated accidental drop events

  6. Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    International Nuclear Information System (INIS)

    DeVault, G.P.; Bell, C.R.

    1985-01-01

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

  7. Proposed model for fuel-coolant mixing during a core-melt accident

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1983-01-01

    If complete failure of normal and emergency coolant flow occurs in a light water reactor, fission product decay heat would eventually cause melting of the reactor fuel and cladding. The core melt may then slump into the lower plenum and later into the reactor cavity and contact residual liquid water. A model is proposed to describe the fuel-coolant mixing process upon contact. The model is compared to intermediate scale experiments being conducted at Sandia. The modelling of this mixing process will aid in understanding three important processes: (1) fuel debris sizes upon quenching in water, (2) the hydrogen source term during fuel quench, and (3) the rate of steam production. Additional observations of Sandia data indicate that the steam explosion is affected by this mixing process

  8. 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.

  9. Experimental study of ballooning and failure of WWER-1000 fuel cans during maximum design basis accident

    International Nuclear Information System (INIS)

    Karetnikov, G.V.; Bogdanov, A.S.; Semishkin, V.P.; Bezrukov, Yu.A.; Trushin, A.M.; Frizen, E.A.

    2001-01-01

    The processes of ballooning and fracturing in tubular specimens of Eh635 and Eh110 alloy fuel cans are investigated with the use of cinematography. The investigations are carried out under steady-state conditions in the temperature range from 680 to 900 deg C and at pressure drops on the can from 2 to 12 MPa. Time dependences of circumferential strains are plotted for various temperatures of fuel cans at pressure of 2 MPa. It is shown that strain changes are of linear character at an initial portion of the curve and then an accelerated strain development takes place with transition to fracture. Using methods of nonlinear evaluation for time to fracture the approximation dependences are obtained for fuel cans. Experimental data are intended to form the equations of state for fuel can materials and to verify the program TVEL-3 [ru

  10. An analytical method to assess the damage and predict the residual strength of a ship in a shoal grounding accident scenario

    Directory of Open Access Journals (Sweden)

    Sun Bin

    2016-04-01

    Full Text Available In this paper, a simplified analytical method used to predict the residual ultimate strength of a ship hull after a shoal grounding accident is proposed. Shoal grounding accidents always lead to severe denting, though not tearing, of the ship bottom structure, which may threaten the global hull girder resistance and result in even worse consequences, such as hull collapse. Here, the degree of damage of the bottom structure is predicted by a series of analytical methods based on the plastic-elastic deformation mechanism. The energy dissipation of a ship bottom structure is obtained from individual components to determine the sliding distance of the seabed obstruction. Then, a new approach to assess the residual strength of the damaged ship subjected to shoal grounding is proposed based on the improved Smith's method. This analytical method is verified by comparing the results of the proposed method and those generated by numerical simulation using the software ABAQUS. The proposed analytical method can be used to assess the safety of a ship with a double bottom during its design phase and predict the residual ultimate strength of a ship after a shoal grounding accident occurs.

  11. Calculation of stricken to mortality and incidence cancers due to beyond design basis accidents of the Esfahan Fuel Production Factory

    International Nuclear Information System (INIS)

    Heydari Azar, A.; Shahshahani, M.; Roshanzamir, M.; Sabouhi, R.

    2008-01-01

    In this investigation the amount of absorbed doses by the different pathways of Cloud shine, Ground shine, deposition of radioactive materials on skin and cloths, ingestion, inhalation and the consequences of radioactive material releases due to Beyond Design Basis Accidents such as fire, sintering furnace explosion, criticality and earthquake in Esfahan Fuel Production factory by the residents are evaluated. The calculations related to atomic cloud distribution, estimation of delivered dose and decay chains are performed by PCCOSYMA dose. These computations are based on radioactive source terms, distribution height of radioactive materials. actions for reducing the absorbed dose, human body physiological characteristics, metrological condition and population distribution. Finally, the number of peoples who are stricken to mortality and morbidity cancers and risk values are calculated for 1 year and 50 years

  12. Design of and experience with the gamma-detecting criticality accident alarm system at ALKEM MOX fuel fabrication plant

    International Nuclear Information System (INIS)

    Kindleben, G.

    1988-01-01

    At ALKEM mixed oxide fuel fabrication plant there are two criticality accident alarm systems in operation and another one is planned for different buildings. They use ionization chambers for gamma-measuring. The measuring channels are self controlled with implemented test sources. The order of limit transgression at the detectors is registrated. The interpretation indicates the room of the radiation source, which is signaled by flash lights. Extensive radiation protection shieldings make detector-placing a complex problem with secondary gamma-radiation to be taken into account. Most of the appearing defects can easily be repaired by exchange of components. Some of them have been eliminated by technical modification. Redundancy prevents total system failure. Some false alarms occurred during the operation time of the alarm systems. The main reason is pulse induction, resulting from lightning strike. Measures to prevent such events have been taken, while further measures are being considered

  13. Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

    International Nuclear Information System (INIS)

    Bhaskaran, G.; McCleery, J.E.

    1979-10-01

    This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report

  14. Accident conditions analysis of spent fuel storage pool RA research reactor in Vinca; Analiza udesnih stanja u odlagalistu isluzenog goriva istrazivackog rektora RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Jovic, V; Jovic, L [Institute of Nuclear Sciences VINCA, Belgrade (Serbia and Montenegro)

    2000-07-01

    Based on Safety analysis of the spent fuel pool RA research reactor in Vinca, conditions and possibilities accident sequences in present configuration storage facility are considered (author) [Serbo-Croat] Na osnovu Analize sigurnosti odlagalista isluzenog goriva istrazivackog reaktora RA u Vinci razmatraju se uslovi i mogucnosti pojave udesnih stanja u postojecoj konfiguraciji odlagalista (author)

  15. International Experts’ Meeting on Reactor and Spent Fuel Safety in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant. Presentations

    International Nuclear Information System (INIS)

    2012-01-01

    The primary objectives of this International Experts’ Meeting (IEM) were: to analyse relevant technical aspects of reactor and spent nuclear fuel management safety and performance related to severe accidents; to review what is known to date about the accident at the Fukushima Daiichi nuclear power plant in order to understand more fully its root causes; and to share the lessons learned from the accident. The meeting identified the necessary priorities for further actions in these areas in different power reactor types, focusing in particular on boiling water reactors (BWRs) and pressurized water reactors (PWRs). The meeting provided a forum for discussions and exchange of information among technical experts from Member States on reactor and spent nuclear fuel safety and performance under severe conditions. The meeting was of particular interest to technical experts from utilities, research and design organizations, regulatory bodies, manufacturing and service companies and other stakeholders. In particular, the objectives of the meeting was to: • Identify and analyse reactor and spent nuclear fuel safety and performance issues; • Consider the design, engineering and analysis of current and new systems for accident prevention and mitigation; • Exchange information on national assessments of reactor and spent nuclear fuel safety and performance; and • Identify potential priority areas for research and development, technology development and management

  16. Sensitivity Study of the Peak Cladding Temperature for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

    International Nuclear Information System (INIS)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R.

    2005-12-01

    The effect of the thermal hydraulic operation parameters, the stroke times of safety-related valves, the node number of test fuel for MARS modeling, and the axial power distribution on the peak cladding temperature (PCT) has been investigated for the loss of coolant accident of the 3-pin fuel test loop. The thermal hydraulic operation parameters investigated are the thermal power of the fuel test loop and the flow rate, temperature, and pressure of the main cooling water. The effect of the thermal power and the coolant temperature on the peak cladding temperature is dominant as compared with that of the coolant flow rate and pressure. The maximum PCT increases up to about 34.3K for the room 1 LOCA when the thermal power increase by 5% of the normal operation power and decreases up to about 38.9K for the room 1 LOCA when the coolant temperature decrease by 2% of the normal operation temperature. The effect of the stroke time of the loop isolation valves on the PCT is also dominant. However the effect of the stroke time of the safety injection valves and depressurization vent valves are negligible. Especially the maximum PCT increases up to 25.7K with the increase of the design stroke time of the cold leg loop isolation valve by 13% and decreases up to 25.1K with the decrease of the design stroke time by 13%. The maximum PCT increases by 3.3K as the number of nodes increases from 7 to 14 for the MARS model of test fuel. Three different axial power distributions are also investigated. The maximum PCT occurs for the room 1 LOCA in case the peak power is shifted to the downstream by 20cm

  17. Accident progression event tree analysis for postulated severe accidents at N Reactor

    International Nuclear Information System (INIS)

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M.; Medford, G.T.

    1990-06-01

    A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied