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Sample records for fuel damage conditions

  1. 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

  2. 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

  3. 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)

  4. 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

  5. Out-of-pile UO2/Zircaloy-4 experiments under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Hofmann, P.

    1983-01-01

    Chemical interactions between UO 2 fuel and Zircaloy-4 cladding up to the melting point of zircaloy (Zry) are described. Out-of-pile UO 2 /zircaloy reaction experiments have been performed to investigate the chemical interaction behavior under possible severe fuel damage conditions (very high temperatures and external overpressure). The tests have been conducted in inert gas (1 to 80 bar) with 10-cm-long zircaloy cladding specimens filled with UO 2 pellets. The annealing temperature varied between 1000 and 1700 deg. C and the annealing period between 1 and 150 min. The extent of the chemical reaction depends decisively on whether or not good contact between UO 2 and zircaloy has been established. If solid contact exists, zircaloy reduces the UO 2 to form oxygen-stabilized α-Zr(O) and uranium metal. The uranium reacts with zircaloy to form a (U,Zr) alloy rich in uranium. The (U,Zr) alloy, which is liquid above approx. 1150 deg. C, lies between two α-Zr(O) layers. The UO 2 /zircaloy reaction obeys a parabolic rate law. The degree of chemical interaction is determined by the extent of oxygen diffusion into the cladding, and hence by the time and temperature. The affinity of zirconium for oxygen, which results in an oxygen gradient across the cladding, is the driving force for the reaction. The growth of the reaction layers can be represented in an Arrhenius diagram. The UO 2 /Zry-4 reaction occurs as rapidly as the steam/Zry-4 reaction above about 1100 deg. C. The extent of the interaction is independent of external pressure above about 10 bar at 1400 deg. C and 5 bar at 1700 deg. C. The maximum measured oxygen content of the cladding is approx. 6wt.%. Up to approx. 9 volume % of the UO 2 can be chemically dissolved by the zircaloy. In an actual fuel rod, complete release of the fission products in this region of the fuel must therefore be assumed. (author)

  6. 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

  7. 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

  8. Conditioning of nuclear reactor fuel

    International Nuclear Information System (INIS)

    1975-01-01

    A method of conditioning the fuel of a nuclear reactor core to minimize failure of the fuel cladding comprising increasing the fuel rod power to a desired maximum power level at a rate below a critical rate which would cause cladding damage is given. Such conditioning allows subsequent freedom of power changes below and up to said maximum power level with minimized danger of cladding damage. (Auth.)

  9. 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)

  10. Out-of-pile bundle temperature escalation under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Hagen, S.; Peck, S.O.

    1983-08-01

    This report provides an overview of the test conduct, results, and posttest appearance of bundle test ESBU-1. The purpose of the test was to investigate fuel rod temperature escalation due to the exothermal zircaloy/steam reaction in a bundle geometry. The 3x3 bundle was surrounded by a zircaloy shroud and 6 mm of fiber ceramic insulation. The center rod escalated to a maximum of 2,250 0 C. Runoff of the melt apparently limited the escalation. Posttest visual examination of the bundle showed that cladding from every rod had melted, liquefied some fuel, flowed down the rod, and frozen in a solid mass that substantially blocked all flow channels. A large amount of powdery rubble, probably fuel that fractured during cooldown, was found on top of the blockage. Metallographic, EMP, and SEM examinations showed that the melt had dissolved both fuel and oxidized cladding, and had itself been oxidized by steam. (orig.) [de

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. Behaviour of a VVER-1000 fuel element with boron carbide/steel absorber tested under severe fuel damage conditions in the CORA facility (Results of experiment CORA-W2)

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.; Sepold, L.

    1994-10-01

    The 'Severe Fuel Damage' (SFD) experiments of the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, were carried out in the out-of-pile facility 'CORA' as part of the international Severe Fuel Damage (SFD) research. The experimental program was set up to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in few cases up to 2400 C. Between 1987 and 1992 a total of 17 CORA experiments with two different bundle configurations, i.e. PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles were performed. These assemblies represented 'Western-type' fuel elements with the pertinent materials for fuel, cladding, grid spacer, and absorber rod. At the end of the experimental program two VVER-1000 specific tests were run in the CORA facility with identical objectives but with genuine VVER-type materials. The experiments, designated CORA-W1 and CORA-W2 were conducted on February 18, 1993 and April 21, 1993, respectively. Test bundle CORA-W1 was without absorber material whereas CORA-W2 contained one absorber rod (boron carbide/steel). As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zircon/niobium-steam reaction started at about 1200 C, leading the bundles to maximum temperatures of approximately 1900 C. The thermal response of bundle CORA-W2 is comparable to that of CORA-W1. In test CORA-W2, however, the temperature front moved faster from the top to the bottom compared to test CORA-W1 [de

  20. 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)

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. CANDU fuel behaviour under transient conditions

    International Nuclear Information System (INIS)

    Segel, A.W.L.

    1979-04-01

    The Canadian R and D program to understand CANDU fuel behaviour under transient conditions is described. Fuel sheath behaviour studies have led to the development of a model of transient plastic strain in inert gas, which integrates the deformation due to several mechanisms. Verification tests demonstrated that on average the model overpredicts strain by 20%. From oxidation kinetics studies a sheath failure embrittlement criterion based on oxygen distribution has been developed. We have also established a rate equation for high-temperature stress-dependent crack formation due to embrittlement of the sheath by beryllium. An electric, simulated fuel element is being used in laboratory tests to characterize the behaviour of fuel in the horizontal. In-reactor, post-dryout tests have been done for several years. There is an axially-segmented, axisymmetric fuel element model in place and a fully two-dimensional code is under development. Laboratory testing of bundles, in its early stages, deals with the effects of geometric distortion and sheath-to-sheath interaction. In-reactor, post-dryout tests of CANDU fuel bundles with extensive central UO 2 melting did not result in fuel fragmentation nor damage to the pressure tube. (author)

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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)

  12. Creep-fatigue damage under multiaxial conditions

    International Nuclear Information System (INIS)

    Lobitz, D.W.; Nickell, R.E.

    1977-02-01

    ASME Code rules for design against creep-fatigue damage for Class 1 nuclear components operating at elevated temperatures are currently being studied by ASME working groups and task forces with a view toward major modification. In addition, the design rules being developed for Class 2 and Class 3 components would be affected by any major modifications of Class 1 Rules. The report represents an attempt to evaluate the differences between two competing procedures--linear damage summation and strainrange partitioning--for multiaxial stress conditions. A modified version of strainrange partitioning is also developed to alleviate some limitations on nonproportional loading

  13. 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)

  14. Defect accumulation under cascade damage conditions

    DEFF Research Database (Denmark)

    Trinkaus, H.; Singh, B.N.; Woo, C.H.

    1994-01-01

    in terms of this reaction kinetics taking into account cluster production, dissociation, migration and annihilation at extended sinks. Microstructural features which are characteristic of cascade damage and cannot be explained in terms of the conventional single defect reaction kinetics are emphasized......There is now ample evidence from both experimental and computer simulation studies that in displacement cascades not only intense recombination takes place but also efficient clustering of both self-interstitial atoms (SIAs) and vacancies. The size distributions of the two types of defects produced...... reactions kinetics associated with the specific features of cascade damage is described, with emphasis on asymmetries between SIA and vacancy type defects concerning their production, stability, mobility and interactions with other defects. Defect accumulation under cascade damage conditions is discussed...

  15. Creep fatigue damage under multiaxial conditions

    International Nuclear Information System (INIS)

    Lobitz, D.W.; Nickell, R.E.

    1977-01-01

    When structural components are subjected to severe cyclic loading conditions with intermittent periods of sustained loading at elevated temperature, the designer must guard against a failure mode caused by the interaction of time-dependent and time-independent deformation. This phenomena is referred to as creep-fatigue interaction. The most elementary form of interaction theory (called linear damage summation) is now embodied in the ASME Boiler and Pressure Vessel Code. In recent years, a competitor for the linear damage summation theory has emerged, called strainrange partitioning. This procedure is based upon the visualization of the cyclic strain in a uniaxial creep-fatigue test as a hysteresis loop, with the inelastic strains in the loop counter-balanced in one of two ways. The two theories are compared and contrasted in terms of ease of use, possible inconsistencies, and component life prediction. Future work to further test the damage theories is recommended

  16. Fuel conditioning facility material accountancy

    International Nuclear Information System (INIS)

    Yacout, A.M.; Bucher, R.G.; Orechwa, Y.

    1995-01-01

    The operation of the Fuel conditioning Facility (FCF) is based on the electrometallurgical processing of spent metallic reactor fuel. It differs significantly, therefore, from traditional PUREX process facilities in both processing technology and safeguards implications. For example, the fissile material is processed in FCF only in batches and is transferred within the facility only as solid, well-characterized items; there are no liquid steams containing fissile material within the facility, nor entering or leaving the facility. The analysis of a single batch lends itself also to an analytical relationship between the safeguards criteria, such as alarm limit, detection probability, and maximum significant amount of fissile material, and the accounting system's performance, as it is reflected in the variance associated with the estimate of the inventory difference. This relation, together with the sensitivity of the inventory difference to the uncertainties in the measurements, allows a thorough evaluation of the power of the accounting system. The system for the accountancy of the fissile material in the FCF has two main components: a system to gather and store information during the operation of the facility, and a system to interpret this information with regard to meeting safeguards criteria. These are described and the precision of the inventory closure over one batch evaluated

  17. 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

  18. Fuel corrosion processes under waste disposal conditions

    International Nuclear Information System (INIS)

    Shoesmith, D.W.

    2000-01-01

    The release of the majority of radionuclides from spent nuclear fuel under permanent disposal conditions will be controlled by the rate of dissolution of the UO 2 fuel matrix. In this manuscript the mechanism of the coupled anodic (fuel dissolution) and cathodic (oxidant reduction) reactions which constitute the overall fuel corrosion process is reviewed, and the many published observations on fuel corrosion under disposal conditions discussed. The primary emphasis is on summarizing the overall mechanistic behaviour and establishing the primary factors likely to control fuel corrosion. Included are discussions on the influence of various oxidants including radiolytic ones, pH, temperature, groundwater composition, and the formation of corrosion product deposits. The relevance of the data recorded on unirradiated UO 2 to the interpretation of spent fuel behaviour is included. Based on the review, the data used to develop fuel corrosion models under the conditions anticipated in Yucca Mountain (NV, USA) are evaluated

  19. 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)

  20. 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)

  1. 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

  2. International experience in conditioning spent fuel elements

    International Nuclear Information System (INIS)

    Ashton, P.

    1991-04-01

    The purpose of this report is to compile and present in a clear form international experience (USA, Canada, Sweden, FRG, UK, Japan, Switzerland) gained to date in conditioning spent fuel elements. The term conditioning is here taken to mean the handling and packaging of spent fuel elements for short- or long-term storage or final disposal. Plants of a varying nature fall within this scope, both in terms of the type of fuel element treated and the plant purpose eg. experimental or production plant. Emphasis is given to plants which bear some similarity to the concept developed in Germany for direct disposal of spent fuel elements. Worldwide, however, relatively few conditioning plants are in existence or have been conceived. Hence additional plants have been included where aspects of the experience gained are also of relevance eg. plants developed for the consolidation of spent fuel elements. (orig./HP) [de

  3. 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)

  4. 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

  5. Fuel corrosion processes under waste disposal conditions

    International Nuclear Information System (INIS)

    Shoesmith, D.W.

    1999-09-01

    Under the oxidizing conditions likely to be encountered in the Yucca Mountain Repository, fuel dissolution is a corrosion process involving the coupling of the anodic dissolution of the fuel with the cathodic reduction of oxidants available within the repository. The oxidants potentially available to drive fuel corrosion are environmental oxygen, supplied by the transport through the permeable rock of the mountain and molecular and radical species produced by the radiolysis of available aerated water. The mechanism of these coupled anodic and cathodic reactions is reviewed in detail. While gaps in understanding remain, many kinetic features of these reactions have been studied in considerable detail, and a reasonably justified mechanism for fuel corrosion is available. The corrosion rate is determined primarily by environmental factors rather than the properties of the fuel. Thus, with the exception of increase in rate due to an increase in surface area, pre-oxidation of the fuel has little effect on the corrosion rate

  6. The conditions of gaseous fuels development

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    Face to the actual situation of petrol and gas oil in France, the situation of gaseous fuels appears to be rather modest. However, the aim of gaseous fuels is not to totally supersede the liquid fuels. Such a situation would imply a complete overturn which has not been seriously considered yet. This short paper describes the essential conditions to promote the wider use of gaseous fuels: the intervention of public authorities to adopt a more advantageous tax policy in agreement with the ''Clean Air''law project, a suitable distribution network for gaseous fuels, a choice of vehicles consistent with the urban demand, the development of transformation kits of quality and of dual-fuel vehicles by the car manufacturers. (J.S.)

  7. Fuel corrosion processes under waste disposal conditions

    Energy Technology Data Exchange (ETDEWEB)

    Shoesmith, D.W. [Univ. of Western Ontario, Dept. of Chemistry, London, Ontario (Canada)

    1999-09-01

    Under the oxidizing conditions likely to be encountered in the Yucca Mountain Repository, fuel dissolution is a corrosion process involving the coupling of the anodic dissolution of the fuel with the cathodic reduction of oxidants available within the repository. The oxidants potentially available to drive fuel corrosion are environmental oxygen, supplied by the transport through the permeable rock of the mountain and molecular and radical species produced by the radiolysis of available aerated water. The mechanism of these coupled anodic and cathodic reactions is reviewed in detail. While gaps in understanding remain, many kinetic features of these reactions have been studied in considerable detail, and a reasonably justified mechanism for fuel corrosion is available. The corrosion rate is determined primarily by environmental factors rather than the properties of the fuel. Thus, with the exception of increase in rate due to an increase in surface area, pre-oxidation of the fuel has little effect on the corrosion rate.

  8. 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

  9. 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

  10. The LP-FP-2 severe fuel damage scenario and discussion of the relative influence of the transient and reflood phases in affecting the final condition of the bundle

    International Nuclear Information System (INIS)

    Modro, S.M.; Carboneau, M.L.

    1990-01-01

    The purpose of this paper is to review the evidence from the OECD LP-FP-2 experiment that a high temperature excursion occurred within the center fuel module (CFM) during the reflood portion of the test, was caused by rapid metal-water reaction. It is shown that this reflood scenario explains many perplexing observations from the experiment, in particular, the small amount of fission products and hydrogen transported to the blowdown suppression tank (BST) as compared with the larger quantities trapped within the primary coolant system (PCS). The timing and destruction of the CFM upper tie plate, as well as the transport of fuel debris to the top of this plate, are also explained. In general, all measurements, observations, and analyses of the LP-FP-2 data indicate that most of the CFM damage occurred during a relatively short period of time coincident with the reflood portion of the experiment. 4 refs., 6 figs

  11. Evaluation of LWR fuel rod behavior under operational transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Hiramoto, K.; Maru, A.

    1984-01-01

    To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. CANDU fuel behaviour under LOCA conditions

    International Nuclear Information System (INIS)

    Kohn, E.

    1989-07-01

    This report summarizes the current understanding of CANDU fuel-element behaviour under loss-of-coolant (LOCA) accidents. It focuses on a key in-reactor verification experiment conducted at Idaho National Engineering Laboratory (INEL) and on three Canadian in-reactor tests. The in-reactor data, and the considerable body of supporting information developed from out-reactor tests, support the general conclusion that CANDU fuel behaviour during LOCA transients is well understood. Four elements of 37-element CANDU fuel-bundle design were tested under conditions typical of a large-break LOCA blowdown in a CANDU reactor. The purpose of the test was to confirm our current understanding of fuel behaviour under loss-of-coolant accident blowdown conditions. The test also provided data for comparison with predictions made with the steady-state and transient fuel-element performance codes ELESIM and ELOCA. Key components of typical LOCA transients were incorporated in the test: namely, a rapid depressurization rate of the hot coolant, a simultaneous power increase before decreasing to decay values (a power pulse), and prototype fuel element under pre-transient power and burnup conditions. The test was successfully completed in the Power Burst Facility (PBF) reactor at INEL under contract to Ontario Hydro and AECL. The three CANDU Owners Group LOCA tests performed at Chalk River Nuclear Laboratories measured both the thermal-mechanical response and fission-gas release resulting from exposure to a LOCA transient. Results from these three tests provided further confirmation that the behaviour of the fuel under LOCA conditions is understood

  17. Methods of conditioning direct methanol fuel cells

    Science.gov (United States)

    Rice, Cynthia; Ren, Xiaoming; Gottesfeld, Shimshon

    2005-11-08

    Methods for conditioning the membrane electrode assembly of a direct methanol fuel cell ("DMFC") are disclosed. In a first method, an electrical current of polarity opposite to that used in a functioning direct methanol fuel cell is passed through the anode surface of the membrane electrode assembly. In a second method, methanol is supplied to an anode surface of the membrane electrode assembly, allowed to cross over the polymer electrolyte membrane of the membrane electrode assembly to a cathode surface of the membrane electrode assembly, and an electrical current of polarity opposite to that in a functioning direct methanol fuel cell is drawn through the membrane electrode assembly, wherein methanol is oxidized at the cathode surface of the membrane electrode assembly while the catalyst on the anode surface is reduced. Surface oxides on the direct methanol fuel cell anode catalyst of the membrane electrode assembly are thereby reduced.

  18. 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.

  19. 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

  20. Safeguardability of advanced spent fuel conditioning process

    Energy Technology Data Exchange (ETDEWEB)

    Li, T. K. (Tien K.); Lee, S. Y. (Sang Yoon); Burr, Tom; Russo, P. A. (Phyllis A.); Menlove, Howard O.; Kim, H. D.; Ko, W. I. (Won Il); Park, S. W.; Park, H. S.

    2004-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) is an electro-metallurgical treatment technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology since 1977 for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. By using of this technology, a significant reduction of the volume and heat load of spent fuel is expected, which would lighten the burden of final disposal in terms of disposal size, safety and economics. In the framework of collaboration agreement to develop the safeguards system for the ACP, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and the KAERI since 2002. In this study, the safeguardability of the ACP technology was examined for the pilot-scale facility. The process and material flows were conceptually designed, and the uncertainties in material accounting were estimated with international target values.

  1. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  2. 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.

  3. 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)

  4. 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

  5. Development and engineering plan for graphite spent fuels conditioning program

    International Nuclear Information System (INIS)

    Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

    1993-09-01

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

  6. Random accumulated damage evaluation under multiaxial fatigue loading conditions

    Directory of Open Access Journals (Sweden)

    V. Anes

    2015-07-01

    Full Text Available Multiaxial fatigue is a very important physical phenomenon to take into account in several mechanical components; its study is of utmost importance to avoid unexpected failure of equipment, vehicles or structures. Among several fatigue characterization tools, a correct definition of a damage parameter and a load cycle counting method under multiaxial loading conditions show to be crucial to estimate multiaxial fatigue life. In this paper, the SSF equivalent stress and the virtual cycle counting method are presented and discussed, regarding their physical foundations and their capability to characterize multiaxial fatigue damage under complex loading blocks. Moreover, it is presented their applicability to evaluate random fatigue damage.

  7. 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

  8. Predicting fuel performance for SP-100 conditions

    International Nuclear Information System (INIS)

    Baars, R.E.

    1985-01-01

    This paper reports on methods for analyzing fuel designs proposed for the thermionic and thermoelectric concepts for SP-100 application. The proposed fuel design for the thermionic concept consisted of fully-enriched oxide fuel clad in chemical vapor deposition (CVD) tungsten, which also served as the emitter for the thermionic fuel element (TFE). The fuel density was 95% of theoretical with the linear heat rate flattened radially by removing fuel from the center of the fuel pellet. The fuel inner diameter varied from approx.0.45 in. at the core center to zero at the edge of the core. The as-fabricated gap between fuel and emitter was 10 mils radial. The emitter thickness was 80 mils, and the outer diameter was 1.099 in. The LIFE-4 code was used for evaluation of this concept after extensive review of the code and development of a procedure that corrects certain deficiencies noted in analysis of several tests

  9. Effect of site conditions on ground motion and damage

    Science.gov (United States)

    Borcherdt, R.; Glassmoyer, G.; Andrews, M.; Cranswick, E.

    1989-01-01

    Results of seismologic studies conducted by the U.S. reconnaissance team in conjunction with Soviet colleagues following the tragic earthquakes of December 7, 1988, suggest that site conditions may have been a major factor in contributing to increased damage levels in Leninakan. As the potential severity of these effects in Leninakan had not been previously identified, this chapter presents results intended to provide a preliminary quantification of these effects on both damage and levels of ground motion observed in Leninakan. The article describes the damage distribution geologic setting, ground motion amplification in Leninakan, including analog amplifications and spectral amplifications. Preliminary model estimates for site response are presented. It is concluded that ground motion amplification in the 0.5-2.5-second period range was a major contributing factor to increased damage in Leninakan as compared with Kirovakan. Leninakan is located on thick water saturated alluvial deposits.

  10. 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)

  11. Optimization of Fuel Cell System Operating Conditions for Fuel Cell Vehicles

    OpenAIRE

    Zhao, Hengbing; Burke, Andy

    2008-01-01

    Proton Exchange Membrane fuel cell (PEMFC) technology for use in fuel cell vehicles and other applications has been intensively developed in recent decades. Besides the fuel cell stack, air and fuel control and thermal and water management are major challenges in the development of the fuel cell for vehicle applications. The air supply system can have a major impact on overall system efficiency. In this paper a fuel cell system model for optimizing system operating conditions was developed wh...

  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. Fuel conditioning facility electrorefiner volume calibration

    International Nuclear Information System (INIS)

    Bucher, R.G.; Orechwa, Y.

    1995-01-01

    In one of the electrometallurgical process steps of the Fuel Conditioning Facility (FCF), die in-process nuclear material is dissolved in the electrorefiner tank in an upper layer of a mixture of liquid LiCl-KCl salt and a lower layer of liquid cadmium. The electrorefiner tank, as most process tanks, is not a smooth right-circular cylinder for which a single linear volume calibration curve could be fitted over the whole height of the tank. Rather, the tank contains many internal components, which cause systematic deviations from a single linear function. The nominal operating temperature of the electrorefiner is 500 degrees C although the salt and cadmium are introduced at 410 degrees C. The operating materials and temperatures preclude multiple calibration runs at operating conditions. In order to maximize the calibration information, multiple calibration runs were performed with water at room temperature. These data allow identification of calibration segments, and preliminary estimation of the calibration function and calibration uncertainties. The final calibration function is based on a combination of data from die water calibrations and the measurements made during the filling of the electrorefiner with salt and cadmium for operation

  14. Studies in Phebus reactor of fuel behaviour upon LOCA conditions

    International Nuclear Information System (INIS)

    Manin, A.; Del Negro, R.; Reocreux, M.

    1980-09-01

    The fuel behaviour upon LOCA conditions is studied in an in-pile loop, in Phebus reactor. This paper presents: a short description of Phebus reactor; the current program (adjusting the thermohydraulic conditions in order to get cladding failure); the program developments (consequences involved by cladding failure); the fuel test conditions determination [fr

  15. 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.

  16. 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

  17. Failure analysis of carbide fuels under transient overpower (TOP) conditions

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1980-06-01

    The failure of carbide fuels in the Fast Test Reactor (FTR) under Transient Overpower (TOP) conditions has been examined. The Beginning-of-Cycle Four (BOC-4) all-oxide base case, at $.50/sec ramp rate was selected as the reference case. A coupling between the advanced fuel performance code UNCLE-T and HCDA Code MELT-IIIA was necessary for the analysis. UNCLE-T was used to determine cladding failure and fuel preconditioning which served as initial conditions for MELT-III calculations. MELT-IIIA determined the time of molten fuel ejection from fuel pin

  18. Fuel conditioning facility electrorefiner start-up results

    International Nuclear Information System (INIS)

    Goff, K.M.; Mariani, R.D.; Vaden, D.; Bonomo, N.L.; Cunningham, S.S.

    1996-01-01

    At ANL-West, there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. The treatment operations will make use of an electrometallurgical process employing molten salts and liquid metals. The treatment equipment is presently undergoing testing with depleted uranium. Operations with irradiated fuel will commence when the environmental evaluation for FCF is complete

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

    International Nuclear Information System (INIS)

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

    1982-11-01

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

  20. 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

  1. 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

  2. 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.)

  3. 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.)

  4. Aspects of microstructure evolution under cascade damage conditions

    International Nuclear Information System (INIS)

    Singh, B.N.; Trinkaus, H.; Barashev, A.V.

    1997-01-01

    The conventional theoretical models describing the damage accumulation, particularly void swelling, under cascade damage conditions do not include treatments of important features such as intracascade clustering of self-interstitial atoms (SIAs) and one-dimensional glide of SIA clusters produced in the cascades. Recently, it has been suggested that the problem can be treated in terms of 'production bias' and one-dimensional glide of small SIA clusters. In the earlier treatments a 'mean size approximation' was used for the defect clusters and cavities evolving during irradiation. In the present work, we use the 'size distribution function' to determine the dose dependence of sink strengths, vacancy supersaturation and void swelling as a function of dislocation density and grain size within the framework of production bias model and glide of small SIA clusters. In this work, the role of the sessile-glissile loop transformation (due to vacancy supersaturation) on the damage accumulation behaviour is included. The calculated results on void swelling are compared with the experimental results as well as the results of the earlier calculations using the 'mean size approximation'. The calculated results agree very well with the experimental results. (orig.)

  5. Results of tests under normal and abnormal operating conditions concerning LMFBR fuel element behaviour

    International Nuclear Information System (INIS)

    Languille, A.; Bergeonneau, P.; Essig, C.; Guerin, Y.

    1985-04-01

    The objective of this paper is to improve the knowledge on LMFBR fuel element behaviour during protected and unprotected transients in RAPSODIE and PHENIX reactors in order to evaluate its reliability. The range of the tests performed in these reactors is sufficiently large to cover normal and also extreme off normal conditions such as fuel melting. Results of such tests allow to better establish transient design limits for reactor structural components in particular for fuel pin cladding which play a lead role in controlling the accident sequence. Three main topics are emphasized in this paper: fuel melting during slow over-power excursions; influence of the fuel element geometrical evolution on reactivity feedback effects and reactor dynamic behaviour; clad damage evaluation during a transient (essentially very severe loss of flow)

  6. Electrostatic fuel conditioning of internal combustion engines

    Science.gov (United States)

    Gold, P. I.

    1982-01-01

    Diesel engines were tested to determine if they are influenced by the presence of electrostatic and magnetic fields. Field forces were applied in a variety of configurations including pretreatment of the fuel and air, however, no affect on engine performance was observed.

  7. Final Report - Durable Catalysts for Fuel Cell Protection during Transient Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Atanasoski, Radoslav [3M Company, St. Paul, MN (United States); van der Vliet, Dennis [3M Company, St. Paul, MN (United States); Cullen, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Atanasoska, Ljiljana [3M Company, St. Paul, MN (United States)

    2015-01-26

    The objective of this project was to develop catalysts that will enable proton exchange membranes (PEM) fuel cell systems to weather the damaging conditions in the fuel cell at voltages beyond the thermodynamic stability of water during the transient periods of start-up/shut-down and fuel starvation. Such catalysts are required to make it possible for the fuel cell to satisfy the 2015 DOE targets for performance and durability. The project addressed a key issue of importance for successful transition of PEM fuel cell technology from development to pre-commercial phase. This issue is the failure of the catalyst and the other thermodynamically unstable membrane electrode assembly (MEA) components during start-up/shut-down and local fuel starvation at the anode, commonly referred to as transient conditions. During these periods the electrodes can reach potentials higher than the usual 1.23V upper limit during normal operation. The most logical way to minimize the damage from such transient events is to minimize the potential seen by the electrodes. At lower positive potentials, increased stability of the catalysts themselves and reduced degradation of the other MEA components is expected.

  8. 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)

  9. 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.)

  10. 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.)

  11. Criticality control during conditioning of spent nuclear fuel in the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Khalil, H.S.

    1994-01-01

    Spent nuclear fuel may be unacceptable for direct repository storage because of composition, enrichment, form, physical condition, or the presence of undesirable materials such as sodium. Fuel types which are not acceptable for direct storage must be processed or conditioned to produce physical forms which can safely be stored in a repository. One possible approach to conditioning is the pyroprocess implemented in the Fuel Cycle Facility (FCF) at Argonne National Laboratory-West. Conditioning of binary (U-Zr) and ternary (U-Pu-Zr) metallic fuels from the EBR-2 reactor is used to demonstrate the process. Criticality safety considerations limit batch sizes during the conditioning steps and provide one constraint on the final form of conditioned material. Criticality safety during conditioning is assured by the integration of criticality safety analysis, equipment design, process development, a measurement program, accountability procedures, and a computerized Mass Tracking System. Criticality issues related to storage and shipment of conditioned material have been examined

  12. 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.

  13. 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.

  14. Methods for measuring of fuel can deformation under radiation conditions

    International Nuclear Information System (INIS)

    Zelenchuk, A.V.; Fetisov, B.V.; Lakin, Yu.G.; Tonkov, V.Yu.

    1978-01-01

    The possibility for measuring fuel can deformation under radiation conditions by means of the acoustic method and tensoresistors is considered. The construction and operation of the in-pile facility for measuring creep of the fuel can specimen loaded by the internal pressure is described. The data on neutron radiation effect on changes in creep rate for zirconium fuel can are presented. The results obtained with tensoresistors are in a good agreement with those obtained by the acoustic method, which enables to recommend the use of both methods for the irradiation creep investigation of the fuel element cans

  15. Biomass co-firing under oxy-fuel conditions

    DEFF Research Database (Denmark)

    Álvarez, L.; Yin, Chungen; Riaza, J.

    2014-01-01

    This paper presents an experimental and numerical study on co-firing olive waste (0, 10%, 20% on mass basis) with two coals in an entrained flow reactor under three oxy-fuel conditions (21%O2/79%CO2, 30%O2/70%CO2 and 35%O2/65%CO2) and air–fuel condition. Co-firing biomass with coal was found...... to have favourable synergy effects in all the cases: it significantly improves the burnout and remarkably lowers NOx emissions. The reduced peak temperatures during co-firing can also help to mitigate deposition formation in real furnaces. Co-firing CO2-neutral biomass with coals under oxy-fuel conditions...... the model can be used to aid in design and optimization of large-scale biomass co-firing under oxy-fuel conditions....

  16. Review of the Effects of Normal Conditions of Transport on Spent Fuel Integrity in Transportation Casks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Junggoo; Yoo, Youngik; Lee, Seongki; Lim, Chaejoon [Korea Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2014-10-15

    Spent fuel(SF) storage capacity of each domestic nuclear power plant will reach a saturated state in the near future. Although there are several methods of SF disposal, interim storage is suggested as the most realistic and promising alternative. SF integrity evaluation is a regulatory requirement that is described in Part 71 of Code of Federal Regulations, Title 10 of the U..S. NRC licensing requirement. In this paper, the report is reviewed written by EPRI in US and it is helpful to a development of domestic SF integrity evaluation technology. EPRI report about integrity evaluation method on normal conditions of high burn-up spent fuel transport is reviewed. First, dynamic forces occurred in one-foot side drop are calculated. And deformation patterns and fuel rods responses by dynamic forces calculated from spent fuel and cask model are analyzed. It is shown that the damage of fuel rods is not occurred by the dynamic forces on normal conditions. Assembly distortion is not predicted, by virtue of the facts that the spacer grids do not experience significant permanent deformation. Axial forces, bending moments and pinch forces of fuel rods are calculated and compared with the results under the hypothetical accident conditions. No occurrence of transverse tearing mode that is the most serious damage mode in side drop case is predicted. Till now, in Korea, regulatory requirements related with structural integrity of spent fuel are not specified such as 10CFR71. To establish own regulation standards, producing and analyzing sufficient experimental data must be performed preferentially. Based on this, failure analysis and criteria establishment are necessary through modeling and analyzing of spent fuel.

  17. 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.

  18. 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

  19. INPR ACPR utilization in fuel behaviour studies under accidental condition

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Popov, Mircea

    1990-01-01

    This paper is dedicated to the experimental program, investigating CANDU type fuel behaviour in transient condition, as well as the facilities supporting this program. The tests Reactivity Initiated Accident type. The experiments were performed within TRIGA ACPR facility, installed at INSTITUTE for NUCLEAR POWER REACTORS, Pitesti, ROMANIA. Studies of the safety issues took a great international developement during last years. In USA, Japan, owners of the similar reactors, and USSR there are a big commitment to such programs, intended to establish the nuclear fuel behaviour under RIA-conditions. In our country, too, there are programs aiming a complete testing of the CANDU type fuels. As it is known, RIA is not a CANDU specific accident, but the fuel behaviour in such conditions can give useful informations on the fuel cladding failure threshold and about reflooding post LOCA heat transfer condition. Based on some papers and specific requirements it was initiated and developed a safety research program on CANDU type fuel using the ACPR. The paper describes the reactor,test capsule, instrumentation, fuel samples, tests, post irradiation results. (orig.)

  20. 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

  1. 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

  2. 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.

  3. 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

  4. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-15

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

  5. Behaviour of Spent WWER fuel under long term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kadarmetov, I M [A.A.Bochvar All-Russia Research Institute of Inorganic Materials, Moscow (Russian Federation)

    1999-07-02

    Results of experimental investigation into thermomechanical properties of pre-irradiated Zr-1%Nb alloy over a range temperatures 500-570 grad C are presented. Safety examination of the Ventilation Storage Casks dry storage system has been carried out. Preliminary safety criteria under dry storage conditions in an environment of inert gas are follows: maximum cladding temperature under normal conditions of dry storage should not exceed 330 grad C after 5-year cooling in water-filled pools; maximum allowable temperature of spent fuel rod cladding under operational mode with infringement of heat removal should not exceed 440 grad C over 8 hours. As each SFA dry storage project comprises its individual technology of spent fuel management, it is necessary to evaluate allowable parameters (terms of storage, maximum temperatures of fuel) for each project respectively. The programme of experimental investigations for the justification of safety criteria for WWER-1000 dry spent fuel storage systems is underway. (author)

  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. 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

  8. Effects of ambient conditions on fuel cell vehicle performance

    Science.gov (United States)

    Haraldsson, K.; Alvfors, P.

    Ambient conditions have considerable impact on the performance of fuel cell hybrid vehicles. Here, the vehicle fuel consumption, the air compressor power demand, the water management system and the heat loads of a fuel cell hybrid sport utility vehicle (SUV) were studied. The simulation results show that the vehicle fuel consumption increases with 10% when the altitude increases from 0 m up to 3000 m to 4.1 L gasoline equivalents/100 km over the New European Drive Cycle (NEDC). The increase is 19% on the more power demanding highway US06 cycle. The air compressor is the major contributor to this fuel consumption increase. Its load-following strategy makes its power demand increase with increasing altitude. Almost 40% of the net power output of the fuel cell system is consumed by the air compressor at the altitude of 3000 m with this load-following strategy and is thus more apparent in the high-power US06 cycle. Changes in ambient air temperature and relative humidity effect on the fuel cell system performance in terms of the water management rather in vehicle fuel consumption. Ambient air temperature and relative humidity have some impact on the vehicle performance mostly seen in the heat and water management of the fuel cell system. While the heat loads of the fuel cell system components vary significantly with increasing ambient temperature, the relative humidity did not have a great impact on the water balance. Overall, dimensioning the compressor and other system components to meet the fuel cell system requirements at the minimum and maximum expected ambient temperatures, in this case 5 and 40 °C, and high altitude, while simultaneously choosing a correct control strategy are important parameters for efficient vehicle power train management.

  9. 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

  10. Effect of Intake Air Filter Condition on Vehicle Fuel Economy

    Energy Technology Data Exchange (ETDEWEB)

    Norman, Kevin M [ORNL; Huff, Shean P [ORNL; West, Brian H [ORNL

    2009-02-01

    The U.S. Department of Energy (DOE) Office of Energy Efficiency and Renewable Energy and the U.S. Environmental Protection Agency (EPA) jointly maintain a fuel economy website (www.fueleconomy.gov), which helps fulfill their responsibility under the Energy Policy Act of 1992 to provide accurate fuel economy information [in miles per gallon (mpg)] to consumers. The site provides information on EPA fuel economy ratings for passenger cars and light trucks from 1985 to the present and other relevant information related to energy use such as alternative fuels and driving and vehicle maintenance tips. In recent years, fluctuations in the price of crude oil and corresponding fluctuations in the price of gasoline and diesel fuels have renewed interest in vehicle fuel economy in the United States. (User sessions on the fuel economy website exceeded 20 million in 2008 compared to less than 5 million in 2004 and less than 1 million in 2001.) As a result of this renewed interest and the age of some of the references cited in the tips section of the website, DOE authorized the Oak Ridge National Laboratory (ORNL) Fuels, Engines, and Emissions Research Center (FEERC) to initiate studies to validate and improve these tips. This report documents a study aimed specifically at the effect of engine air filter condition on fuel economy. The goal of this study was to explore the effects of a clogged air filter on the fuel economy of vehicles operating over prescribed test cycles. Three newer vehicles (a 2007 Buick Lucerne, a 2006 Dodge Charger, and a 2003 Toyota Camry) and an older carbureted vehicle were tested. Results show that clogging the air filter has no significant effect on the fuel economy of the newer vehicles (all fuel injected with closed-loop control and one equipped with MDS). The engine control systems were able to maintain the desired AFR regardless of intake restrictions, and therefore fuel consumption was not increased. The carbureted engine did show a decrease in

  11. 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

  12. 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

  13. Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

    Energy Technology Data Exchange (ETDEWEB)

    Boussard, F.; Huillery, R. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Combustibles; Averseng, J.L.; Serpantie, J.P. [Novatome Industries, 92 - Le Plessis-Robinson (France)

    1994-12-31

    This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs.

  14. Spent fuels conditioning and irradiated nuclear fuel elements examination: the STAR facility and its abilities

    International Nuclear Information System (INIS)

    Boussard, F.; Huillery, R.

    1994-01-01

    This paper is a presentation of the STAR facility, a high activity laboratory located in Cadarache Nuclear Research Center (France). The purpose of the STAR facility and of the associated processes, is the treatment, cleaning and conditioning of spent fuels from Gas Cooled Reactors (GCR) and in particular of about 2300 spent GCR fuel cartridges irradiated more than 20 years ago in Electricite de France (EDF) or CEA Uranium Graphite GCR. The processes are: to separate the nuclear fuel from the clad remains, to chemically stabilize the nuclear material and to condition it in sealed canisters. An additional objective of STAR consists in non-destructive or destructive examinations and tests on PWR rods or FBR pins in the frame of fuel development programs. The paper describes the STAR facility conceptual design (safety design rules, hot cells..) and the different options corresponding to the GCR reconditioning process and to further research and development works on various fuel types. (J.S.). 3 figs

  15. Analysis of WWER-440 fuel performance under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gunduz, Oe; Koese, S; Akbas, T [Atomenerjisi Komisyonu, Ankara (Turkey); Colak, Ue [Ankara Nuclear Research and Training Center (Turkey)

    1994-12-31

    FRAPCON-2 code originally developed for LWR fuel behaviour simulation is used to analyse the WWER-440 fuel rod behaviour at normal operational conditions. The code is capable of utilizing different models for mechanical analysis and gas release calculations. Heat transfer calculations are accomplished through a collocation technique by the method of weighted residuals. Temperature and burnup element properties are evaluated using MATPRO package. As the material properties of Zr-1%Nb used as cladding in WWER-440s are not provided in the code, Zircaloy-4 is used as a substitute for Zr-1%Nb. Mac-Donald-Weisman model is used for gas release calculation. FRACAS-1 and FRACAS-2 models are used in the mechanical calculations. It is assumed that the reactor was operated for 920 days (three consecutive cycles), the burnup being 42000 Mwd/t U. Results of the fuel rod behaviour analysis are given for three axial nodes: bottom node, central node and top node. The variations of the following characteristic fuel rod parameters are studied through the prescribed power history: unmoved gap thickness, gap heat transfer coefficient, fuel axial elongation, cladding axial elongation, fuel centerline temperature and ZrO-thickness at cladding surface. The value of each parameter is calculated as a function of the effective power days for the three nodes by using FRACAS-1 and FRACAS-2 codes for comparison.The results show that calculations with deformable pellet approximation with FRACAS-II model could provide better information for the behaviour of a typical fuel rod. Calculations indicate that fuel rod failure is not observed during the operation. All fuel rod parameters investigated are found to be within the safety limits. It is concluded, however, that for better assessment of reactor safety these calculations should be extended for transient conditions such as LOCA. 1 tab., 10 figs., 4 refs.

  16. Conditioning of metallic Magnox fuel element debris

    International Nuclear Information System (INIS)

    Kaye, C.J.

    1983-01-01

    The conditioning of metallic Magnox debris poses particular problems arising from its chemical reactivity and from the presence in discrete amounts of highly radioactive components. The treatment of this waste is currently being studied by the Central Electricity Generating Board. Following retrieval from store it is envisaged that the debris will be dried and comminuted to facilitate the removal for further storage of the highly active components from the bulk debris. A satisfactory means of sorting the debris appears to be by magnetic induction. The relatively low activity but potentially reactive Magnox will then be directly encapsulated prior to disposal off-site. Currently the only disposal route open for this waste is to the deep ocean. Matrices for encapsulating Magnox have been developed and others are under investigation. The desirable features of such matrices include low chemical reactivity and impermeability to water. The methods used to characterize the resultant waste forms and the results obtained are presented. Thermosetting polymers produce suitable waste forms for sea disposal, exhibiting high mechanical strength and resistance to leaching, and possessing very low chemical reactivity with respect to the Magnox waste. Low viscosity matrices are advantageous from the point of view of the process plant engineering as they enable the comminuted waste to be directly encapsulated. (author)

  17. Conditioning of spent nuclear fuel for permanent disposal

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1994-01-01

    A compact, efficient method for conditioning spent nuclear fuel is under development This method, known as pyrochemical processing, or open-quotes pyroprocessing,close quotes provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the United States Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and preclude the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

  18. Conditioning of spent nuclear fuel for permanent disposal

    International Nuclear Information System (INIS)

    Laidler, J.J.

    1994-01-01

    A compact, efficient method for conditioning spent nuclear fuel is under development. This method, known as pyrochemical processing, or pyroprocessing, provides a separation of fission products from the actinide elements present in spent fuel and further separates pure uranium from the transuranic elements. The process can facilitate the timely and environmentally-sound treatment of the highly diverse collection of spent fuel currently in the inventory of the US Department of Energy (DOE). The pyroprocess utilizes elevated-temperature processes to prepare spent fuel for fission product separation; that separation is accomplished by a molten salt electrorefining step that provides efficient (> 99.9%) separation of transuranics. The resultant waste forms from the pyroprocess are stable under envisioned repository environment conditions and highly leach-resistant. Treatment of any spent fuel type produces a set of common high-level waste forms, one a mineral and the other a metal alloy, that can be readily qualified for repository disposal and that avoid the substantial costs that would be associated with the qualification of the numerous spent fuel types included in the DOE inventory

  19. Thermal conditions and functional requirements for molten fuel containment

    International Nuclear Information System (INIS)

    Kang, C.S.; Torri, A.

    1980-05-01

    This paper discusses the configuration and functional requirements for the molten fuel containment system (MFCS) in the GCFR demonstration plant design. Meltdown conditions following a loss of shutdown cooling (LOSC) accident were studied to define the core debris volume for a realistic meltdown case. Materials and thicknesses of the molten fuel container were defined. Stainless steel was chosen as the sacrificial material and magnesium oxide was chosen as the crucible material. Thermal conditions for an expected quasi-steady state were analyzed. Highlights of the functional requirements which directly affect the MFCS design are discussed

  20. 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.

  1. Safeguards approach for conditioning facility for spent fuel

    International Nuclear Information System (INIS)

    Younkin, J.M.; Barham, M.; Moran, B.W.

    1999-01-01

    A safeguards approach has been developed for conditioning facilities associated with the final disposal of spent fuel in geologic repositories. The proposed approach is based on a generic conditioning facility incorporating common features of conditioning facility designs currently proposed. The generic facility includes a hot cell for consolidation of spent fuel pins and repackaging of spent fuel items such as assemblies and cans of pins. The consolidation process introduces safeguards concerns which have not previously been addressed in traditional safeguards approaches. In developing the safeguards approach, diversion of spent fuel was assessed in terms of potential target items, operational activities performed on the items, containment of the items, and concealment activities performed on the items. The combination of these factors defines the potential diversion pathways. Diversion pathways were identified for spent fuel pellets, pins, assemblies, canisters, and casks. Diversion activities provide for opportunities of detection along the diversion paths. Potential detection methods were identified at several levels of diversion activities. Detection methods can be implemented through safeguards measures. Safeguards measures were proposed for each of the primary safeguards techniques of design information verification (DIV), containment and surveillance (C/S), and material accountancy. Potential safeguards approaches were developed by selection of appropriate combinations of safeguards measures. For all candidate safeguards approaches, DIV is a fundamental component. Variations in the approaches are mainly in the degree of C/S measures and in the types and numbers of material accountancy verification measures. The candidate safeguards approaches were evaluated toward the goal of determining a model safeguards approach. This model approach is based on the integrated application of selected safeguards measures to use International Atomic Energy Agency resources

  2. Structural damage monitoring of harbor caissons with interlocking condition

    Energy Technology Data Exchange (ETDEWEB)

    Huynh, Thanh Canh; Lee, So Young; Nauyen, Khac Duy; Kim, Jeong Tae [Pukyong National Univ., Busan (Korea, Republic of)

    2012-12-15

    The objective of this study is to monitor the health status of harbor caissons which have potential foundation damage. To obtain the objective, the following approaches are performed. Firstly, a structural damage monitoring(SDM) method is designed for interlocked multiple caisson structures. The SDM method utilizes the change in modal strain energy to monitor the foundation damage in a target caisson unit. Secondly, a finite element model of a caisson system which consists of three caisson units is established to verify the feasibility of the proposed method. In the finite element simulation, the caisson units are constrained each other by shear key connections. The health status of the caisson system against various levels of foundation damage is monitored by measuring relative modal displacements between the adjacent caissons.

  3. Structural damage monitoring of harbor caissons with interlocking condition

    International Nuclear Information System (INIS)

    Huynh, Thanh Canh; Lee, So Young; Nauyen, Khac Duy; Kim, Jeong Tae

    2012-01-01

    The objective of this study is to monitor the health status of harbor caissons which have potential foundation damage. To obtain the objective, the following approaches are performed. Firstly, a structural damage monitoring(SDM) method is designed for interlocked multiple caisson structures. The SDM method utilizes the change in modal strain energy to monitor the foundation damage in a target caisson unit. Secondly, a finite element model of a caisson system which consists of three caisson units is established to verify the feasibility of the proposed method. In the finite element simulation, the caisson units are constrained each other by shear key connections. The health status of the caisson system against various levels of foundation damage is monitored by measuring relative modal displacements between the adjacent caissons

  4. 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

  5. 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

  6. Impact of Fuel Type on the Internal Combustion Engine Condition

    Directory of Open Access Journals (Sweden)

    Zdravko Schauperl

    2012-07-01

    Full Text Available The paper studies the influence of liquefied petroleum gas as alternative fuel on the condition of the internal combustion engine. The traffic, energy, economic and ecological influence as well as the types of fuel are studied and analyzed in an unbiased manner, objectively, and in detail, and the obtained results are compared with the condition of the engine of a vehicle powered by the stipulated fuel, petrol Eurosuper 95. The study was carried out on two identical passenger cars with one being fitted with gas installation. The obtained results show that properly installed gas installations in vehicles and the usage of LPG have no significant influence on the driving performances, but they affect significantly the ecological and economic parameters of using passenger cars.

  7. Corrosion mechanisms of spent fuel under oxidizing conditions

    International Nuclear Information System (INIS)

    Finn, P.A.; Finch, R.; Buck, E.; Bates, J.

    1997-01-01

    The release of 99 Tc can be used as a reliable marker for the extent of spent oxide fuel reaction under unsaturated high-drip-rate conditions at 90 degrees C. Evidence from leachate data and from scanning and transmission electron microscopy (SEM and TEM) examination of reacted fuel samples is presented for radionuclide release, potential reaction pathways, and the formation of alteration products. In the ATM-103 fuel, 0.03 of the total inventory of 99 Tc is released in 3.7 years under unsaturated and oxidizing conditions. Two reaction pathways that have been identified from SEM are (1) through-grain dissolution with subsequent formation of uranyl alteration products, and (2) grain-boundary dissolution. The major alteration product identified by x-ray diffraction (XRD) and SEM, is Na-boltwoodite, Na[(UO 2 )(SiO 3 OH)]lg-bullet H 2 O, which is formed from sodium and silicon in the water leachant

  8. 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

  9. 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

  10. 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

  11. 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

  12. Improving of spent fuel monitoring in condition of Slovak wet interim spent fuel storage facility

    International Nuclear Information System (INIS)

    Miklos, M.; Krsjak, V.; Bozik, M.; Vasina, D.

    2008-01-01

    Monitoring of WWER fuel assemblies condition in Slovakia is presented in the paper. The leak tightness results of fuel assemblies used in Slovak WWER units in last 20 years are analyzed. Good experiences with the 'Sipping system' are described. The Slovak wet interim spent fuel storage facility in NPP Jaslovske Bohunice was build and put in operation in 1986. Since 1999, leak tests of WWER-440 fuel assemblies are provided by special leak tightness detection system 'Sipping in Pool' delivered by Framatome-ANP facility with external heating for the precise detection of active specimens. Another system for monitoring of fuel assemblies condition was implemented in December 2006 under the name 'SVYPP-440'. First non-active tests started at February 2007 and are described in the paper. Although those systems seems to be very effective, the detection time of all fuel assemblies in one storage pool is too long (several months). Therefore, a new 'on-line' detection system, based on new sorbent KNiFC-PAN for effective 134 Cs and 137 Cs activity was developed. This sorbent was compared with another type of sorbent NIFSIL and results are presented. The design of this detection system and its possible application in the Slovak wet spent fuel storage facility is discussed. For completeness, the initial results of the new system are also presented. (authors)

  13. 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

  14. Structural analysis of advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    Gu, J. H.; Jung, W. M.; Jo, I. J.; Gug, D. H.; Yoo, K. S.

    2003-01-01

    An advanced spent fuel conditioning process (ACP) is developing for the safe and effective management of spent fuels which arising from the domestic nuclear power plants. And its demonstration facility is under design. This facility will be prepared by modifying IMEF's reserve hot cell facility which reserved for future usage by considering the characteristics of ACP. This study presents a basic structural architecture design and analysis results of ACP hot cell including modification of the IMEF. The results of this study will be used for the detail design of ACP demonstration facility, and utilized as basic data for the licensing of the ACP facility

  15. Fuel conditioning facility electrorefiner cadmium vapor trap operation

    International Nuclear Information System (INIS)

    Vaden, D. E.

    1998-01-01

    Processing sodium-bonded spent nuclear fuel at the Fuel Conditioning Facility at Argonne National Laboratory-West involves an electrometallurgical process employing a molten LiCl-KCl salt covering a pool of molten cadmium. Previous research has shown that the cadmium dissolves in the salt as a gas, diffuses through the salt layer and vaporizes at the salt surface. This cadmium vapor condenses on cool surfaces, causing equipment operation and handling problems. Using a cadmium vapor trap to condense the cadmium vapors and reflux them back to the electrorefiner has mitigated equipment problems and improved electrorefiner operations

  16. 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

  17. Methods and Piezoelectric Imbedded Sensors for Damage Detection in Composite Plates Under Ambient and Cryogenic Conditions

    Science.gov (United States)

    Engberg, Robert; Ooi, Teng K.

    2004-01-01

    New methods for structural health monitoring are being assessed, especially in high-performance, extreme environment, safety-critical applications. One such application is for composite cryogenic fuel tanks. The work presented here attempts to characterize and investigate the feasibility of using imbedded piezoelectric sensors to detect cracks and delaminations under cryogenic and ambient conditions. A variety of damage detection methods and different Sensors are employed in the different composite plate samples to aid in determining an optimal algorithm, sensor placement strategy, and type of imbedded sensor to use. Variations of frequency, impedance measurements, and pulse echoing techniques of the sensors are employed and compared. Statistical and analytic techniques are then used to determine which method is most desirable for a specific type of damage. These results are furthermore compared with previous work using externally mounted sensors. Results and optimized methods from this work can then be incorporated into a larger composite structure to validate and assess its structural health. This could prove to be important in the development and qualification of any 2" generation reusable launch vehicle using composites as a structural element.

  18. Computational simulation of the microstructure of irradiation damaged regions for the plate type fuel of UO2 microspheres dispersed in stainless steel matrix

    International Nuclear Information System (INIS)

    Reis, S.C. dos; Lage, A.F.; Braga, D.; Ferraz, W.B.

    2006-01-01

    Plate type fuel elements have high efficiency of thermal transference what benefits the heat flux with high rates of power output. In reactor cores, fuel elements, in general, are subject to a high neutrons flux, high working temperatures, severe corrosion conditions, direct interference of fission products that result from nuclear reactions and radiation interaction-matter. For plate type fuels composed of ceramic particles dispersed in metallic matrix, one can observe the damage regions that arise due to the interaction fission products in the metallic matrix. Aiming at evaluating the extension of the damage regions in function of the particles and its diameters, in this paper, computational geometric simulations structure of plate type fuel cores, composed of UO 2 microspheres dispersed in stainless steel in several fractions of volume and diameters were carried out. The results of the simulations were exported to AutoCAD R where it was possible its visualization and analysis. (author)

  19. 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

  20. Safeguarding of spent fuel conditioning and disposal in geological repositories

    International Nuclear Information System (INIS)

    Forsstroem, H.; Richter, B.

    1997-01-01

    Disposal of spent nuclear fuel in geological formations, without reprocessing, is being considered in a number of States. Before disposal the fuel will be encapsulated in a tight and corrosion resistant container. The method chosen for disposal and the design of the repository will be determined by the geological conditions and the very strict requirements on long-term safety. From a safeguards perspective spent fuel disposal is a new issue. As the spent fuel still contains important amounts of material under safeguards and as it can not be considered practicably irrecoverable in the repository, the IAEA has been advised not to terminate safeguards, even after closure of the repository. This raises a number of new issues where there could be a potential conflict of interests between safety and safeguards demands, in particular in connection with the safety principle that burdens on future generations should be avoided. In this paper some of these issues are discussed based on the experience gained in Germany and Sweden about the design and future operation of encapsulation and disposal facilities. The most important issues are connected to the required level of safeguards for a closed repository, the differences in time scales for waste management and safeguards, the need for verification of the fissile content in the containers and the possibility of retrieving the fuel disposed of. (author)

  1. Ion irradiation damage in ilmenite under cryogenic conditions

    International Nuclear Information System (INIS)

    Mitchell, J.N.; Yu, N.; Devanathan, R.; Sickafus, K.E.; Nastasi, M.A.

    1996-01-01

    A natural single crystal of ilmenite was irradiated at 100 K with 200 keV Ar 2+ . Rutherford backscattering spectroscopy and ion channeling with 2 MeV He + ions were used to monitor damage accumulation in the surface region of the implanted crystal. At an irradiation fluence of 1 x 10 15 Ar 2+ cm -2 , considerable near-surface He + ion dechanneling was observed, to the extent that ion yield from a portion of the aligned crystal spectrum reached the yield level of a random spectrum. This observation suggests that the near-surface region of the crystal was amorphized by the implantation. Cross-sectional transmission electron microscopy and electron diffraction on this sample confirmed the presence of a 150 mm thick amorphous layer. These results are compared to similar investigations on geikielite (MgTiO 3 ) and spinel (MgAl 2 O 4 ) to explore factors that may influence radiation damage response in oxides

  2. 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

  3. 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

  4. Aqueous alteration of VHTR fuels particles under simulated geological conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ait Chaou, Abdelouahed, E-mail: aitchaou@subatech.in2p3.fr; Abdelouas, Abdesselam; Karakurt, Gökhan; Grambow, Bernd

    2014-05-01

    Very High Temperature Reactor (VHTR) fuels consist of the bistructural-isotropic (BISO) or tristructural-isotropic (TRISO)-coated particles embedded in a graphite matrix. Management of the spent fuel generated during VHTR operation would most likely be through deep geological disposal. In this framework we investigated the alteration of BISO (with pyrolytic carbon) and TRISO (with SiC) particles under geological conditions simulated by temperatures of 50 and 90 °C and in the presence of synthetic groundwater. Solid state (scanning electron microscopy (SEM), micro-Raman spectroscopy, electron probe microanalyses (EPMA) and X-ray photoelectron spectroscopy (XPS)) and solution analyses (ICP-MS, ionique chromatography (IC)) showed oxidation of both pyrolytic carbon and SiC at 90 °C. Under air this led to the formation of SiO{sub 2} and a clay-like Mg–silicate, while under reducing conditions (H{sub 2}/N{sub 2} atmosphere) SiC and pyrolytic carbon were highly stable after a few months of alteration. At 50 °C, in the presence and absence of air, the alteration of the coatings was minor. In conclusion, due to their high stability in reducing conditions, HTR fuel disposal in reducing deep geological environments may constitute a viable solution for their long-term management.

  5. Condition Assessment for Wastewater Pipes: Method for Assessing Cracking and Surface Damage of Concrete Pipes

    OpenAIRE

    Hauge, Petter

    2013-01-01

    The objective of the Master Thesis has been to provide an improved method for condition assessment, which will give a better correlation between Condition class and actual Condition of concrete pipes with cracking and/or surface damages. Additionally improvement of the characterization of cracking (SR) and surface (KO) damages was a sub goal.Based on the findings described in my Thesis and my Specialization Project (Hauge 2012), I recommend that the Norwegian condition assessment method based...

  6. Nuclear Fuel Fretting Mechanisms in a Room Temperature Unlubricated Condition

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Ho; Kim, Hyung Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    Recently, efforts for evaluating the fretting wear mechanism have been carried out by many researchers in various conditions. In an unlubricated condition, especially, effects of a wear debris and/or its layer on the fretting wear behavior were proposed that the formation of a well-developed glaze layer has a beneficial effect for decreasing a friction coefficient. Otherwise, a wear rate was accelerated by a third-body abrasion. At this time, it is well known that wear debris behaviors are affected by test variables such as a temperature, environment, material characteristics, etc. In a nuclear fuel fretting, however, its contact condition is quite different when compared with general fretting wear studies and could be summarized as the following; first, a fuel rod is supported by spacer grid springs and dimples that were elastically deformable. This results in a unique friction loop and a different fretting mechanism when a fuel rod is vibrated due to a flow-induced vibration (FIV). Next, it is possible that some region of the wear scar area with a specific spring shape condition could be hidden due to different wear debris behavior. So, some of the wear debris layers could be found on the worn surfaces in previous studies even though fretting wear tests were performed in a water lubricated condition. Finally, initial contact condition could be changed both an actual operating condition in power plants (i.e. high temperature and pressurized water (HTHP) under severe irradiation conditions) and the fretting wear tests for evaluating the wear resistant spring in lab conditions (i.e. from room temperature to HTHP without irradiation conditions) due to material degradations and the formation of the wear scar, respectively. In summary, the spring shape effect and the variation of the contact condition with increasing fretting cycle should be evaluated in order to improve the wear resistance of the spacer grid spring. So, in this study, fretting wear tests have been

  7. Reduction of uranium in disposal conditions of spent nuclear fuel

    International Nuclear Information System (INIS)

    Myllykylae, E.

    2008-02-01

    This literature study is a summary of publications, in which the reduction of uranium by iron has been investigated in anaerobic groundwater conditions or in aqueous solution in general. The basics of the reduction phenomena and the oxidation states, complexes and solubilities of uranium and iron in groundwaters are discussed as an introduction to the subject, as well as, the Finnish disposal concept of spent nuclear fuel. The spent fuel itself mainly (∼96 %) consists of a sparingly soluble uranium(IV) dioxide, UO 2 (s), which is stable phase in the anticipated reducing disposal conditions. If spent fuel gets in contact with groundwater, oxidizing conditions might be induced by the radiolysis of water, or by the intrusion of oxidizing glacial melting water. Under these conditions, the oxidation and dissolution of uranium dioxide to more soluble U(VI) species could occur. This could lead to the mobilization of uranium and other components of spent fuel matrix including fission products and transuranium elements. The reduction of uranium back to oxidation state U(IV) can be considered as a favourable immobilization mechanism in a long-term, leading to precipitation due to the low solubility of U(IV) species. The cast iron insert of the disposal canister and its anaerobic corrosion products are the most important reductants under disposal conditions, but dissolved ferrous iron may also function as reductant. Other iron sources in the buffer or near-field rock, are also considered as possible reductants. The reduction of uranium is a very challenging phenomenon to investigate. The experimental studies need e.g. well-controlled anoxic conditions and measurements of oxidation states. Reduction and other simultaneous phenomena are difficult to distinghuish. The groundwater conditions (pH, Eh and ions) influence on the prevailing complexes of U and Fe and on forming corrosion products of iron and, thus they determine also the redox chemistry. The partial reduction of

  8. 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

  9. Spent fuel management in France: Reprocessing, conditioning, recycling

    International Nuclear Information System (INIS)

    Giraud, J.P.; Montalembert, J.A. de

    1994-01-01

    The French energy policy has been based for 20 years on the development of nuclear power. The some 75% share of nuclear in the total electricity generation, representing an annual production of 317 TWh requires full fuel cycle control from the head-end to the waste management. This paper presents the RCR concept (Reprocessing, Conditioning, Recycling) with its industrial implementation. The long lasting experience acquired in reprocessing and MOX fuel fabrication leads to a comprehensive industrial organization with minimized impact on the environment and waste generation. Each 900 MWe PWR loaded with MOX fuel avoids piling up 2,500 m 3 per year of mine tailings. By the year 2000, less than 500 m 3 of high-level and long-lived waste will be annually produced at La Hague for the French program. The fuel cycle facilities and the associated MOX loading programs are ramping-up according to schedule. Thus, the RCR concept is a reality as well as a policy adopted in several countries. Last but not least, RCR represents a strong commitment to non-proliferation as it is the way to fully control and master the plutonium inventory

  10. 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

  11. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    Energy Technology Data Exchange (ETDEWEB)

    You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P

    2006-09-15

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO{sub 2} into U-metal. For demonstration of this process, {alpha}-{gamma} type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for {gamma}-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration.

  12. Radiation Monitoring System in Advanced Spent Fuel Conditioning Process Facility

    International Nuclear Information System (INIS)

    You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P.

    2006-09-01

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO 2 into U-metal. For demonstration of this process, α-γ type new hot cell was built in the IMEF basement . To secure against radiation hazard, this facility needs radiation monitoring system which will observe the entire operating area before the hot cell and service area at back of it. This system consists of 7 parts; Area Monitor for γ-ray, Room Air Monitor for particulate and iodine in both area, Hot cell Monitor for hot cell inside high radiation and rear door interlock, Duct Monitor for particulate of outlet ventilation, Iodine Monitor for iodine of outlet duct, CCTV for watching workers and material movement, Server for management of whole monitoring system. After installation and test of this, radiation monitoring system will be expected to assist the successful ACP demonstration

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  15. 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)

  16. 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

  17. Solid oxide fuel cell performance under severe operating conditions

    DEFF Research Database (Denmark)

    Koch, Søren; Hendriksen, P.V.; Mogensen, Mogens Bjerg

    2006-01-01

    The performance and degradation of Solid Oxide Fuel Cells (SOFC) were studied under severe operating conditions. The cells studied were manufactured in a small series by ECN, in the framework of the EU funded CORE-SOFC project. The cells were of the anode-supported type with a double layer LSM...... cathode. They were operated at 750 °C or 850 °C in hydrogen with 5% or 50% water at current densities ranging from 0.25 A cm–2 to 1 A cm–2 for periods of 300 hours or more. The area specific cell resistance, corrected for fuel utilisation, ranged between 0.20 Ω cm2 and 0.34 Ω cm2 at 850 °C and 520 m......V, and between 0.51 Ω cm2 and 0.92 Ω cm2 at 750 °C and 520 mV. The degradation of cell performance was found to be low (ranging from 0 to 8%/1,000 hours) at regular operating conditions. Voltage degradation rates of 20 to 40%/1,000 hours were observed under severe operating conditions, depending on the test...

  18. 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.)

  19. 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.

  20. Self-Healing Characteristics of Damaged Rock Salt under Different Healing Conditions

    Directory of Open Access Journals (Sweden)

    Lin Li

    2013-08-01

    Full Text Available Salt deposits are commonly regarded as ideal hosts for geologic energy reservoirs. Underground cavern construction-induced damage in salt is reduced by self-healing. Thus, studying the influencing factors on such healing processes is important. This research uses ultrasonic technology to monitor the longitudinal wave velocity variations of stress-damaged rock salts during self-recovery experiments under different recovery conditions. The influences of stress-induced initial damage, temperature, humidity, and oil on the self-recovery of damaged rock salts are analyzed. The wave velocity values of the damaged rock salts increase rapidly during the first 200 h of recovery, and the values gradually increase toward stabilization after 600 h. The recovery of damaged rock salts is subjected to higher initial damage stress. Water is important in damage recovery. The increase in temperature improves damage recovery when water is abundant, but hinders recovery when water evaporates. The presence of residual hydraulic oil blocks the inter-granular role of water and restrains the recovery under triaxial compression. The results indicate that rock salt damage recovery is related to the damage degree, pore pressure, temperature, humidity, and presence of oil due to the sealing integrity of the jacket material.

  1. 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.

  2. Experimental conditions at Osiris with the new CARAMEL fuel

    International Nuclear Information System (INIS)

    Beylot, J.

    1979-01-01

    Replacing the former highly enriched (93%) U-Al fuel by low enrichment (7%) oxide has brought about some changes in the experimental conditions for irradiations. The advantages for the experiments placed right in the lattice are shown to be a great improvement in the neutron spectrum (fast/thermal) and a very significant reduction in heating due to gamma radiation. In the case of the peripherally placed experiments there is an increase in the number of high thermal flux sites. In all cases, there is found to be an increase in the duration of the irradiation cycle, between two partial reloadings, permitted by the significant amount of 235 U tied up in the loading. The drawbacks observed are reduced thermo-hydraulic performance of the new fuel elements that does not allow working with a core of a size under a 7x7 configuration, increased surveillance of the kind of experiments placed in the lattice to avoid excessive power rises on the neighbouring fuel elements and moderate reduction in the level of thermal neutron fluxes in the peripheral irradiation sites [fr

  3. Smelting Associated with the Advanced Spent Fuel Conditioning Process

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  4. 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

  5. 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.

  6. 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.)

  7. Material Control and Accountability Experience at the Fuel Conditioning Facility

    International Nuclear Information System (INIS)

    Vaden, D.; Fredrickson, G.L.

    2007-01-01

    The Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) treats spent nuclear fuel using an electrometallurgical process that separates the uranium from the fission products, sodium thermal bond, and cladding materials. Material accountancy is necessary at FCF for two reasons: 1) it provides a mechanism for detecting a potential loss of nuclear material for safeguards and security, and 2) it provides a periodic check of inventories to ensure that processes and materials are within control limits. Material Control and Accountability is also a Department of Energy (DOE) requirement (DOE Order 474.1). The FCF employs a computer based Mass Tracking (MTG) System to collect, store, retrieve, and process data on all operations that directly affect the flow of materials through the FCF. The MTG System is important for the operations of the FCF because it supports activities such as material control and accountability, criticality safety, and process modeling. To conduct material control and accountability checks and to monitor process performance, mass balances are routinely performed around the process equipment. The equipment used in FCF for pyro-processing consists of two mechanical choppers and two electro-refiners (the Mark-IV with the accompanying element chopper and Mark-V with the accompanying blanket chopper for processing driver fuel and blanket, respectively), and a cathode processor (used for processing both driver fuel and blanket) and casting furnace (mostly used for processing driver fuel). Performing mass balances requires the measurement of the masses and compositions of several process streams and equipment inventories. The masses of process streams are obtained via in-cell balances (i.e., load cells) that weigh containers entering and leaving the process equipment. Samples taken at key locations are analyzed to determine the composition of process streams and equipment inventories. In cases where equipment or containers cannot be

  8. Modelling spent fuel and HLW behaviour in repository conditions

    Energy Technology Data Exchange (ETDEWEB)

    Esparza, A M; Esteban, J A

    2003-07-01

    The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

  9. Modelling spent fuel and HLW behaviour in repository conditions

    International Nuclear Information System (INIS)

    Esparza, A. M.; Esteban, J. A.

    2003-01-01

    The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

  10. 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.)

  11. 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.

  12. Methods of conditioning waste fuel decladding hulls and dissolver residues

    International Nuclear Information System (INIS)

    De Regge, P.; Loida, A.; Schmidt-Hansberg, T.; Sombret, C.

    1985-01-01

    Several methods for conditioning spent fuel decladding hulls or dissolver residues have been considered in various countries of the European Community. Five of these methods used embedding technique with or without prior compaction: they are based on incorporation in metallic alloys, glass, ceramics, cements and metals or graphite compounds. A sixth one consists in melting the decladding materials. The corresponding research programmes have been pursued to varying states of progress with regard to demonstrating their feasibility on an industrial scale and the use of genuine wastes in bench scale experiments. The properties of the conditioned wastes have been investigated. Special attention has been paid to the corrosion resistance to various aqueous media as tap water, brine or clayey water. Although no categorical conclusion can be drawn from the initial results, the available findings provide a basis for assessing the different processes

  13. Experimental observation of dynamic ductile damage development under various triaxiality conditions - description of the principle

    Directory of Open Access Journals (Sweden)

    Pillon L.

    2012-08-01

    Full Text Available The Gurson model has been extended by Perrin to describe damage evolution in ductile viscoplastic materials. The so-called Gurson-Perrin model allows representing damage development with respect to strain-rate conditions. In order to fill a lack in current experimental procedures, we propose an experimental project able to test and validate the Gurson-Perrin model under various dynamic conditions and for different stress triaxiality levels.

  14. 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.

  15. Safeguards System for the Advanced Spent Fuel Conditioning Process Facility

    International Nuclear Information System (INIS)

    Kim, Ho-dong; Lee, T.H.; Yoon, J.S.; Park, S.W; Lee, S.Y.; Li, T.K.; Menlove, H.; Miller, M.C.; Tolba, A.; Zarucki, R.; Shawky, S.; Kamya, S.

    2007-01-01

    The advanced spent fuel conditioning process (ACP) which is a part of a pyro-processing has been under development at Korean Atomic Energy Research Institute (KAERI) since 1997 to tackle the problem of an accumulation of spent fuel. The concept is to convert spent oxide fuel into a metallic form in a high temperature molten salt in order to reduce the heat energy, volume, and radioactivity of a spent fuel. Since the inactive tests of the ACP have been successfully implemented to confirm the validity of the electrolytic reduction technology, a lab-scale hot test will be undertaken in a couple of years to validate the concept. For this purpose, the KAERI has built the ACP Facility (ACPF) at the basement of the Irradiated Material Examination Facility (IMEF) of KAERI, which already has a reserved hot-cell area. Through the bilateral arrangement between US Department of Energy (DOE) and Korean Ministry of Science and Technology (MOST) for safeguards R and D, the KAERI has developed elements of safeguards system for the ACPF in cooperation with the Los Alamos National Laboratory (LANL). The reference safeguards design conditions and equipment were established for the ACPF. The ACPF safeguards system has many unique design specifications because of the particular characteristics of the pyro-process materials and the restrictions during a facility operation. For the material accounting system, a set of remote operation and maintenance concepts has been introduced for a non-destructive assay (NDA) system. The IAEA has proposed a safeguards approach to the ACPF for the different operational phases. Safeguards measures at the ACPF will be implemented during all operational phases which include a 'Cold Test', a 'Hot Test' and at the end of a 'Hot test'. Optimization of the IAEA's inspection efforts was addressed by designing an effective safeguards approach that relies on, inter alia, remote monitoring using cameras, installed NDA instrumentation, gate monitors and seals

  16. Comparison of Fuel-Nox Formation Characteristics in Conventional Air and Oxy fuel Combustion Conditions

    International Nuclear Information System (INIS)

    Woo, Mino; Park, Kweon Ha; Choi, Byung Chul

    2013-01-01

    Nitric oxide (NO x ) formation characteristics in non-premixed diffusion flames of methane fuels have been investigated experimentally and numerically by adding 10% ammonia to the fuel stream, according to the variation of the oxygen ratio in the oxidizer with oxygen/carbon dioxide and oxygen/nitrogen mixtures. In an experiment of co flow jet flames, in the case of an oxidizer with oxygen/carbon dioxide, the NO x emission increased slightly as the oxygen ratio increased. On the other hand, in case of an oxygen/nitrogen oxidizer, the NO x emission was the maximum at an oxygen ratio of 0.7, and it exhibited non-monotonic behavior according to the oxygen ratio. Consequently, the NO x emission in the condition of oxy fuel combustion was overestimated as compared to that in the condition of conventional air combustion. To elucidate the characteristics of NO x formation for various oxidizer compositions, 1a and 2a numerical simulations have been conducted by adopting one kinetic mechanism. The result of 2 simulation for an oxidizer with oxygen/nitrogen well predicted the trend of experimentally measured NO x emissions

  17. 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.

  18. 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.

  19. 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

  20. 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.

  1. Serviceability of rod ceramic fuel pins on motoring conditions of FTP or NEMF reactor

    International Nuclear Information System (INIS)

    Deryavko, I.I.

    2004-01-01

    The operation conditions of rod ceramic fuel pins in the running hydrogen-cooled technological canals of FTP or NEMF reactor on the motoring conditions are considered. The available postreactor researches of the fuel pins are presented and the additional postreactor researches of fuel pins, tested on this mode in IVG.1 and IRGIT reactors, are carried out. The fuel pins serviceability on motoring conditions of FTP or NEF reactor operation is concluded. (author)

  2. 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

  3. 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

  4. Material accountancy in an electrometallurgical Fuel Conditioning Facility

    International Nuclear Information System (INIS)

    Vaden, D.; Benedict, R.W.; Goff, K.M.; Keyes, R.W.; Mariani, R.D.; Bucher, R.G.; Yacout, A.M.

    1996-01-01

    The Fuel Conditioning Facility (FCF) treats spent nuclear fuel using an electrometallurgical process that separates the uranium from the fission products, sodium thermal bond and cladding materials. Material accountancy is necessary at FCF for two reasons: first, it provides a mechanism for detecting a potential loss of nuclear material for safeguards and security; second, it provides a periodic check of inventories to ensure that processes and material are under control. By weighing material entering and leaving a process, and using sampling results to determine composition, an inventory difference (ID) results when the measured inventory is compared to the predicted inventory. The ID and its uncertainty, based on error propagation, determines the degree of assurance that an operation proceeded according to expectations. FCF uses the ID calculation in two ways: closeout, which is the ID and uncertainty for a particular operational step, and material accountancy, which determines an ID and its associated uncertainty for a material balance area through several operational steps. Material accountancy over the whole facility for a specified time period assists in detecting diversion of nuclear material. Data from depleted uranium operations are presented to illustrate the method used in FCF

  5. Fuel conditioning facility zone-to-zone transfer administrative controls

    International Nuclear Information System (INIS)

    Pope, C. L.

    2000-01-01

    The administrative controls associated with transferring containers from one criticality hazard control zone to another in the Argonne National Laboratory (ANL) Fuel Conditioning Facility (FCF) are described. FCF, located at the ANL-West site near Idaho Falls, Idaho, is used to remotely process spent sodium bonded metallic fuel for disposition. The process involves nearly forty widely varying material forms and types, over fifty specific use container types, and over thirty distinct zones where work activities occur. During 1999, over five thousand transfers from one zone to another were conducted. Limits are placed on mass, material form and type, and container types for each zone. Ml material and containers are tracked using the Mass Tracking System (MTG). The MTG uses an Oracle database and numerous applications to manage the database. The database stores information specific to the process, including material composition and mass, container identification number and mass, transfer history, and the operators involved in each transfer. The process is controlled using written procedures which specify the zone, containers, and material involved in a task. Transferring a container from one zone to another is called a zone-to-zone transfer (ZZT). ZZTs consist of four distinct phases, select, request, identify, and completion

  6. Microbial Condition of Water Samples from Foreign Fuel Storage Facilities

    International Nuclear Information System (INIS)

    Berry, C.J.

    1998-01-01

    In order to assess the microbial condition of foreign spent nuclear fuel storage facilities and their possible impact on SRS storage basins, twenty-three water samples were analyzed from 12 different countries. Fifteen of the water samples were analyzed and described in an earlier report (WSRC-TR-97-00365 [1]). This report describes nine additional samples received from October 1997 through March 1998. The samples include three from Australia, two from Denmark and Germany and one sample from Italy and Greece. Each water sample was analyzed for microbial content and activity as determined by total bacteria, viable aerobic bacteria, viable anaerobic bacteria, viable sulfate-reducing bacteria, viable acid-producing bacteria and enzyme diversity. The results for each water sample were then compared to all other foreign samples analyzed to date and monthly samples pulled from the receiving basin for off-site fuel (RBOF), at SRS. Of the nine samples analyzed, four samples from Italy, Germany and Greece had considerably higher microbiological activity than that historically found in the RBOF. This microbial activity included high levels of enzyme diversity and the presence of viable organisms that have been associated with microbial influenced corrosion in other environments. The three samples from Australia had microbial activities similar to that in the RBOF while the two samples from Denmark had lower levels of microbial activity. These results suggest that a significant number of the foreign storage facilities have water quality standards that allow microbial proliferation and survival

  7. Safety analysis of spent fuel transport and storage casks under extreme impact conditions

    International Nuclear Information System (INIS)

    Wolff, D.; Wieser, G.; Ballheimer, V.; Voelzke, H.; Droste, B.

    2005-01-01

    Full text: Worldwide the security of transport and storage of spent fuel with respect to terrorism threats is a matter of concern. In Germany a spent nuclear fuel management program was developed by the government including a new concept of dry on-site interim storage instead of centralized interim storage. In order to minimize transports of spent fuel casks between nuclear power plants, reprocessing plants and central storage facilities, the operators of NPPs have to erect and to use interim storage facilities for spent nuclear fuel on the site or in the vicinity of nuclear power plants. Up to now, 11 on-site interim storage buildings, one storage tunnel and 4 on-site interim storage areas (preliminary cask storage till the on-site interim storage building is completed) have been licensed at 12 nuclear power plant sites. Inside the interim storage buildings the casks are kept in upright position, whereas at the preliminary interim storage areas horizontal storage of the casks on concrete slabs is used and each cask is covered by concrete elements. Storage buildings and concrete elements are designed only for gamma and neutron radiation shielding reasons and as weather protection. Therefore the security of spent fuel inside a dual purpose transport and storage cask depends on the inherent safety of the cask itself. For nearly three decades BAM has been investigating cask safety under severe accident conditions like drop tests from more than 9 m onto different targets and without impact limiters as well as artificially damaged prototype casks. Since the terror attacks of 11 September 2001 the determination of casks' inherent safety also under extreme impact conditions due to terrorist attacks has been of our increasing interest. With respect to spent fuel storage one of the most critical scenarios of a terrorist attack for a cask is the centric impact of a dynamic load onto the lid-seal-system caused e.g. by direct aircraft crash or its engine as well as by a

  8. Severe fuel damage experiments performed in the QUENCH facility with 21-rod bundles of LWR-type

    International Nuclear Information System (INIS)

    Sepold, L.; Hering, W.; Schanz, G.; Scholtyssek, W.; Steinbrueck, M.; Stuckert, J.

    2006-01-01

    The objective of the QUENCH experimental program at the Karlsruhe Research Center is to investigate core degradation and the hydrogen source term that results from quenching/flooding an uncovered core, to examine the physical/chemical behavior of overheated fuel elements under different flooding conditions, and to create a data base for model development and improvement of severe fuel damage (SFD) code systems. The large-scale 21-rod bundle experiments conducted in the QUENCH out-of-pile facility are supported by an extensive separate-effects test program, by modeling activities as well as application and improvement of SFD code systems. International cooperations exist with institutions mainly within the European Union but e.g. also with the Russian Academy of Science (IBRAE, Moscow) and the CSARP program of the USNRC. So far, eleven experiments have been performed, two of them with B 4 C absorber material. Experimental parameters were: the temperature at initiation of reflood, the degree of peroxidation, the quench medium, i.e. water or steam, and its injection rate, the influence of a B 4 C absorber rod, the effect of steam-starved conditions before quench, the influence of air oxidation before quench, and boil-off behavior of a water-filled bundle with subsequent quenching. The paper gives an overview of the QUENCH program with its organizational structure, describes the test facility and the test matrix with selected experimental results. (author)

  9. ISP-31 OECD/NEA/CSNI International Standard Problem n.31. Cora-13 experiment on severe fuel damage. Comparison report

    International Nuclear Information System (INIS)

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

    1993-07-01

    The severe fuel damage experiment CORA-13 has been offered as CSNI-International Standard Problem (ISP) No. 31. The out-of-pile experiment CORA-13 was executed in November 1990 at Kernforschungszentrum Karlsruhe. The major objectives of this 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. The ISP was conducted as a blind exercise. 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. Results to the ISP were submitted by 9 participants using different versions of SCDAP/RELAP5, and codes such as FRAS-SFD, ICARE2, KESS-III, MELCOR. The thermal behavior up to significant oxidation has been predicted quite well by most of the codes. In general, 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

  10. Experimental program on fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Languille, A.; Cecchi, P.

    1985-01-01

    During LMFBR plant operation, fuel developments are primarily concerned with the fuel pin irradiation behaviour under steady-state conditions up to high burn-up levels. But additional studies under off-normal conditions are necessary in order to assess fuel pin performance and to define operational limits. (author)

  11. Dry Storage at long term of nuclear fuels: Influence of the fuel design and commercial irradiation conditions

    International Nuclear Information System (INIS)

    Marino, Armando C

    2009-01-01

    The BaCo code was applied to simulate the behaviour for a PHWR fuel under storage conditions showing a strong dependence on the original design of the fuel and the irradiation history. In particular, the results of the statistical analysis of BaCo indicate that the integrity of the fuel is influenced by the manufacture tolerances and the solicitations during the NPP irradiation. The main conclusion of the present study is that the fuel temperature of the device should be carefully controlled in order to ensure safe storage conditions. [es

  12. 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

  13. The significance of the pilot conditioning plant (PKA) for spent fuel management

    International Nuclear Information System (INIS)

    Willax, H.O.

    1996-01-01

    The pilot conditioning plant (PKA) is intended as a multi-purpose facility and thus may serve various purposes involved in the conditioning or disposal of spent fuel elements or radwaste. Its design as a pilot plant permits development and trial of various methods and processes for fuel element conditioning, as well as for radwaste conditioning. (orig./DG) [de

  14. The risk of PCI damage to 8x8 fuel rods during limit cycle instability

    Energy Technology Data Exchange (ETDEWEB)

    Schrire, D.; Oguma, R.; Malen, K.

    1994-12-31

    A BWR reactor core may experience thermal-hydraulic instability under certain operating conditions. Generally, the instability results in neutron flux (i e generated neutronic power) and coolant flow and pressure oscillations, which reach a maximum `limit cycle` amplitude. The cladding response to power transients has been studied using noise analysis. These results have been compared to results from code calculations using the fuel code TOODEE 2. From these results the risk for fuel rod failure due to pellet-clad mechanical interaction and possible failure due to stress corrosion cracking (PCI) has been estimated. It turns out that for the oscillation frequencies of interest (0,3-0,5 Hz) the fuel response amplitude reduction makes PCI-failure improbable. 17 refs.

  15. Mixing Characteristics during Fuel Coolant Interaction under Reactor Submerged Conditions

    International Nuclear Information System (INIS)

    Hong, S. W.; Na, Y. S.; Hong, S. H.; Song, J. H.

    2014-01-01

    A molten material is injected into an interaction chamber by free gravitation fall. This type of fuel coolant interaction could happen to operating plants. However, the flooding of a reactor cavity is considered as SAM measures for new PWRs such as APR-1400 and AP1000 to assure the IVR of a core melt. In this case, a molten corium in a reactor is directly injected into water surrounding the reactor vessel without a free fall. KAERI has carried out fuel coolant interaction tests without a free fall using ZrO 2 and corium to simulate the reactor submerged conditions. There are four phases in a steam explosion. The first phase is a premixing phase. The premixing is described in the literature as follows: during penetration of melt into water, hydrodynamic instabilities, generated by the velocities and density differences as well as vapor production, induce fragmentation of the melt into particles; the particles fragment in turn into smaller particles until they reach a critical size such that the cohesive forces (surface tension) balance exactly the disruptive forces (inertial); and the molten core material temperature (>2500 K) is such that the mixing always occurs in the film boiling regime of the water: It is very important to qualify and quantify this phase because it gives the initial conditions for a steam explosion This paper mainly focuses on the observation of the premixing phase between a case with 1 m free fall and a case without a free fall to simulate submerged reactor condition. The premixing behavior between a 1m free fall case and reactor case submerged without a free fall is observed experimentally. The average velocity of the melt front passing through 1m water pool; - Case without a free fall: The average velocity of corium, 2.7m/s, is faster than ZrO 2 , 2.3m/s, in water. - Cases of with a 1 m free fall and without a free fall : The case without a free fall is about two times faster than a case with a 1 m free fall. Bubble characteristics; - Case

  16. Corrosion of copper under Canadian nuclear fuel waste disposal conditions

    International Nuclear Information System (INIS)

    King, F.; Litke, C.D.

    1990-01-01

    The corrosion of copper was studied under Canadian nuclear fuel waste disposal conditions. The groundwater in a Canadian waste vault is expected to be saline, with chloride concentrations from 0.1 to 1.0 mol/l. The container would be packed in a sand/clay buffer, and the maximum temperature on the copper surface would be 100C; tests were performed up to 150C. Radiation fields will initially be around 500 rad/h, and conditions will be oxidizing. Sulfides may be present. The minimum design lifetime for the container is 500 years. Most work has been done on uniform corrosion, although pitting has been considered. It was found that the rate of uniform corrosion in aerated NaCl at room temperature is limited by the rate of the anodic reaction, which is controlled mainly by the rate of transport of dissolved metal species away from the copper surface. The rate of corrosion should become controlled by the transport of oxygen to the copper surface only at very low oxygen concentrations. In the presence of gamma radiation the corrosion rate may never become cathodically transport limited. In compacted buffer material, the corrosion rate appears to be limited by the rate of transport of copper species away from the corroding surface. The authors recommend that long-term predictions of container lifetime should be based on the known rate-determining step for the overall corrosion process. 8 refs

  17. Thermal behaviour of fuel: influence on the behavior of fuel elements in nominal and incidental operating conditions

    International Nuclear Information System (INIS)

    Languille, A.

    1984-02-01

    The behaviour of the oxide, in normal conditions as well as in incidental conditions is an important care at the fuel element design level in a fast reactor. In nominal operating conditions, the probability of melt to core of the pellet is very low and even for high burnup. The behaviour in incidental operating conditions is also satisfying, especially for inadvertent rod ejections [fr

  18. Effect of aviation fuel type and fuel injection conditions on the spray characteristics of pressure swirl and hybrid air blast fuel injectors

    Science.gov (United States)

    Feddema, Rick

    Feddema, Rick T. M.S.M.E., Purdue University, December 2013. Effect of Aviation Fuel Type and Fuel Injection Conditions on the Spray Characteristics of Pressure Swirl and Hybrid Air Blast Fuel Injectors. Major Professor: Dr. Paul E. Sojka, School of Mechanical Engineering Spray performance of pressure swirl and hybrid air blast fuel injectors are central to combustion stability, combustor heat management, and pollutant formation in aviation gas turbine engines. Next generation aviation gas turbine engines will optimize spray atomization characteristics of the fuel injector in order to achieve engine efficiency and emissions requirements. Fuel injector spray atomization performance is affected by the type of fuel injector, fuel liquid properties, fuel injection pressure, fuel injection temperature, and ambient pressure. Performance of pressure swirl atomizer and hybrid air blast nozzle type fuel injectors are compared in this study. Aviation jet fuels, JP-8, Jet A, JP-5, and JP-10 and their effect on fuel injector performance is investigated. Fuel injector set conditions involving fuel injector pressure, fuel temperature and ambient pressure are varied in order to compare each fuel type. One objective of this thesis is to contribute spray patternation measurements to the body of existing drop size data in the literature. Fuel droplet size tends to increase with decreasing fuel injection pressure, decreasing fuel injection temperature and increasing ambient injection pressure. The differences between fuel types at particular set conditions occur due to differences in liquid properties between fuels. Liquid viscosity and surface tension are identified to be fuel-specific properties that affect the drop size of the fuel. An open aspect of current research that this paper addresses is how much the type of aviation jet fuel affects spray atomization characteristics. Conventional aviation fuel specifications are becoming more important with new interest in alternative

  19. Fuel mechanical design as a boundary condition for fuel management optimization

    International Nuclear Information System (INIS)

    Wunderlich, F.; Aisch, F.W.; Heins, L.

    1988-01-01

    The incentive to reduce fuel cycle costs as well as the amount of active waste requires, among others, measures to optimize fuel management. Improved fuel management in this sense calls, e.g., for reduction of parasitic neutron absorption, for reduction of neutron leakage, and particularly for burnup extension. Such measures result in increased demands for fuel mechanical design. In the first part of this paper their impact on fuel mechanical behaviour is described. In the second part, some examples of practical importance for the interaction between fuel management optimization and fuel mechanical design are discussed. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    1981-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-09-01

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

  2. Concrete Materials with Ultra-High Damage Resistance and Self- Sensing Capacity for Extended Nuclear Fuel Storage Systems

    Energy Technology Data Exchange (ETDEWEB)

    Li, Mo [Univ. of California, Irvine, CA (United States); Nakshatrala, Kalyana [Univ. of Houston, TX (United States); William, Kasper [Univ. of Houston, TX (United States); Xi, Yungping [Univ. of Colorado, Boulder, CO (United States)

    2017-02-08

    The objective of this project is to develop a new class of multifunctional concrete materials (MSCs) for extended spent nuclear fuel (SNF) storage systems, which combine ultra-high damage resistance through strain-hardening behavior with distributed multi-dimensional damage self-sensing capacity. The beauty of multifunctional concrete materials is two-fold: First, it serves as a major material component for the SNF pool, dry cask shielding and foundation pad with greatly improved resistance to cracking, reinforcement corrosion, and other common deterioration mechanisms under service conditions, and prevention from fracture failure under extreme events (e.g. impact, earthquake). This will be achieved by designing multiple levels of protection mechanisms into the material (i.e., ultrahigh ductility that provides thousands of times greater fracture energy than concrete and normal fiber reinforced concrete; intrinsic cracking control, electrochemical properties modification, reduced chemical and radionuclide transport properties, and crack-healing properties). Second, it offers capacity for distributed and direct sensing of cracking, strain, and corrosion wherever the material is located. This will be achieved by establishing the changes in electrical properties due to mechanical and electrochemical stimulus. The project will combine nano-, micro- and composite technologies, computational mechanics, durability characterization, and structural health monitoring methods, to realize new MSCs for very long-term (greater than 120 years) SNF storage systems.

  3. Acoustic Emission Analysis of Damage Progression in Thermal Barrier Coatings Under Thermal Cyclic Conditions

    Science.gov (United States)

    Appleby, Matthew; Zhu, Dongming; Morscher, Gregory

    2015-01-01

    Damage evolution of electron beam-physical vapor deposited (EBVD-PVD) ZrO2-7 wt.% Y2O3 thermal barrier coatings (TBCs) under thermal cyclic conditions was monitored using an acoustic emission (AE) technique. The coatings were heated using a laser heat flux technique that yields a high reproducibility in thermal loading. Along with AE, real-time thermal conductivity measurements were also taken using infrared thermography. Tests were performed on samples with induced stress concentrations, as well as calcium-magnesium-alumino-silicate (CMAS) exposure, for comparison of damage mechanisms and AE response to the baseline (as-produced) coating. Analysis of acoustic waveforms was used to investigate damage development by comparing when events occurred, AE event frequency, energy content and location. The test results have shown that AE accumulation correlates well with thermal conductivity changes and that AE waveform analysis could be a valuable tool for monitoring coating degradation and provide insight on specific damage mechanisms.

  4. Reduction of damage initiation density in fused silica optics via UV laser conditioning

    Science.gov (United States)

    Peterson, John E.; Maricle, Stephen M.; Brusasco, Raymond M.; Penetrante, Bernardino M.

    2004-03-16

    The present invention provides a method for reducing the density of sites on the surface of fused silica optics that are prone to the initiation of laser-induced damage, resulting in optics which have far fewer catastrophic defects and are better capable of resisting optical deterioration upon exposure for a long period of time to a high-power laser beam having a wavelength of about 360 nm or less. The initiation of laser-induced damage is reduced by conditioning the optic at low fluences below levels that normally lead to catastrophic growth of damage. When the optic is then irradiated at its high fluence design limit, the concentration of catastrophic damage sites that form on the surface of the optic is greatly reduced.

  5. The Effect of Fuel Injector Nozzle Configuration on JP-8 Sprays at Diesel Engine Conditions

    Science.gov (United States)

    2014-10-01

    The Effect of Fuel Injector Nozzle Configuration on JP-8 Sprays at Diesel Engine Conditions by Matthew Kurman, Luis Bravo, Chol-Bum Kweon...Fuel Injector Nozzle Configuration on JP-8 Sprays at Diesel Engine Conditions Matthew Kurman, Luis Bravo, and Chol-Bum Kweon Vehicle Technology...March 2014 4. TITLE AND SUBTITLE The Effect of Fuel Injector Nozzle Configuration on JP-8 Sprays at Diesel Engine Conditions 5a. CONTRACT NUMBER 5b

  6. Damage to E. coli cells induced by tritium decay: secondary lethality under nongrowth conditions

    International Nuclear Information System (INIS)

    Koukalova, B.; Kuhrova, V.

    1980-01-01

    Cells containing incorporated 3 H-thymidine are damaged by its decay. It was found with E.coli TAU-bar cells that a small part of the damage is lethal whereas most of it is reparable and only potentially lethal. If cells are subjected to nongrowth conditions, the potentially lethal damage changes to lethal damage. This process is called secondary lethality (SL). The extent of SL and some changes in DNA under three different modes of growth inhibition were determined. It was found that: (i) SL is maximal under conditions of amino acid starvation (-AA), the viable count decreasing by two orders of magnitude. (ii) SL is 4 times lower in the presence of chloramphenicol (-AA+CLP) and 6.5 times lower under +AA+CLP conditions. Changes in the sedimentation rate of DNA determined in alkaline sucrose gradient correlate with the differences in SL: under -AA conditions the sedimentation rate of DNA decreases whereas in the presence of CLP no decrease occurs. The results suggest that certain enzymatic processes take place under -AA conditions which lead to irreparable changes in DNA. (author)

  7. Damage to E. coli cells induced by tritium decay: secondary lethality under nongrowth conditions

    Energy Technology Data Exchange (ETDEWEB)

    Koukalova, B; Kuhrova, V [Ceskoslovenska Akademie Ved, Brno. Biofysikalni Ustav

    1980-05-01

    Cells containing incorporated /sup 3/H-thymidine are damaged by its decay. It was found with E.coli TAU-bar cells that a small part of the damage is lethal whereas most of it is reparable and only potentially lethal. If cells are subjected to nongrowth conditions, the potentially lethal damage changes to lethal damage. This process is called secondary lethality (SL). The extent of SL and some changes in DNA under three different modes of growth inhibition were determined. It was found that: (i) SL is maximal under conditions of amino acid starvation (-AA), the viable count decreasing by two orders of magnitude. (ii) SL is 4 times lower in the presence of chloramphenicol (-AA+CLP) and 6.5 times lower under +AA+CLP conditions. Changes in the sedimentation rate of DNA determined in alkaline sucrose gradient correlate with the differences in SL: under -AA conditions the sedimentation rate of DNA decreases whereas in the presence of CLP no decrease occurs. The results suggest that certain enzymatic processes take place under -AA conditions which lead to irreparable changes in DNA.

  8. Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

  9. Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments

    International Nuclear Information System (INIS)

    Sato, Ikken; Lemoine, Francette; Struwe, Dankward

    2004-01-01

    In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear density and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long

  10. Behavior of pre-irradiated fuel under a simulated RIA condition

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide

    1994-07-01

    This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-3, conducted in the Nuclear Safety Research Reactor (NSSR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and analyses, interpretations, and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 19.6MWd/kgU with average linear heat rate of 25.3 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a deposited energy of 174±6 cal/g·fuel and a peak fuel enthalpy of 130±5 cal/g·fuel under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The cladding surface temperature increased to only 150degC, and the test resulted in slight fuel deformation and no fuel failure. An estimated rod-average fission gas release during the transient was about 2.2%. Through the detailed fuel examinations, the information concerning microstructural change in the fuel pellets were also obtained. (author)

  11. The Intrinsic Inferiority of Efficiency Wages to Damages and Conditional Bonuses

    NARCIS (Netherlands)

    de Geest, G.G.A.; Dari Mattiacci, G.; Siegers, J.J.

    In this paper, we argue that, as an enforcement mechanism, efficiency wages are intrinsically inferior to damages and to conditional bonuses – an alternative positive sanction system overlooked in the labor economics literature, under which rents are only paid if monitoring has effectively taken

  12. The intrinsic inferiority of efficiency wages to damages and conditional bonuses

    NARCIS (Netherlands)

    Geest, Gerrit de; Dari Mattiacci, G.; Siegers, J.J.

    2004-01-01

    In this paper, we argue that, as an enforcement mechanism, efficiency wages are intrinsically inferior to damages and to conditional bonuses an alternative positive sanction system overlooked in the labor economics literature, under which rents are only paid if monitoring has effectively taken

  13. 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

  14. Utility residual fuel oil market conditions: An update

    International Nuclear Information System (INIS)

    Mueller, H.A. Jr.

    1992-01-01

    Planning for residual fuel oil usage and management remains an important part of the generation fuel planning and management function for many utilities. EPRI's Utility Planning Methods Center has maintained its analytical overview of the fuel oil markets as part of its overall fuel planning and management research program. This overview provides an update of recent fuel oil market directions. Several key events of the past year have had important implications for residual fuel oil markets. The key events have been the changes brought about by the Persian Gulf War and its aftermath, as well as continuing environmental policy developments. The Persian Gulf conflict has created renewed interest in reducing fuel oil use by utilities as part of an overall reduction in oil imports. The policy analysis performed to date has generally failed to properly evaluate utility industry capability. The Persian Gulf conflict has also resulted in an important change in the structure of international oil markets. The result of this policy-based change is likely to be a shift in oil pricing strategy. Finally, continued change in environmental requirements is continuing to shift utility residual oil requirements, but is also changing the nature of the US resid market itself

  15. 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

  16. Modeling of fuel retention in the pre-damaged tungsten with MeV W ions after exposure to D plasma

    Directory of Open Access Journals (Sweden)

    Zhenhou Wang

    2017-12-01

    Full Text Available Modeling of high-Z ion irradiated-induced damages on fuel retention inside tungsten (W material has been performed in this work. The upgraded Hydrogen Isotope Inventory Processes Code (HIIPC is applied to model the deuterium (D retention inside pre-damaged W during exposed to low-energy D flux, and the W is pre-irradiated by 20 MeV W-ion before exposed to D flux. Three types of trap, i.e. mono-vacancies, dislocations and grain boundary vacancies, are considered in the present model. The mono-vacancy defects induced by energetic W ions are calculated by SRIM code. First, the model is validated against the available experimental data under the same D flux exposure conditions, showing the reasonable agreement. Then, the effect of radiation-induced defects produced by pre-exposed energetic W-ion with different energy and fluence on the fuel retention are studied, confirming that the irradiation-induced traps play a dominated role on the fuel retention in the surface of the material (∼ micrometer. Finally, the effects of different type of defect, D fluence, and wall temperature on the fuel retention are discussed systemically, and these modeling results are in well agreement with the previous studies.

  17. Research program on conditions to failure of high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Regarding the power ramp test to verify the out-of-pile test results on hydrogen-induced cladding failure, situation of the shipping port restoration after the earthquake disaster was investigated for the overseas transportation of test fuel rods which had been interrupted. Its reopening schedule was still currently uncertain and the power ramp test plan also remained suspended. The information about the fuel irradiation performance obtained from JNES projects and international projects, etc. is prepared as database, and based on the recent findings, the fuel irradiation performance models and analysis codes are developed and/or improved. (author)

  18. Main conditions and effectiveness of gas fuel use for powering of dual fuel IC self-ignition engine

    Directory of Open Access Journals (Sweden)

    Stefan POSTRZEDNIK

    2015-09-01

    Full Text Available Internal combustion engines are fuelled mostly with liquid fuels (gasoline, diesel. Nowadays the gaseous fuels are applied as driving fuel of combustion engines. In case of spark ignition engines the liquid fuel (petrol can be totally replaced by the gas fuels. This possibility in case of compression engines is essentially restricted through the higher self-ignition temperatures of the combustible gases in comparison to classical diesel oil. Solution if this problem can be achieved by using of the dual fuel system, where for ignition of the prepared fuel gas - air mixture a specified amount of the liquid fuel (diesel oil should be additionally injected into the combustion chamber. For assurance that the combustion process proceeds without mistakes and completely, some basic conditions should be satisfied. In the frame of this work, three main aspects of this problem are taken into account: a. filling efficiency of the engine, b. stoichiometry of the combustion, c. performance of mechanical parameters (torque, power. A complex analysis of these conditions has been done and some achieved important results are presented in the paper.

  19. Change in geometrical parameters of WWER high burnup fuel rods under operational conditions and transient testing

    International Nuclear Information System (INIS)

    Kanashov, B.; Amosov, S.; Lyadov, G.; Markov, D.; Ovchinnikov, V; Polenok, V.; Smirnov, A.; Sukhikh, A.; Bek, E.; Yenin, A.; Novikov, V.

    2001-01-01

    The paper discusses changes in fuel rods geometric parameters as result of operation conditions and burnups. The degree of geometry variability of fuel rods, cladding and column is one of the most important characteristics affecting fuel serviceability. On the other hand, changes in fuel rod geometric parameters influence fuel temperature, fission gas release, fuel-to-cladding stress strained state as well as the degree of interaction with FA skeleton elements and skeleton rigidity. Change in fuel-to-cladding gap is measured using compression technique. The axial distribution of fuel-to-cladding gap demonstrates the largest decrease of the gap in the region 500 to 2000 mm from the bottom of the fuel rod (WWER-440) and in the region of 500 to 3000 mm for WWER-1000. The cladding material creep in WWER fuel rods together with the radiation growth results in fuel rod cladding elongation. A set of transient tests for spent WWER-440 and WWER-1000 fuel rods carried out in SSC RIAR during a period 1995-1999, with the aim to estimate the changes in geometric parameters of FRs. The estimation of changes in outer diameter of cladding and fuel column and fuel-to-cladding gap are performed in transient conditions (changes in linear power range of 180 to 400 W/cm) for both WWER-440 and WWER-1000. WWER-440 fuel rods having the same burnup and close fuel-cladding contact before testing are subjected to considerable hoop cladding strain in testing up to 300 W/cm. But the hoop strain does not grow due to the structural changes in fuel column and decrease in central hole diameter occurred when the power is higher

  20. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Phillpot, Simon; Tulenko, James

    2011-09-08

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  1. Fundamental Processes of Coupled Radiation Damage and Mechanical Behavior in Nuclear Fuel Materials for High Temperature Reactors

    International Nuclear Information System (INIS)

    Phillpot, Simon; Tulenko, James

    2011-01-01

    The objective of this work has been to elucidate the relationship among microstructure, radiation damage and mechanical properties for nuclear fuel materials. As representative nuclear materials, we have taken an hcp metal (Mg as a generic metal, and Ti alloys for fast reactors) and UO2 (representing fuel). The degradation of the thermo-mechanical behavior of nuclear fuels under irradiation, both the fissionable material itself and its cladding, is a longstanding issue of critical importance to the nuclear industry. There are experimental indications that nanocrystalline metals and ceramics may be more resistant to radiation damage than their coarse-grained counterparts. The objective of this project look at the effect of microstructure on radiation damage and mechanical behavior in these materials. The approach to be taken was state-of-the-art, large-scale atomic-level simulation. This systematic simulation program of the effects of irradiation on the structure and mechanical properties of polycrystalline Ti and UO2 identified radiation damage mechanisms. Moreover, it will provided important insights into behavior that can be expected in nanocrystalline microstructures and, by extension, nanocomposites. The fundamental insights from this work can be expected to help in the design microstructures that are less susceptible to radiation damage and thermomechanical degradation.

  2. The analysis of the annular fuel performance in steady state condition by using AFPAC code

    International Nuclear Information System (INIS)

    He Xiaojun; Ji Songtao; Zhang Yingchao

    2012-01-01

    The fuel performance code AFPAC v1.0 is used to analyze annular fuel's behavior under steady state conditions, including neutronics, thermal hydraulic, rod deformation, fission gas release and rod internal pressure. The calculation results show that: 1) Annular fuel has a good steady irradiation performance at 150% power level as current LWRs' with burnup up to 50 GWd/t, and all parameters, such as temperature, rod internal pressure and rod deformation, are meet the rod design criteria for current fuel of PWRs: 2) Compared to the solid fuel under the same irradiation condition. annular fuel has lower temperature, smaller deformation, lower fission gas release and lower pressure at EOL. From the point of view of steady irradiation performance, the safety of reactors can significantly improved by u sing the annular fuel. (authors)

  3. Development of a laser multi-layer cladding technology for damage mitigation of fuel spacers in Hanaro reactor

    International Nuclear Information System (INIS)

    Kim, J. S.; Lee, D. H.; Hwang, S. S.; Suh, J. H.

    2002-01-01

    A laser multi-layer cladding technology was developed to mitigate the fretting wear damages occurred at fuel spacers in Hanaro reactor. The detailed experimental results are as follows. 1) Analyses of fretting wear damages and fabrication process of fuel spacers 2) Development and analysis of spherical Al 6061 T-6 alloy powders for the laser cladding 3) Analysis of parameter effects on laser cladding process for clad bids, and optimization of laser cladding process 4) Analysis on the changes of cladding layers due to overlapping factor change 5) Microstructural observation and phase analysis 6) Characterization of materials properties (hardness and wear tests) 7) Manufacture of prototype fuel spacers 8) Development of a vision system and revision of its related softwares

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  5. 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)

  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. Study of fuel control strategy based on an fuel behavior model for starting conditions; Nenryo kyodo model ni motozuita shidoji no nenryo hosei hosho ni tsuite no kosatsu

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Y; Uchida, M; Iwano, H; Oba, H [Nissan Motor Co. Ltd., Tokyo (Japan)

    1997-10-01

    We have applied a fuel behavior model to a fuel injection system which we call SOFIS (Sophisticated and Optimized Fuel Injection System) so that we get air/fuel ratio control accuracy and good driveability. However the fuel behavior under starting conditions is still not clear. To meet low emission rules and to get better driveability under starting conditions, better air/fuel ratio control is necessary. Now we have understood the ignition timing, injection timing, and injection pulse width required in such conditions. In former days, we analyzed the state of the air/fuel mixture under cold conditions and made a new fuel behavior model which considered fuel loss such as hydrocarbons and dissolution into oil and so on. Al this time, we have applied this idea to starting. We confirm this new model offers improved air/fuel ratio control. 6 refs., 9 figs., 3 tabs.

  8. Repairable-conditionally repairable damage model based on dual Poisson processes.

    Science.gov (United States)

    Lind, B K; Persson, L M; Edgren, M R; Hedlöf, I; Brahme, A

    2003-09-01

    The advent of intensity-modulated radiation therapy makes it increasingly important to model the response accurately when large volumes of normal tissues are irradiated by controlled graded dose distributions aimed at maximizing tumor cure and minimizing normal tissue toxicity. The cell survival model proposed here is very useful and flexible for accurate description of the response of healthy tissues as well as tumors in classical and truly radiobiologically optimized radiation therapy. The repairable-conditionally repairable (RCR) model distinguishes between two different types of damage, namely the potentially repairable, which may also be lethal, i.e. if unrepaired or misrepaired, and the conditionally repairable, which may be repaired or may lead to apoptosis if it has not been repaired correctly. When potentially repairable damage is being repaired, for example by nonhomologous end joining, conditionally repairable damage may require in addition a high-fidelity correction by homologous repair. The induction of both types of damage is assumed to be described by Poisson statistics. The resultant cell survival expression has the unique ability to fit most experimental data well at low doses (the initial hypersensitive range), intermediate doses (on the shoulder of the survival curve), and high doses (on the quasi-exponential region of the survival curve). The complete Poisson expression can be approximated well by a simple bi-exponential cell survival expression, S(D) = e(-aD) + bDe(-cD), where the first term describes the survival of undamaged cells and the last term represents survival after complete repair of sublethal damage. The bi-exponential expression makes it easy to derive D(0), D(q), n and alpha, beta values to facilitate comparison with classical cell survival models.

  9. Critical stability conditions of the fuel element cladding; Kriticni uslovi stabilnosti kosuljice G.E

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, M; Savic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-12-15

    The role of the fuel element cladding being the first safety barrier, is to prevent contamination by the fission products. Construction of the fuel element cladding depends on the reactor type, coolant type, fuel type, technology of material fabrication, influence of the material on the neutron economy, thermal conditions, etc. That is why an optimum solution has to be found. This paper deals with mechanical properties of ceramic natural UO{sub 2} sintered fuel pellets in the zircaloy-2 cladding. This type of fuel is used in heavy water reactors.

  10. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Takats, F [TS Enercon Kft. (Hungary)

    2012-07-01

    Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. The fresh fuel is imported from Russia so far. The spent fuel assemblies were shipped back to Russia until 1997 after about 6 years cooling at the plant. A dry storage facility (MVDS type) has been constructed and is operational since then. By 1 January 2008, there were 5107 assemblies in dry storage. The objectives are: 1) Wet AR storage of spent fuel from the NPP Paks: Measurements of conditions for spent fuel storage in the at-reactor (AR) storage pools of Paks NPP (physical and chemical characteristics of pool water, corrosion product data); Measurements and visual control of storage pool component characteristics; Evaluation of storage characteristics and conditions with respect to long-term stability (corrosion of fuel cladding, construction materials); 2) Dry AFR storage at Paks NPP: Calculation and measurement of spent fuel conditions during the transfer from the storage pool to the modular vault dry storage (MVDS) on the site; Calculation and measurement of spent fuel conditions during the preparation of fuel for dry storage (drying process), such as crud release, activity build-up; Measurement of spent fuel conditions during the long-term dry storage, activity data in the storage tubes and amount of crud.

  11. Fuel-coolant interaction-phenomena under prompt burst conditions

    International Nuclear Information System (INIS)

    Jacobs, H.; Young, M.F.; Reil, K.O.

    1979-01-01

    The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2

  12. Production bias: A proposed modification of the driving force for void swelling under cascade damage conditions

    International Nuclear Information System (INIS)

    Woo, C.H.; Garner, F.A.

    1991-11-01

    A new concept of point-defect production as the main driving force for void swelling under cascade damage conditions is proposed. This concept takes into account the recombination and formation of immobile clusters and loops of vacancies and interstitials in the cascade region. The life times of the clusters and loops due to desolution are strong functions of the temperature, as well as their vacancy and interstitial nature. The resulting biased production of free point defects from the internal sources is shown to be a strong driving force for void swelling. The characteristics of void swelling due to production bias are described and compared with experimental results. We conclude that the production bias concept provides a good description of void swelling under cascade damage conditions

  13. Production bias: A proposed modification of the driving force for void swelling under cascade damage conditions

    International Nuclear Information System (INIS)

    Woo, C.H.; Singh, B.N.; Garner, F.A.

    1992-01-01

    A new concept of point defect production as the main driving force for void swelling under cascade damage conditions is proposed. This concept takes into account the recombination and formation of immobile clusters and loops of vacancies and interstitials in the cascade region. The lifetimes of the clusters and loops due to desolution are strong functions of the temperature, as well as their vacancy and interstitial nature. The resulting biased production of free point defects from the internal sources is shown to be a strong driving force for void swelling. The characteristics of void swelling due to production bias are described and compared with experimental results. We conclude that the production bias concept provides a good description of void swelling under cascade damage conditions. (orig.)

  14. A seismic analysis of Korean standard PWR fuels under transition core conditions

    International Nuclear Information System (INIS)

    Kim, Hyeong Koo; Park, Nam Kyu; Jang, Young Ki; Kim, Jae Ik; Kim, Kyu Tae

    2005-01-01

    The PLUS7 fuel is developed to achieve higher thermal performance, burnup and more safety margin than the conventional fuel used in the Korean Standard Nuclear Plants (KSNPs) and to sustain structural integrity under increased seismic requirement in Korea. In this study, a series of seismic analysis have been performed in order to evaluate the structural integrity of fuel assemblies associated with seismic loads in the KSNPs under transition core conditions replacing the Guardian fuel, which is a resident fuel in the KSNP reactors, with the PLUS7 fuel. For the analysis, transition core seismic models have been developed, based on the possible fuel loading patterns. And the maximum impact forces on the spacer grid and various stresses acting on the fuel components have been evaluated and compared with the through-grid strength of spacer grids and the stress criteria specified in the ASME code for each fuel component, respectively. Then three noticeable parameters regarding as important parameters governing fuel assembly dynamic behavior are evaluated to clarify their effects on the fuel impact and stress response. As a result of the study, it has been confirmed that both the PLUS7 and the Guardian fuel sustain their structural integrity under the transition core condition. And when the damping ratio is constant, increasing the natural frequency of fuel assembly results in a decrease in impact force. The fuel assembly flexural stiffness has an effect increasing the stress of fuel assembly, but not the impact force. And the spacer grid stiffness is directly related with the impact force response. (author)

  15. Frost damage of roof tiles: A study on moisture boundary conditions

    OpenAIRE

    Iba, Chiemi; Ueda, Ayumi; Hokoi, Shuichi

    2015-01-01

    Freeze-thaw cycles are the most serious cause of roof tile deterioration; thus, it is important to know the temperature and moisture distributions in tile materials for protection against frost damage. This study focused on moisture boundary conditions for air layers under the tile. Temperature and humidity were measured using model structures with different types of roof tiles. The results showed that the temperatures around the roof were strongly influenced by solar and longwave radiation, ...

  16. Calculation of fuel pin failure timing under LOCA conditions

    International Nuclear Information System (INIS)

    Jones, K.R.; Wade, N.L.; Siefken, L.J.; Straka, M.; Katsma, K.R.

    1991-10-01

    The objective of this research was to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

  17. Behavior and failure of fresh, hydrided and irradiated Zircaloy-4 fuel claddings under RIA conditions

    International Nuclear Information System (INIS)

    Le Saux, M.

    2008-01-01

    The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fresh, hydrided and irradiated (in pressurized water reactors) cold-worked stress relieved Zircaloy-4 fuel claddings under reactivity initiated accident conditions. A model is proposed to describe the anisotropic viscoplastic mechanical behavior of the material as a function of temperature (from 20 C up to 1100 C), strain rate (from 3.10 -4 s -1 up to 5 s -1 ), fluence (from 0 up to 1026 n.m -2 ) and irradiation conditions. Axial tensile, hoop tensile, expansion due to compression and hoop plane strain tensile tests are performed at 25 C, 350 C and 480 C in order to analyse the anisotropic plastic and failure properties of the non-irradiated material hydrided up to 1200 ppm. Material strength and strain hardening depend on temperature and hydrogen in solid solution and precipitated hydride contents. Plastic anisotropy is not significantly modified by hydrogen. The material is embrittled by hydrides at room temperature. The plastic strain that leads to hydride cracking decreases with increasing hydrogen content. The material ductility, which increases with increasing temperature, is not deteriorated by hydrogen at 350 C and 480 C. Macroscopic fracture modes and damage mechanisms depend on specimen geometry, temperature and hydrogen content. A Gurson type model is finally proposed to describe both the anisotropic viscoplastic behavior and the ductile fracture of the material as a function of temperature and hydrogen content. (author) [fr

  18. On the corrosion behavior of zircaloy-4 in spent fuel pools under accidental conditions

    International Nuclear Information System (INIS)

    Lavigne, O.; Shoji, T.; Sakaguchi, K.

    2012-01-01

    Highlights: ► Corrosion behavior of oxidized Zr-4 in alkaline media in presence of chloride and radical forms. ► Generation of radical forms by sonolysis of water. ► Limited increase of the passive current densities under polarization with the increase of pH and the presence of radicals. ► Decrease of the passive range of oxidized Zr-4 with presence of Cl − (E pit ∼ 0.6 V SCE ). ► Decrease of the pitting potential when oxide layer is scratched or damaged (E pit ∼ 0.16 V SCE ). - Abstract: After zircaloy cladding tubes have been subjected to irradiation in the reactor core, they are stored temporarily in spent fuel pools. In case of an accident, the integrity of the pool may be affected and the composition of the coolant may change drastically. This was the case in Fukushima Daiichi in March 2011. Successive incidents have led to an increase in the pH of the coolant and to chloride contamination. Moreover, water radiolysis may occur owing to the remnant radioactivity of the spent fuel. In this study, we propose to evaluate the corrosion behavior of oxidized Zr-4 (in autoclave at 288 °C for 32 days) in function of the pH and the presence of chloride and radical forms. The generation of radicals is achieved by the sonolysis of the solution. It appears that the increase in pH and the presence of radicals lead to an increase in current densities. However, the current densities remain quite low (depending on the conditions, between 1 and 10 μA cm −2 ). The critical parameter is the presence of chloride ions. The chloride ions widely decrease the passive range of the oxidized samples (the pitting potential is measured around +0.6 V (vs. SCE)). Moreover, if the oxide layer is scratched or damaged (which is likely under accidental conditions), the pitting potential of the oxidized sample reaches the pitting potential of the non-oxidized sample (around +0.16 V (vs. SCE)), leaving a shorter stable passive range for the Zr-4 cladding tubes.

  19. Enhanced taurine release in cell-damaging conditions in the developing and ageing mouse hippocampus.

    Science.gov (United States)

    Saransaari, P; Oja, S S

    1997-08-01

    Taurine has been shown to be essential for neuronal development and survival in the central nervous system. The release of preloaded [3H]taurine was studied in hippocampal slices from seven-day-, three-month- and 18-22-month-old mice in cell-damaging conditions. The slices were superfused in hypoxic, hypoglycemic and ischemic conditions and exposed to free radicals and oxidative stress. The release of taurine was greatly enhanced in the above conditions in all age groups, except in oxidative stress. The release was large in ischemia, particularly in the hippocampus of aged mice. Potassium stimulation was still able to release taurine in cell-damaging conditions in immature mice, whereas in adult and aged animals the release was so substantial that this additional stimulus failed to work. Taurine release was partially Ca2+-dependent in all cases. The massive release of the inhibitory amino acid taurine in ischemic conditions could act neuroprotectively, counteracting in several ways the effects of simultaneous release of excitatory amino acids. This protection could be of great importance in developing brain tissue, while also having an effect in aged brains.

  20. New oil condition monitoring system, Wearsens® enables continuous, online detection of critical operating conditions and wear damage

    Directory of Open Access Journals (Sweden)

    Manfred Mauntz

    2015-12-01

    Full Text Available A new oil sensor system is presented for the continuous, online measurement of the wear in turbines, industrial gears, generators, hydraulic systems and transformers. Detection of change is much earlier than existing technologies such as particle counting, vibration measurement or recording temperature. Thus targeted, corrective procedures and/or maintenance can be carried out before actual damage occurs. Efficient machine utilization, accurately timed preventive maintenance, increased service life and a reduction of downtime can all be achieved. The presented sensor system effectively controls the proper operation conditions of bearings and cogwheels in gears. The online diagnostics system measures components of the specific complex impedance of oils. For instance, metal abrasion due to wear debris, broken oil molecules, forming acids or oil soaps, result in an increase of the electrical conductivity, which directly correlates with the degree of contamination of the oil. For additivated lubricants, the stage of degradation of the additives can also be derived from changes in the dielectric constant. The determination of impurities or reduction in the quality of the oil and the quasi continuous evaluation of wear and chemical aging follow the holistic approach of a real-time monitoring of an alteration in the condition of the oil-machine system. Once the oil condition monitoring sensors are installed on the wind turbine, industrial gearbox and test stands, the measuring data can be displayed and evaluated elsewhere. The signals are transmitted to a web-based condition monitoring system via LAN, WLAN or serial interfaces of the sensor unit. Monitoring of the damage mechanisms during proper operation below the tolerance limits of the components enables specific preventive maintenance independent of rigid inspection intervals.

  1. Alternative fuels from forest residues for passenger cars - an assessment under German framework conditions

    OpenAIRE

    Hurtig, O.; Leible, L.; Kälber, S.; Kappler, g.; Spicher, U.

    2014-01-01

    Background Due to the available volumes, biogenic residues are a promising resource for renewable fuels for passenger cars to reduce greenhouse gas (GHG) emissions. In this study, we compare three fuels from forest residues under German framework conditions: biogenic electricity, substitute natural gas (SNG), and Fischer-Tropsch (FT) diesel. Methods Fuels from forest residues are compared with regard to their technical efficiency (here defined as ‘pkm per kg b...

  2. Low temperature spent fuel oxidation under tuff repository conditions

    International Nuclear Information System (INIS)

    Einziger, R.E.; Woodley, R.E.

    1985-01-01

    The Nevada Nuclear Waste Storage Investigations Project is studying the suitability of tuffaceous rocks at Yucca Mountain, Nye County, Nevada, for high level waste disposal. The oxidation state of LWR spent fuel in a tuff repository may be a significant factor in determining its ability to inhibit radionuclide migration. Long term exposure at low temperatures to the moist air expected in a tuff repository is expected to increase the oxidation state of the fuel. A program is underway to determine the spent fuel oxidation mechanisms which might be active in a tuff repository. Initial work involves a series of TGA experiments to determine the effectiveness of the technique and to obtain preliminary oxidation data. Tests were run at 200 0 C and 225 0 C for as long as 720 hours. Grain boundary diffusion appears to open up a greater surface area for oxidation prior to onset of bulk diffusion. Temperature strongly influences the oxidation rates. The effect of moisture is small but readily measurable. 25 refs., 7 figs., 4 tabs

  3. 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.

  4. 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

  5. The Effect of Nozzle Design and Operating Conditions on the Atomization and Distribution of Fuel Sprays

    Science.gov (United States)

    Lee, Dana W

    1933-01-01

    The atomization and distribution characteristics of fuel sprays from automatic injection valves for compression-ignition engines were determined by catching the fuel drops on smoked-glass plates, and then measuring and counting the impressions made in the lampblack. The experiments were made in an air-tight chamber in which the air density was raised to values corresponding to engine conditions.

  6. 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

  7. 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)

  8. The influence of operational conditions on radiation damage in JFET-input operational amplifiers

    International Nuclear Information System (INIS)

    Zheng Yuzhan; Wang Yiyuan; Chen Rui; Fei Wuxiong; Lu Wu; Ren Diyuan

    2010-01-01

    High- and low-dose-rate irradiation have been performed on JFET-input operational amplifiers (op-amps) with normal operational and zero biased conditions, respectively. The experimental results show that operational conditions have a great influence on the radiation effects and damage in JFET-input operational amplifiers. Under normal condition, the JFET-input op-amps have exhibited time-dependent effect (TDE); while they show enhanced low-dose-rate sensitivity (ELDRS) at zero biased condition. Compared with zero biased condition, the JFET-input op-amps would degrade more severely at normal condition for high-dose-rate irradiation; while for the low-dose-rate case, they have more degradation at normal condition. Irradiation would induce positive oxide-trapped charge and interface traps in bipolar transistors, which are the basic components in JFET-input op-amps. From the dependence of oxide trapped charge and interface traps on operational conditions, the degradation behavior is discussed. (authors)

  9. Conditioning of spent fuel for interim and final storage in the pilot conditioning plant (PKA) at Gorleben

    International Nuclear Information System (INIS)

    Lahr, H.; Willax, H.O.; Spilker, H.

    1999-01-01

    In 1994, due to the change of the nuclear law in Germany, the concept of direct final disposal for spent fuel was developed as an equivalent alternative to the waste management with reprocessing. Since 1979, tests for the direct final disposal of spent fuel have been conducted in Germany. In 1985, the State and the utilities came to an agreement to develop this concept of waste management to technical maturity. Gesellschaft fuer Nuklear-Service (GNS) was commissioned by the utilities with the following tasks: to develop and test components with regard to conditioning technology, to construct and operate the pilot conditioning plant (PKA), and to develop casks suitable for final disposal. Since 1990, the construction of the PKA has taken place at the Brennelementlager Gorleben site. The PKA has been designed as a multipurpose facility and can thus fulfil various tasks within the framework of the conditioning and management of spent fuel assemblies and radioactive waste. The pilot character of the plant allows for development and testing in the field of spent fuel assembly conditioning. The objectives of the PKA may be summarized as follows: to condition spent fuel assemblies, to reload spent fuel assemblies and waste packages, to condition radioactive waste, and to do maintenance work on transport and storage casks as well as on waste packages. Currently, the buildings of the PKA are constructed and the technical facilities are installed. The plant will be ready for service in the middle of 1999. It is the first plant of its kind in the world. (author)

  10. Conditioning spent fuels from research nuclear reactor in ceramic dies

    International Nuclear Information System (INIS)

    Russo, D.O; Rodriguez, D.S; Mateos, P; Heredia, A; Sangilippo, M; Sterba, M

    2002-01-01

    The problem of immobilizing nuclear wastes is a complex one and is vitally important in the nuclear fuels cycle. In the case of spent elements from research reactors, the presence of large amounts of aluminum makes the procedure more complex and, therefore, onerous. There are various alternatives proposed for processing these materials. Two methods were studied in the Nuclear Materials Division for obtaining, as a final product, a vitreous block that could be place definitively in a geological repository. The processes are briefly, as follows: 1.By mechanical and chemical processes eliminating all the exterior aluminum from the fuel plates and then placing the product which we will call 'meat' (with some additional treatment and mixing with the amount needed to produce a natural uranium compound or weakened by decreasing the isotope enrichment in U-235) in a vitreous matrix. 2.Mechanically eliminate the aluminum from the exterior frame (as shown below) by shearing and cutting off the sectors containing only the Al, but leaving the rest of the aluminum, a big part of which is still present (4511.03), then doing the same procedure as in the case above: mixing with a natural uranium compound or weakening and vitrifying this mixture. In both cases, the vitrification can be carried out by fusion as well as by sintering. Given that these methods imply a big increase in volume together with a big mass of uranium and an even bigger amount of glass we decided to study an alternative. The proposed process involves synthesizing the mixtures obtained from the pre-treatment of the fuel plates (as described later) with natural isotope uranium oxide in order to obtain a block with the appropriate properties for its final disposal in a deep geological repository (CW)

  11. Selection of optimal conditions for preparation of emulsified fuel fluids

    Science.gov (United States)

    Ivanov, V. A.; Berg, V. I.; Frolov, M. D.

    2018-05-01

    The aim of the article is to derive the optimal concept of physical and chemical effects, and its application to the production of water-fuel emulsions. The authors set a research task to attempt to estimate the influence of the surfactant concentration on such indicator as the time before the beginning of emulsion breaking. The analysis, based on experimental data, showed that an increase in the concentration of sodium lauryl sulfate is expedient to a certain point, corresponding to 0.05% of the total mass fraction. The main advantage of the model is a rational combination of methods of physical and chemical treatment used in the production of emulsions.

  12. Trends for Methane Oxidation at Solid Oxide Fuel Cell Conditions

    DEFF Research Database (Denmark)

    Kleis, Jesper; Jones, Glenn; Abild-Pedersen, Frank

    2009-01-01

    First-principles calculations are used to predict a plausible reaction pathway for the methane oxidation reaction. In turn, this pathway is used to obtain trends in methane oxidation activity at solid oxide fuel cell (SOFC) anode materials. Reaction energetics and barriers for the elementary...... the Ni surfaces to other metals of interest. This allows the reactivity over the different metals to be understood in terms of two reactivity descriptors, namely, the carbon and oxygen adsorption energies. By combining a simple free-energy analysis with microkinetic modeling, activity landscapes of anode...

  13. Experimental loop for fast neutron fuels under normal, abnormal, transient and emergency conditions

    International Nuclear Information System (INIS)

    Bauge, M.; Colomez, G.; Marfaing, R.J.; Mourain, M.

    1976-01-01

    Within the scope of safety experiments on power reactor fuels, an experimental loop is described which can, by reduction of the flow, flush the sodium joint of vented mixed carbide fuel elements and allow the study of the resulting phenomena. With the help of the annex laboratories at OSIRIS, the control test can be analyzed and followed, with special attention to the study of the migration of fission products inside and outside the fuel. This apparatus can, of course, also be used for testing the fuels under normal and abnormal working conditions [fr

  14. 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.)

  15. BWR fuel performance under advanced water chemistry conditions – a delicate journey towards zero fuel failures – a review

    International Nuclear Information System (INIS)

    Hettiarachchi, S.

    2015-01-01

    Boiling Water Reactors (BWRs) have undergone a variety of chemistry evolutions over the past few decades as a result of the need to control stress corrosion cracking of reactor internals, radiation fields and personnel exposure. Some of the advanced chemistry changes include hydrogen addition, zinc addition, iron reduction using better filtration technologies, and more recently noble metal chemical addition to many of the modern day operating BWRs. These water chemistry evolutions have resulted in changes in the crud distribution on fuel cladding material, Co-60 levels and the Rod oxide thickness (ROXI) measurements using the conventional eddy current techniques. A limited number of Post-Irradiation Examinations (PIE) of fuel rods that exhibited elevated oxide thickness using eddy current techniques showed that the actual oxide thickness by metallography is much lower. The difference in these observations is attributed to the changing magnetic properties of the crud affecting the rod oxide thickness measurement by the eddy current technique. This paper will review and summarize the BWR fuel cladding performance under these advanced and improved water chemistry conditions and how these changes have affected the goal to reach zero fuel failures. The paper will also provide a brief summary of some of the results of hot cell PIE, results of crud composition evaluation, crud spallation, oxide thickness measurements, hydrogen content in the cladding and some fuel failure observations. (author) Key Words: Boiling Water Reactor, Fuel Performance, Hydrogen Addition, Zinc Addition, Noble Metal Chemical Addition, Zero Leakers

  16. Air conditioning facilities in a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Kawasaki, Michitaka; Oka, Tsutomu

    1987-01-01

    Reprocessing plants are the facilities for separating the plutonium produced by nuclear reaction and unconsumed remaining uranium from fission products in the spent fuel taken out of nuclear reactors and recovering them. The fuel reprocessing procedure is outlined. In order to ensure safety in handling radioactive substances, triple confinement using vessels, concrete cells and buildings is carried out in addition to the prevention of criticality and radiation shielding, and stainless steel linings and drip trays are installed as occasion demands. The ventilation system in a reprocessing plant is roughly divided into three systems, that is, tower and tank ventilation system to deal with offgas, cell ventilation system for the cells in which main towers and tanks are installed, and building ventilation system. Air pressure becomes higher from tower and tank system to building system. In a reprocessing plant, the areas in a building are classified according to dose rate. The building ventilation system deals with green and amber areas, and the cell ventilation system deals with red area. These three ventilation systems are explained. Radiation monitors are installed to monitor the radiation dose rate and air contamination in working places. The maintenance and checkup of ventilation systems are important. (Kako, I.)

  17. Performance Analysis of Air Breathing Proton Exchange Membrane Fuel Cell Stack (PEMFCS) At Different Operating Condition

    Science.gov (United States)

    Sunil, V.; Venkata siva, G.; Yoganjaneyulu, G.; Ravikumar, V. V.

    2017-08-01

    The answer for an emission free power source in future is in the form of fuel cells which combine hydrogen and oxygen producing electricity and a harmless by product-water. A proton exchange membrane (PEM) fuel cell is ideal for automotive applications. A single cell cannot supply the essential power for any application. Hence PEM fuel cell stacks are used. The effect of different operating parameters namely: type of convection, type of draught, hydrogen flow rate, hydrogen inlet pressure, ambient temperature and humidity, hydrogen humidity, cell orientation on the performance of air breathing PEM fuel cell stack was analyzed using a computerized fuel cell test station. Then, the fuel cell stack was subjected to different load conditions. It was found that the stack performs very poorly at full capacity (runs only for 30 min. but runs for 3 hours at 50% capacity). Hence, a detailed study was undertaken to maximize the duration of the stack’s performance at peak load.

  18. The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

    International Nuclear Information System (INIS)

    Rohde, U.

    1998-01-01

    A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported. Fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D model. (author)

  19. 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)

  20. Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wiss, Thierry, E-mail: thierry.wiss@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Hiernaut, Jean-Pol [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Roudil, Danièle [Commissariat à l’Energie Atomique et aux Energie Alternatives, Centre de Marcoule, BP 30207 Bagnols-sur-Cèze (France); Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J.M.; Matzke, Hans-Joachim [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Weber, William J. [Department of Materials Science and Engineering, The University of Tennessee, Knoxville, TN 37996 (United States); Division of Materials Science and Technology, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2014-08-01

    Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted.

  1. High Burnup Fuel Behaviour under LOCA Conditions as Observed in Halden Reactor Experiments

    International Nuclear Information System (INIS)

    Kolstad, E.; Wiesenack, W.; Oberlander, B.; Tverberg, T.

    2013-01-01

    In the context of assessing the validity of safety criteria for loss of coolant accidents with high burnup fuel, the OECD Halden Reactor Project has implemented an integral in-pile LOCA test series. In this series, fuel fragmentation and relocation, axial gas communication in high burnup rods as affected by gap closure and fuel- clad bonding, and secondary cladding oxidation and hydriding are of major interest. In addition, the data are being used for code validation as well as model development and verification. So far, nine tests with irradiated fuel segments (burnup 40-92 MW.d.kg -1 ) from PWR, BWR and VVER commercial nuclear power plants have been carried out. The in-pile measurements and the PIE results show a good repeatability of the experiments. The paper describes the experimental setup as well as the principal features and main results of these tests. Fuel fragmentation and relocation have occurred to varying degrees in these tests. The paper compares the conditions leading to the presence or absence of fuel fragmentation, e.g., burnup and loss of constraint. Axial gas flow is an important driving force for clad ballooning, fuel relocation and fuel expulsion. The experiments have provided evidence that such gas flow can be impeded in high burnup fuel with a potential impact on the ballooning and fuel dispersal. Although the results of the Halden LOCA tests are, to some extent, amplified by conditions and features deliberately introduced into the test series, the fuel behaviour identified in the Halden tests has an impact on the safety assessment of high burnup fuel and should give rise to improvements of the predictive capabilities of LOCA modelling codes. (author)

  2. Synthesis on power electronics for large fuel cells: From power conditioning to potentiodynamic analysis technique

    International Nuclear Information System (INIS)

    De Bernardinis, Alexandre

    2014-01-01

    Highlights: • Active load for fuel cell managing electrical drive constraints: frequency and current ripple can be adjusted independently. • Multi-port resonant soft-switched topology for power management of a thirty kilowatt segmented PEM fuel cell. • Splitting current control strategy for power segmented PEM fuel cell in case of a segment is under fault. • Reversible Buck topology for large fuel cell with control of the fuel cell potential linked to current density nonlinearity. - Abstract: The work addressed in this paper deals with a synthesis on power electronic converters used for fuel cells. The knowledge gap concerns conceptually different electronic converter architectures for PEM (Proton Exchange Membrane) fuel cells able to perform three types of functionalities: The first one is the capacity of emulating an active load representative of electrical drive constraints. In that case, frequency and fuel cell current ripple can be set independently to investigate the dynamic behavior of the fuel cell. The second one is power conditioning applied to large high power and segmented fuel cell systems (“Large” represents several tens of cells and multi-kilowatt stacks), which is a non trivial consideration regarding the topological choices to be made for improving efficiency, compactness and ensure operation under faulty condition. A multi-port resonant isolated boost topology is analyzed enabling soft switching over a large operating range for a thirty kilowatt segmented fuel cell. A splitting current control strategy in case of a segment is under fault is proposed. Each considered converter topologies meet specific constraints regarding fuel cell stack design and power level. The third functionality is the ability for the power electronics to perform analysis and diagnosis techniques, like the cyclic voltammetry on large PEM fuel cell assemblies. The latter technique is an uncommon process for large fuel cell stacks since it is rather performed on

  3. Some conditions and prospects of going to a closed fuel cycle in Russia

    International Nuclear Information System (INIS)

    Lependin, A.V.; Oussanov, V.; Lependina, E.V.

    2000-01-01

    Nuclear policy in Russia is based on the necessity of closure of the nuclear fuel cycle. At the same time, the schedule of such a move is not yet defined. In this study, some conditions and possible time frames of taking the nuclear fuel cycle of Russia to closure are discussed. Naturally, the main condition is the revival of the Russian economy wherein nuclear power will turn out to be necessary in a number of Russian regions. The question is whether the closure of nuclear cycle strategy will be implemented in the near future or nuclear power will develop based on the open fuel cycle over a long period of time? (authors)

  4. Hydrogen generation at ambient conditions: application in fuel cells.

    Science.gov (United States)

    Boddien, Albert; Loges, Björn; Junge, Henrik; Beller, Matthias

    2008-01-01

    The efficient generation of hydrogen from formic acid/amine adducts at ambient temperature is demonstrated. The highest catalytic activity (TOF up to 3630 h(-1) after 20 min) was observed in the presence of in situ generated ruthenium phosphine catalysts. Compared to the previously known methods to generate hydrogen from liquid feedstocks, the systems presented here can be operated at room temperature without the need for any high-temperature reforming processes, and the hydrogen produced can then be directly used in fuel cells. A variety of Ru precursors and phosphine ligands were investigated for the decomposition of formic acid/amine adducts. These catalytic systems are particularly interesting for the generation of H2 for new applications in portable electric devices.

  5. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kupca, L [VUJE Inc. (Slovakia)

    2012-07-01

    In the Slovak Republic are under operation 6 units (4 in the Jaslovske Bohunice site, and 2 in the Mochovce), 2 units are under construction in Mochovce site. All units are WWER-440 type. The fresh fuel is imported from the Russian Federation. The spent fuel assemblies are stored in wet conditions in Bohunice Interim Storage Spent Fuel Facility (SFIS). By 15 July 2008, there were 8413 assemblies in SFIS. The objectives are: 1) Wet AR storage of spent fuel from the NPP Bohunice and Mochovce: Surveillance of conditions for spent fuel storage in the at-reactor (AR) storage pools of both NPP's (characteristics of pool water, corrosion product data); Visual control of storage pool components; Evaluation of storage conditions with respect to long-term stability (corrosion of fuel cladding, structural materials); 2) Wet SFIS storage at Bohunice: Measurement of spent fuel conditions during the long-term wet storage, activity data in the storage casks and amount of crud; Surveillance program for SFIS structural materials.

  6. The modeling of fuel rod behaviour under RIA conditions in the code DYN3D

    International Nuclear Information System (INIS)

    Rohde, U.

    2001-01-01

    A description of the fuel rod behaviour and heat transfer model used in the code DYN3D for nuclear reactor core dynamic simulations is given. Besides the solution of heat conduction equations in fuel and cladding, the model comprises a detailed description of heat transfer in the gas gap by conduction, radiation and fuel-cladding contact. The gas gap behaviour is modeled in a mechanistic way taking into account transient changes of the gas gap parameters based on given conditions for the initial state. Thermal, elastic and plastic deformations of fuel and cladding are taken into account within 1D approximation. A creeping law for time-dependent estimation of plastic deformations is implemented. Metal-water reaction of the cladding material in the high temperature region is considered. The cladding-coolant heat transfer regime map covers the region from one-phase liquid convection to dispersed flow with superheated steam. Special emphasis is put on taking into account the impact of thermodynamic non-equilibrium conditions on heat transfer. For the validation of the model, experiments on fuel rod behaviour during RIAs carried out in Russian and Japanese pulsed research reactors with shortened probes of fresh fuel rods are calculated. Comparisons between calculated and measured results are shown and discussed. It is shown, that the fuel rod behaviour is significantly influenced by plastic deformation of the cladding, post crisis heat transfer with sub-cooled liquid conditions and heat release from the metal-water reaction. Numerical studies concerning the fuel rod behaviour under RIA conditions in power reactors are reported on. It is demonstrated, that the fuel rod behaviour at high pressures and flow rates in power reactors is different from the behaviour under atmospheric pressure and stagnant flow conditions in the experiments. The mechanisms of fuel rod failure for fresh and burned fuel reported from the literature can be qualitatively reproduced by the DYN3D

  7. Mechanical stress analysis for a fuel rod under normal operating conditions

    International Nuclear Information System (INIS)

    Pino, Eddy S.; Giovedi, Claudia; Serra, Andre da Silva; Abe, Alfredo Y.

    2013-01-01

    Nuclear reactor fuel elements consist mainly in a system of a nuclear fuel encapsulated by a cladding material subject to high fluxes of energetic neutrons, high operating temperatures, pressure systems, thermal gradients, heat fluxes and with chemical compatibility with the reactor coolant. The design of a nuclear reactor requires, among a set of activities, the evaluation of the structural integrity of the fuel rod submitted to different loads acting on the fuel rod and the specific properties (dimensions and mechanical and thermal properties) of the cladding material and coolant, including thermal and pressure gradients produced inside the rod due to the fuel burnup. In this work were evaluated the structural mechanical stresses of a fuel rod using stainless steel as cladding material and UO 2 with a low degree of enrichment as fuel pellet on a PWR (pressurized water reactor) under normal operating conditions. In this sense, tangential, radial and axial stress on internal and external cladding surfaces considering the orientations of 0 deg, 90 deg and 180 deg were considered. The obtained values were compared with the limit values for stress to the studied material. From the obtained results, it was possible to conclude that, under the expected normal reactor operation conditions, the integrity of the fuel rod can be maintained. (author)

  8. Risk-informed optimal routing of ships considering different damage scenarios and operational conditions

    International Nuclear Information System (INIS)

    Decò, Alberto; Frangopol, Dan M.

    2013-01-01

    The aim of this paper is the development of a risk-informed decision tool for the optimal mission-oriented routing of ships. The strength of the hull is investigated by modeling the midship section with finite elements and by analyzing different damage levels depending on the propagation of plastification throughout the section. Vertical and horizontal flexural interaction is investigated. Uncertainties associated with geometry and material properties are accounted for by means of the implementation of the response surface method. Load effects are evaluated using strip theory. Reliability analysis is performed for several ship operational conditions and considering four different limit states. Then, risk is assessed by including the direct losses associated with five investigated damage states. The effects of corrosion on aged ships are included in the proposed approach. Polar representation of load effects, reliability, and direct risk are presented for a large spectrum of operational conditions. Finally, the optimal routing of ships is obtained by minimizing both the estimated time of arrival and the expected direct risk, which are clearly conflicting objectives. The optimization process provides feasible solutions belonging to the Pareto front. The proposed approach is applied to a Joint High Speed Sealift

  9. Construction Condition and Damage Monitoring of Post-Tensioned PSC Girders Using Embedded Sensors.

    Science.gov (United States)

    Shin, Kyung-Joon; Lee, Seong-Cheol; Kim, Yun Yong; Kim, Jae-Min; Park, Seunghee; Lee, Hwanwoo

    2017-08-10

    The potential for monitoring the construction of post-tensioned concrete beams and detecting damage to the beams under loading conditions was investigated through an experimental program. First, embedded sensors were investigated that could measure pre-stress from the fabrication process to a failure condition. Four types of sensors were installed on a steel frame, and the applicability and the accuracy of these sensors were tested while pre-stress was applied to a tendon in the steel frame. As a result, a tri-sensor loading plate and a Fiber Bragg Grating (FBG) sensor were selected as possible candidates. With those sensors, two pre-stressed concrete flexural beams were fabricated and tested. The pre-stress of the tendons was monitored during the construction and loading processes. Through the test, it was proven that the variation in thepre-stress had been successfully monitored throughout the construction process. The losses of pre-stress that occurred during a jacking and storage process, even those which occurred inside the concrete, were measured successfully. The results of the loading test showed that tendon stress and strain within the pure span significantly increased, while the stress in areas near the anchors was almost constant. These results prove that FBG sensors installed in a middle section can be used to monitor the strain within, and the damage to pre-stressed concrete beams.

  10. Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

    International Nuclear Information System (INIS)

    Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

    1999-02-01

    Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

  11. Finite element modelling of different CANDU fuel bundle types in various refuelling conditions

    International Nuclear Information System (INIS)

    Roman, M. R.; Ionescu, D. V.; Olteanu, G.; Florea, S.; Radut, A. C.

    2016-01-01

    The objective of this paper is to develop a finite element model for static strength analysis of the CANDU standard with 37 elements fuel bundle and the SEU43 with 43 elements fuel bundle design for various refuelling conditions. The computer code, ANSYS7.1, is used to simulate the axial compression in CANDU type fuel bundles subject to hydraulic drag loads, deflection of fuel elements, stresses and displacements in the end plates. Two possible situations for the fuelling machine side stops are considered in our analyses, as follows: the last fuel bundle is supported by the two side stops and a side stop can be blocked therefore, the last fuel bundle is supported by only one side stop. The results of the analyses performed are briefly presented and also illustrated in a graphical form. The finite element model developed in present study is verified against test results for endplate displacement and element bowing obtained from strength tests with fuel bundle string and fuelling machine side-stop simulators. Comparison of ANSYS model predictions with these experimental results led to a very good agreement. Despite the difference in hydraulic load between SEU43 and CANDU standard fuel bundles strings, the maximum stress in the SEU43 endplate is about the same with the maximum stress in the CANDU standard endplate. The comparative assessment reveals that SEU43 fuel bundle is able to withstand high flow rate without showing a significant geometric instability. (authors)

  12. Impact of fission gas on irradiated PWR fuel behaviour at extended burnup under RIA conditions

    International Nuclear Information System (INIS)

    Lemoine, F.; Schmitz, F.

    1996-01-01

    With the world-wide trend to increase the fuel burnup at discharge of the LWRs, the reliability of high burnup fuel must be proven, including its behaviour under energetic transient conditions, and in particular during RIAs. Specific aspects of irradiated fuel result from the increasing retention of gaseous and volatile fission products with burnup. The potential for swelling and transient expansion work under rapid heating conditions characterizes the high burnup fuel behaviour by comparison to fresh fuel. This effect is resulting from the steadily increasing amount of gaseous and volatile fission products retained inside the fuel structure. An attempt is presented to quantify the gas behaviour which is motivated by the results from the global tests both in CABRI and in NSRR. A coherent understanding of specific results, either transient release or post transient residual retention has been reached. The early failure of REP Na1 with consideration given to the satisfactory behaviour of the father rod of the test pin at the end of the irradiation (under load follow conditions) is to be explained both by the transient loading from gas driven fuel swelling and from the reduced clad resistance due to hydriding. (R.P.)

  13. Encapsulation technology of MR6 spent fuel and quality analysis of the EK-10 and WWR-SM spent fuel stored more than 30 years in wet conditions

    Energy Technology Data Exchange (ETDEWEB)

    Borek-Kruszewska, E.; Bykowski, W.; Chwaszczewski, S.; Czajkowski, W.; Madry, M. [Institute of Atomic Energy, Otwock -Swierk (Poland)

    2002-07-01

    The research reactor MARIA has been in operation for more than twenty years and all the spent fuel assemblies used since the first commissioning of the reactor are stored in wet facility on site. The present paper deals with the spent fuel MR-6 encapsulation technology in MARIA reactor. The encapsulated spent MR-6 fuel will be stored under water in the same pool unless some other solution is available. The capsules made of stainless steel are capable to accommodate one MR-6 fuel assembly. The encapsulation process is performed in the hot cell by the MARIA reactor. The spent fuel having its leg cut off is loaded to the transport cylinder manually and next transferred to a trolley. The trolley is moving to a position directly below the entrance to the hot cell and the spent fuel is entering the hot cell. The spent fuel assembly is then put into the drying cell. Dried out spent fuel is moved into the capsule mounted on the grip of the machine. Next, the capsule lid is pressed in and welded. After the leak test and filling up with helium the capsule returns from the hot cell to the pool. The hermetic capsule is sunk back into the water and positioned in the separator . The results presented earlier show, that the limiting time of WWR-SM and Ek-10 type spent fuel residence in wet storage is about 40-45 years. Therefore, the systematic quality investigation of all Ek-10 fuel elements and WWR-SM fuel assemblies discharged from EWA reactor in the period of 1959-1969 was performed. Altogether, about 2500 Ek-10 fuel elements and 47 WWR-SM fuel assemblies were investigated. The results of these investigations are presented in the present work. The sipping test, visual investigation and ultrasonic techniques were used for that purpose. The radioactive isotope Cs-137 was used as the indicator of fission product release from the fuel assembly. Taking into account the value of Cs-137 release from damaged WWR-SM fuel assembly the criteria of damaged fuel assembly were proposed. It

  14. The effect of test configuration on the true operating conditions of PEM fuel cells. Paper no. IGEC-1-124

    International Nuclear Information System (INIS)

    Simpson, T.; Li, X.

    2005-01-01

    The operating conditions of a single PEM fuel cell can be significantly affected by the configuration in which the fuel cell test is setup. This study investigates the effect on the gas dewpoint temperature of not insulating the inlet fittings to a PEM fuel cell and the effect of non-optimal stack control thermocouple placement on fuel cell stack operating temperature. Both of these setup configurations can significantly affect fuel cell membrane humidification conditions, especially in a single fuel cell as demonstrated through the sample test conditions presented in this paper. (author)

  15. Power Conditioning of Fuel Cell Systems in Portable Applications

    Energy Technology Data Exchange (ETDEWEB)

    Moreno-Benitez, E.; Brey, J. J.; Rodriguez-Bordallo, C.; Carrasco, J. M.; Galvan, E.

    2005-07-01

    The achieving of high performance and long useful life are the two fundamental objectives of portable application designers. Cost and size conditions make these objectives more complex and always lead to a compromise solution having to be reached. The most significant parameters as regards portables devices are cost, efficiency (useful life), output crimps and noise, and quiescent current. Most portable products have two fundamental operating modes: active and standby. During the active period, current consumption is generally high and this means that excellent conversion is essential in order to maximize the useful life of the device that supplies current and voltage. However, most portable devices spend most of their time on standby and draw little energy from the source. It is equally important for the source to be very efficient under these conditions. This means that the quiescent current from the source (the current that supplies in low or nil load conditions) must be much lower than the load current in order to maintain high efficiency. Topologies Different power conditioning topologies to be used in portable applications are indicated with their corresponding advantages and inconveniences being specified. Low dropout voltage regulator (LDO) This type of conditioning is one of minimum cost, noise and quiescent current. This makes this device a favorite for many applications. Its external components are minimal: usually a bypass capacity. Its efficiency, although poor when Vin is much greater than Vout, increases greatly when their values are somewhat similar. In this event, the benefits of using LDOs are almost impossible to beat. In fact, these circuits are much used to reach output voltages of up to 3 volts. (Author)

  16. A literature survey on the dissolution mechanism of spent fuel under disposal conditions

    International Nuclear Information System (INIS)

    Ollila, Kaija

    1989-06-01

    In Finland spent nuclear fuel is planned to be disposed of at large depths in crystalline bedrock. As part of the YJT (Nuclear Waste Commission of Finnish Power Companies) - program, the solubiliy and dissolution mechanisms of unirradiated UO 2 are experimentally investigated as a function of groundwater conditions. This study is a literature survey on the leaching and dissolution studies carried out with spent fuel. It consists first a review on characterization studies of spent fuel. Then the solubilities and release mechanisms of the radionuclides from spent fuel in granitic or related groundwaters are discussed, including the dissolution of UO 2 matrix, and the leaching of fission products and actinides. Lastly approaches to modelling the dissolution of spent fuel are shortly discussed

  17. Stress Calculation of a TRISO Coated Particle Fuel by Using a Poisson's Ratio in Creep Condition

    International Nuclear Information System (INIS)

    Cho, Moon-Sung; Kim, Y. M.; Lee, Y. W.; Jeong, K. C.; Kim, Y. K.; Oh, S. C.; Kim, W. K.

    2007-01-01

    KAERI, which has been carrying out the Korean VHTR (Very High Temperature modular gas cooled Reactor) project since 2004, has been developing a performance analysis code for the TRISO coated particle fuel named COPA (COated Particle fuel Analysis). COPA predicts temperatures, stresses, a fission gas release and failure probabilities of a coated particle fuel in normal operating conditions. KAERI, on the other hand, is developing an ABAQUS based finite element(FE) model to cover the non-linear behaviors of a coated particle fuel such as cracking or debonding of the TRISO coating layers. Using the ABAQUS based FE model, verification calculations were carried out for the IAEA CRP-6 benchmark problems involving creep, swelling, and pressure. However, in this model the Poisson's ratio for elastic solution was used for creep strain calculation. In this study, an improvement is made for the ABAQUS based finite element model by using the Poisson's ratio in creep condition for the calculation of the creep strain rate. As a direct input of the coefficient in a creep condition is impossible, a user subroutine for the ABAQUS solution is prepared in FORTRAN for use in the calculations of the creep strain of the coating layers in the radial and hoop directions of the spherical fuel. This paper shows the calculation results of a TRISO coated particle fuel subject to an irradiation condition assumed as in the Miller's publication in comparison with the results obtained from the old FE model used in the CRP-6 benchmark calculations

  18. Fuel Cooling in Absence of Forced Flow at Shutdown Condition with PHTS Partially Drained

    Energy Technology Data Exchange (ETDEWEB)

    Parasca, L.; Pecheanu, D.L., E-mail: laurentiu.parasca@cne.ro, E-mail: doru.pecheanu@cne.ro [Cernavoda Nuclear Power Plant, Cernavoda (Romania)

    2014-09-15

    During the plant outage for maintenance on primary side (e.g. for the main Heat Transport System pumps maintenance, the Steam Generators inspection), there are situations which require the primary heat transport system (HTS) drainage to a certain level for opening the circuit. The primary fuel heat sink for this configuration is provided by the shutdown cooling system (SDCS). In case of losing the forced cooling (e.g. due to the loss of SDCS, design basis earthquake-DBE), flow conditions in the reactor core may become stagnant. Inside the fuel channels, natural circulation phenomena known as Intermittent Buoyancy Induced Flow (IBIF) will initiate, providing an alternate heat sink mechanism for the fuel. However, this heat sink is effective only for a limited period of time (recall time). The recall time is defined as the elapsed time until the water temperature in the HTS headers exceeds a certain limit. Until then, compensatory measures need to be taken (e.g. by re-establishing the forced flow or initiate Emergency Core Cooling system injection) to preclude fuel failures. The present paper briefly presents the results of an analysis performed to demonstrate that fuel temperature remains within acceptable limits during IBIF transient. One of the objectives of this analysis was to determine the earliest moment since the reactor shut down when maintenance activities on the HTS can be started such that IBIF is effective in case of losing the forced circulation. The resulting peak fuel sheath and pressure tube temperatures due to fuel heat up shall be within the acceptable limits to preclude fuel defect or fuel channel defects.Thermalhydraulic circuit conditions were obtained using a CATHENA model for the primary side of HTS (drained to a certain level), an ECC system model and a system model for SDCS. A single channel model was developed in GOTHIC code for the fuel assessment analysis. (author)

  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. Influence of shock absorber condition on pavement fatigue using relative damage concept

    Directory of Open Access Journals (Sweden)

    Pablo Kubo

    2015-12-01

    Full Text Available Considering the importance of the road transportation nowadays, concerns related to pavement deterioration and maintenance have become relevant subjects. Especially for commercial vehicles, the vertical dynamic load (characterized by the tire-road interaction is directly related to wear on the road surface. Given this, the main objective of this paper is to analyse effects of vertical loads applied on the flexible pavement, considering the variation of the condition of shock absorbers from a truck's front suspension. The measurements were performed on a rigid truck, with 2 steering front axles, in a durability test track located in Brazil. With a constant load of 6 tons on the front suspension (the maximum allowed load on front axles according to Brazilian legislation, 3 different shock absorber conditions were evaluated: new, used and failed. By applying the relative damage concept, it is possible to conclude that the variation of the shock absorber conditions will significantly affect the vertical load applied on the pavement. Although the results clearly point to a dependent relationship between the load and the condition of the shock absorbers, it is recommended to repeat the same methodology, in future to analyse the influence of other quarter car model variants (such as spring rate, mass and tire spring stiffness.

  1. 14 CFR 91.151 - Fuel requirements for flight in VFR conditions.

    Science.gov (United States)

    2010-01-01

    ... begin a flight in an airplane under VFR conditions unless (considering wind and forecast weather conditions) there is enough fuel to fly to the first point of intended landing and, assuming normal cruising speed— (1) During the day, to fly after that for at least 30 minutes; or (2) At night, to fly after that...

  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. The behaviour of water-cooled reactor fuel rods in steady state and transient conditions

    International Nuclear Information System (INIS)

    Strupczewski, A.; Marks, P.

    1997-01-01

    In this report, the results of temperature field and filling gas pressure calculations by means of contemporary calculational models for a WWER-440 and WWER-1000 type fuel rod at low and high burnup operating under steady-state conditions are presented. A review of in-core temperature and pressure measurements for various types of LWR fuel is also included. Basing on calculational and collected measured data, the behaviour of fuel cladding during large and small break LOCA, is estimated with special emphasis on their oxidation and failure resistance. (author)

  4. The Influence Of Mass Fraction Of Dressed Coal On Ignition Conditions Of Composite Liquid Fuel Droplet

    Directory of Open Access Journals (Sweden)

    Shlegel Nikita E.

    2015-01-01

    Full Text Available The laws of condition modification of inert heat and ignition in an oxidant flow of composite liquid fuel droplet were studied by the developed experimental setup. Investigations were for composite liquid fuel composition based on the waste of bituminous and nonbaking coal processing, appropriate carbon dust, water, used motor oil. The characteristics of boundary layer inertia heat of composite liquid fuel droplet, thermal decomposition of coal organic part, the yield of volatiles and evaporation of liquid combustion component, ignition of the gas mixture and coke residue were defined.

  5. High-flux He+ irradiation effects on surface damages of tungsten under ITER relevant conditions

    International Nuclear Information System (INIS)

    Liu, Lu; Liu, Dongping; Hong, Yi; Fan, Hongyu; Ni, Weiyuan; Yang, Qi; Bi, Zhenhua; Benstetter, Günther; Li, Shouzhe

    2016-01-01

    A large-power inductively coupled plasma source was designed to perform the continuous helium ions (He + ) irradiations of polycrystalline tungsten (W) under International Thermonuclear Experimental Reactor (ITER) relevant conditions. He + irradiations were performed at He + fluxes of 2.3 × 10 21 –1.6 × 10 22 /m 2  s and He + energies of 12–220 eV. Surface damages and microstructures of irradiated W were observed by scanning electron microscopy. This study showed the growth of nano-fuzzes with their lengths of 1.3–2.0 μm at He + energies of >70 eV or He + fluxes of >1.3 × 10 22 /m 2  s. Nanometer-sized defects or columnar microstructures were formed in W surface layer due to low-energy He + irradiations at an elevated temperature (>1300 K). The diffusion and coalescence of He atoms in W surface layers led to the growth and structures of nano-fuzzes. This study indicated that a reduction of He + energy below 12–30 eV may greatly decrease the surface damage of tungsten diverter in the fusion reactor.

  6. Calculation of the Incremental Conditional Core Damage Probability on the Extension of Allowed Outage Time

    International Nuclear Information System (INIS)

    Kang, Dae Il; Han, Sang Hoon

    2006-01-01

    RG 1.177 requires that the conditional risk (incremental conditional core damage probability and incremental conditional large early release probability: ICCDP and ICLERP), given that a specific component is out of service (OOS), be quantified for a permanent change of the allowed outage time (AOT) of a safety system. An AOT is the length of time that a particular component or system is permitted to be OOS while the plant is operating. The ICCDP is defined as: ICCDP = [(conditional CDF with the subject equipment OOS)- (baseline CDF with nominal expected equipment unavailabilities)] [duration of the single AOT under consideration]. Any event enabling the component OOS can initiate the time clock for the limiting condition of operation for a nuclear power plant. Thus, the largest ICCDP among the ICCDPs estimated from any occurrence of the basic events for the component fault tree should be selected for determining whether the AOT can be extended or not. If the component is under a preventive maintenance, the conditional risk can be straightforwardly calculated without changing the CCF probability. The main concern is the estimations of the CCF probability because there are the possibilities of the failures of other similar components due to the same root causes. The quantifications of the risk, given that a subject equipment is in a failed state, are performed by setting the identified event of subject equipment to TRUE. The CCF probabilities are also changed according to the identified failure cause. In the previous studies, however, the ICCDP was quantified with the consideration of the possibility of a simultaneous occurrence of two CCF events. Based on the above, we derived the formulas of the CCF probabilities for the cases where a specific component is in a failed state and we presented sample calculation results of the ICCDP for the low pressure safety injection system (LPSIS) of Ulchin Unit 3

  7. Analysis of accelerated degradation of a HT-PEM fuel cell caused by cell reversal in fuel starvation condition

    DEFF Research Database (Denmark)

    Zhou, Fan; Andreasen, Søren Juhl; Kær, Søren Knudsen

    2015-01-01

    This paper reports an accelerated degradation test of a high temperature PEM fuel cell under repeated H2 starvation condition. The H2 stoichiometry is cycled between 3.0 and 0.8 every 2 min during the test. The experimental results show that the polarity of the fuel cell is reversed under H2......, there is only a slight decrease in open circuit voltage of the fuel cell which implies the membrane is not affected by the test. The electrochemical impedance spectrum measurement shows that the H2 starvation can cause significant increase in the ohmic resistance and charge transfer resistance. By looking...... starvation condition, and the cell performance indicated by cell voltage at H2 stoichiometry of 3.0 declines from 0.59 V to 0.41 V in 19 cycles. Since CO2 is detected in anode exhaust under H2 starvation condition, carbon corrosion is believed to be the reason for the degradation in this test. After the test...

  8. 40 CFR 80.527 - Under what conditions may motor vehicle diesel fuel subject to the 15 ppm sulfur standard be...

    Science.gov (United States)

    2010-07-01

    ... vehicle diesel fuel subject to the 15 ppm sulfur standard be downgraded to motor vehicle diesel fuel... Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Motor Vehicle Diesel Fuel Standards and Requirements § 80.527 Under what conditions may motor vehicle diesel fuel subject to the 15...

  9. Under what conditions is recognition spared relative to recall after selective hippocampal damage in humans?

    Science.gov (United States)

    Holdstock, J S; Mayes, A R; Roberts, N; Cezayirli, E; Isaac, C L; O'Reilly, R C; Norman, K A

    2002-01-01

    The claim that recognition memory is spared relative to recall after focal hippocampal damage has been disputed in the literature. We examined this claim by investigating object and object-location recall and recognition memory in a patient, YR, who has adult-onset selective hippocampal damage. Our aim was to identify the conditions under which recognition was spared relative to recall in this patient. She showed unimpaired forced-choice object recognition but clearly impaired recall, even when her control subjects found the object recognition task to be numerically harder than the object recall task. However, on two other recognition tests, YR's performance was not relatively spared. First, she was clearly impaired at an equivalently difficult yes/no object recognition task, but only when targets and foils were very similar. Second, YR was clearly impaired at forced-choice recognition of object-location associations. This impairment was also unrelated to difficulty because this task was no more difficult than the forced-choice object recognition task for control subjects. The clear impairment of yes/no, but not of forced-choice, object recognition after focal hippocampal damage, when targets and foils are very similar, is predicted by the neural network-based Complementary Learning Systems model of recognition. This model postulates that recognition is mediated by hippocampally dependent recollection and cortically dependent familiarity; thus hippocampal damage should not impair item familiarity. The model postulates that familiarity is ineffective when very similar targets and foils are shown one at a time and subjects have to identify which items are old (yes/no recognition). In contrast, familiarity is effective in discriminating which of similar targets and foils, seen together, is old (forced-choice recognition). Independent evidence from the remember/know procedure also indicates that YR's familiarity is normal. The Complementary Learning Systems model can

  10. Fuel performance under normal PWR conditions: A review of relevant experimental results and models

    Science.gov (United States)

    Charles, M.; Lemaignan, C.

    1992-06-01

    Experiments conducted at Grenoble (CEA/DRN) over the past 20 years in the field of nuclear fuel behaviour are reviewed. Of particular concern is the need to achieve a comprehensive understanding of and subsequently overcome the limitations associated with high burnup and load-following conditions (pellet-cladding interaction (PCI), fission gas release (FGR), water-side corrosion). A general view is given of the organization of research work as well as some experimental details (irradiation, postirradiation examination — PIE). Based on various experimental programmes (Cyrano, Medicis, Anemone, Furet, Tango, Contact, Cansar, Hatac, Flog, Decor), the main contributions of the thermomechanical behaviour of a PWR fuel rod are described: thermal conductivity, in-pile densification, swelling, fission gas release in steady state and moderate transient conditions, gap thermal conductance, formation of primary and secondary ridges under PCI conditions. Specific programmes (Gdgrif, Thermox, Grimox) are devoted to the behaviour of particular fuels (gadolinia-bearing fuel, MOX fuel). Moreover, microstructure-based studies have been undertaken on fission gas release (fine analysis of the bubble population inside irradiated fuel samples), and on cladding behaviour (PCI related studies on stress-corrosion cracking (SCO, irradiation effects on zircaloy microstructure).

  11. Effects of burnup on fission product release and implications for severe fuel damage events

    International Nuclear Information System (INIS)

    Appelhans, A.D.; Cronenberg, A.W.; Carboneau, M.L.

    1984-01-01

    Xe, Kr, and I fission-product release data from (a) Halden tests where release in intact rods was measured during irradiation at burnups to 18,000 MWd/t and fuel temperatures of 800 to 1800 0 K, and (b) Power Burst Facility (PBF) tests where trace-irradiated fuel (approx. = 90 MWd/t) was driven to temperatures of >2400 0 K and fuel liquefaction occurred are discussed and related to fuel morphology. Results from both indicate that the fission-product morphology and fuel restructuring govern release behavior. The Halden tests show low release at beginning of life with a 10-fold increase at burnups in excess of 10,000 MWd/t, due to the development of grain boundary interlinkage at higher burnups. Such dependence of release on morphology characteristics is consistent with findings from the PBF tests, where for trace-irradiated fuel, the absence of interlinkage accounts for the low release rates observed during initial fuel heatup, with subsequent enhanced Xe, Kr, and I release via liquefaction or quench-induced destruction of the grain structure. Morphology is also shown to influence the chemical release form of I and Cs fission products

  12. TRISO-Coated Fuel Durability Under Extreme Conditions

    International Nuclear Information System (INIS)

    2014-01-01

    The PIs propose to examine TRISO-coated particles (SiC and ZrC coatings) in an integrated two-part study. In the first part, experiments will be performed to assess the reaction kinetics of the carbides under CO-CO2 environments at temperatures up to 1800 degree C. Kinetic model will be applied to describe the degradation. Scanning and transmission electron microscopy will be employed to establish the chemical and microstructure evolution under the imposed environmental conditions. The second part of the proposed work focuses on establishing the role of the high temperature, environmental exposure described above on the mechanical behavior of TRISO-coated particles. Electron microscopy and other advanced techniques will be subsequently performed to evaluate failure mechanisms. The work is expected to reveal relationships between corrosion reactions, starting material characteristics (polytype of SiC, impurity concentration, flaw distribution), flaw healing behavior, and crack growth.

  13. Prediction of pressure tube fretting-wear damage due to fuel vibration

    International Nuclear Information System (INIS)

    Yetisir, M.; Fisher, N.J.

    1997-01-01

    Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington Nuclear Generating Station (NGS) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGSs. (orig.)

  14. Prediction of pressure tube fretting-wear damage due to fuel vibration

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M; Fisher, N J [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1996-12-31

    Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington NGS (nuclear generating station) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGS`s (nuclear generating stations). (author). 12 refs., 2 tabs., 11 figs.

  15. Internal Nozzle Flow Simulations of Gasoline-Like Fuels under Diesel Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Torelli, R.; Som, S.; Pei, Y.; Zhang, Yu; Traver, Michael

    2017-05-15

    Spray formation in internal combustion engines with direct injection is strictly correlated with internal nozzle flow characteristics, which are in turn influenced by fuel physical properties and injector needle motion. This paper pre-sents a series of 3D simulations that model the in-nozzle flow in a 5-hole mini-sac diesel injector. Two gasoline-like naphtha fuels, namely full-range and light naphtha, were tested under operating conditions typical of diesel applica-tions and were compared with n-dodecane, selected from a palette used as diesel surrogates. Validated methodolo-gies from our previous work were employed to account for realistic needle motion. The multi-phase nature of the problem was described by the mixture model assumption with the Volume of Fluid method. Cavitation effects were included by means of the Homogeneous Relaxation Model and turbulence closure was achieved with the Standard k-ε model in an Unsteady Reynolds-Averaged Navier-Stokes formulation. The results revealed that injector perfor-mance and propensity to cavitation are influenced by the fuel properties. Analyses of several physical quantities were carried out to highlight the fuel-to-fuel differences in terms of mass flow rate, discharge coefficients, and fuel vapor volume fraction inside the orifices. A series of parametric investigations was also performed to assess the fuel response to varied fuel injection temperature, injection pressure, and cross-sectional orifice area. For all cases, the strict correlation between cavitation magnitude and saturation pressure was confirmed. Owing to their higher volatil-ity, the two gasoline-like fuels were characterized by higher cavitation across all the simulated conditions. Occur-rence of cavitation was mostly found at the needle seat and at the orifice inlets during the injection event’s transient, when very small gaps exist between the needle and its seat. This behavior tended to disappear at maximum needle lift, where cavitation was

  16. Enhancement of laser induced damage threshold of fused silica by acid etching combined with UV laser conditioning

    International Nuclear Information System (INIS)

    Chen Meng; Xiang Xia; Jiang Yong; Zu Xiaotao; Yuan Xiaodong; Zheng Wanguo; Wang Haijun; Li Xibin; Lu Haibing; Jiang Xiaodong; Wang Chengcheng

    2010-01-01

    Acid etching combined with UV laser conditioning is developed to enhance the laser induced damage threshold (LIDT) of fused silica. Firstly, the fused silica is etched for 1 ∼ 100 min with a buffered 1% HF solution. After acid etching, its transmittance, surface roughness and LIDT are measured. The results reveal that the fused silica has the highest LIDT and transmittance after etching for 10 min. Then UV laser (355 nm) conditioning is adopted to process the 10-min-etched fused silica. When the laser fluence is below 60% of fused silica's zero probability damage threshold, the LIDT increases gradually with the increase of laser conditioning fluence. However, the LIDT rapidly decreases to be lower than the threshold of the 10-min-etched fused silica when the conditioning fluence is up to 80% of the threshold. Proper acid etching and laser conditioning parameters will effectively enhance the laser damage resistance of fused silica. (authors)

  17. Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States); Mei, Zhigang [Argonne National Lab. (ANL), Argonne, IL (United States); Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-10

    Uranium silicide (U3Si2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U3Si2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U3Si2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuel material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U3Si2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U3Si2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U3Si2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U3Si2 fuel as an accident

  18. Crevice corrosion of titanium under nuclear fuel waste conditions

    International Nuclear Information System (INIS)

    Ikeda, B.M.; Bailey, M.G.; Clarke, C.F.; Shoesmith, D.W.

    1989-11-01

    This report describes our experimental program to investigate the localized corrosion of ASTM Grade-2 titanium. In particular, it describes the study of the crevice corrosion of titanium, the process most likely to lead to the failure of nuclear waste containers constructed from this material. The basic mechanisms of crevice corrosion are discussed in detail. This is followed by a description of our laboratory program and the various immersion tests being performed under irradiated conditions. Experiments and tests were performed in NaCl solutions (generally 1.6 wt.%) and in simulated groundwater at 100 or 150 degrees C. A mechanism for crevice corrosion of titanium is presented and justified experimentally using an electrochemical approach. During the initiation stage, the crevice reaction is controlled by the kinetics of the anodic process. As oxygen is consumed in the propagation step, control switches to the cathodic step. Crevice corrosion eventually stops when the oxygen concentration falls to a low value. Propagation of the crevice can be restarted by the addition of oxygen. Our preliminary results on the effect of varying the iron content of the titanium are presented. An increase in iron content from 0.02 wt.% to 0.13 wt.% leads to passivation, as opposed to propagation, of the crevice. The effects of γ-irradiation, temperature, and oxygen concentration are also briefly discussed. Although our conclusions must be considered tentative, the effects of γ-irradiation appear to be beneficial. some crevice corrosion rates from longer-term immersion tests are also presented. Generally the rates are very low

  19. 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

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  20. 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

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  1. Simulating fuel behavior under transient conditions using FRAPTRAN and uncertainty analysis using Dakota

    International Nuclear Information System (INIS)

    Gomes, Daniel S.; Teixeira, Antonio S.

    2017-01-01

    Although regulatory agencies have shown a special interest in incorporating best estimate approaches in the fuel licensing process, fuel codes are currently licensed based on only the deterministic limits such as those seen in 10CRF50, and therefore, may yield unrealistic safety margins. The concept of uncertainty analysis is employed to more realistically manage this risk. In this study, uncertainties were classified into two categories: probabilistic and epistemic (owing to a lack of pre-existing knowledge in this area). Fuel rods have three sources of uncertainty: manufacturing tolerance, boundary conditions, and physical models. The first step in successfully analyzing the uncertainties involves performing a statistical analysis on the input parameters used throughout the fuel code. The response obtained from this analysis must show proportional index correlations because the uncertainties are globally propagated. The Dakota toolkit was used to analyze the FRAPTRAN transient fuel code. The subsequent sensitivity analyses helped in identifying the key parameters with the highest correlation indices including the peak cladding temperature and the time required for cladding failures. The uncertainty analysis was performed using an IFA-650-5 fuel rod, which was in line with the tests performed in the Halden Project in Norway. The main objectives of the Halden project included studying the ballooning and rupture processes. The results of this experiment demonstrate the accuracy and applicability of the physical models in evaluating the thermal conductivity, mechanical model, and fuel swelling formulations. (author)

  2. Simulating fuel behavior under transient conditions using FRAPTRAN and uncertainty analysis using Dakota

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel S.; Teixeira, Antonio S., E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    Although regulatory agencies have shown a special interest in incorporating best estimate approaches in the fuel licensing process, fuel codes are currently licensed based on only the deterministic limits such as those seen in 10CRF50, and therefore, may yield unrealistic safety margins. The concept of uncertainty analysis is employed to more realistically manage this risk. In this study, uncertainties were classified into two categories: probabilistic and epistemic (owing to a lack of pre-existing knowledge in this area). Fuel rods have three sources of uncertainty: manufacturing tolerance, boundary conditions, and physical models. The first step in successfully analyzing the uncertainties involves performing a statistical analysis on the input parameters used throughout the fuel code. The response obtained from this analysis must show proportional index correlations because the uncertainties are globally propagated. The Dakota toolkit was used to analyze the FRAPTRAN transient fuel code. The subsequent sensitivity analyses helped in identifying the key parameters with the highest correlation indices including the peak cladding temperature and the time required for cladding failures. The uncertainty analysis was performed using an IFA-650-5 fuel rod, which was in line with the tests performed in the Halden Project in Norway. The main objectives of the Halden project included studying the ballooning and rupture processes. The results of this experiment demonstrate the accuracy and applicability of the physical models in evaluating the thermal conductivity, mechanical model, and fuel swelling formulations. (author)

  3. Power, heat and chilliness with natural gas - fuel cells and air conditioning

    International Nuclear Information System (INIS)

    Krein, Stephan; Ruehling, Karin

    1999-01-01

    A new and innovative concept of the supply with power, heat and chilliness will realise in the new Malteser-hospital in Kamenz. The core of this demonstration-plant are a fuel cell, an adsorption cooling machine as well as multi-solar collectors. The fuel cell has two goals. Primary it produces power for the own demand. The selected dimension guarantees, that the power will consume nearly continuously. Secondly the produced heat of the fuel cell (and the solar-heat too) will use for heating and preparation of warm water. In the summer, the heat will use for the adsorption cooling machine, which produces chilliness for air-conditioning. The advantage in the face of common concepts of combining power and heat is the high-efficiently use of the fuel-energy for electric power generation on the one hand. Fuel cells work with high efficiency also at partial load. On the other hand, with the adsorption cooling machine the produced heat of fuel cell and multi-solar collectors can be used also in the summer. First experiences with this concept show, that an optimised co-operation of the components with an adaptive, self-learning control system based on the weather forecast as well as various storages for heat and chilliness can be achieve. A continuously operation, high fuel utilisation and reduced environmental pollution can be demonstrated. (author)

  4. RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

    International Nuclear Information System (INIS)

    Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

    1984-01-01

    The present report is the user's manual for the computer code RODSWELL developed at the JRC-Ispra for the thermomechanical analysis of LWR fuel rods under simulated loss-of-coolant accident (LOCA) conditions. The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The essential characteristics of the code are briefly outlined here. The model is particularly designed to perform a full thermal and mechanical analysis in both the azimuthal and radial directions. Thus, azimuthal temperature gradients arising from pellet eccentricity, flux tilt, arbitrary distribution of heat sources in the fuel and the cladding and azimuthal variation of coolant conditions can be treated. The code combines a transient 2-dimensional heat conduction code and a 1-dimentional mechanical model for the cladding deformation. The fuel rod is divided into a number of axial sections and a detailed thermomechanical analysis is performed within each section in radial and azimuthal directions. In the following sections, instructions are given for the definition of the data files and the semi-variable dimensions. Then follows a complete description of the input data. Finally, the restart option is described

  5. Effects of pellet shape on the fuel failure behavior under a RIA condition

    International Nuclear Information System (INIS)

    Hosokawa, Takanori; Hoshi, Tsutao; Yanagihara, Satoshi; Iwamura, Takamichi; Orita, Yoshihiko.

    1980-10-01

    The two types of fuel rods with different pellet shaped, i.e. flat pellets and dished pellets, were tested in the NSRR to investigate the effects of pellet shapes on the fuel failure behavior under an RIA condition and the results were compared with those of the chamfered pellet fuel rods which are used as the reference rod in the NSRR experiments. In addition, the deformation of pellets due to thermal expansion is calculated by using an FEM computer code. Through the above results, following conclusions are obtained. (1) In the experiments, insignificant differences on the cladding surface temperature responses and the appearance of post-irradiated rods are observed in each type of rods. (2) Evident differences on the deformation of fuel pellets have not appeared in the calculation. (3) In the RIA conditions, it is concluded that the fuel failure behavior and threshold energy might not be affected by pellet shape of which size is in the range of the current LWR's fuel rods. (author)

  6. 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)

  7. Sulfur emission from Victorian brown coal under pyrolysis, oxy-fuel combustion and gasification conditions.

    Science.gov (United States)

    Chen, Luguang; Bhattacharya, Sankar

    2013-02-05

    Sulfur emission from a Victorian brown coal was quantitatively determined through controlled experiments in a continuously fed drop-tube furnace under three different atmospheres: pyrolysis, oxy-fuel combustion, and carbon dioxide gasification conditions. The species measured were H(2)S, SO(2), COS, CS(2), and more importantly SO(3). The temperature (873-1273 K) and gas environment effects on the sulfur species emission were investigated. The effect of residence time on the emission of those species was also assessed under oxy-fuel condition. The emission of the sulfur species depended on the reaction environment. H(2)S, SO(2), and CS(2) are the major species during pyrolysis, oxy-fuel, and gasification. Up to 10% of coal sulfur was found to be converted to SO(3) under oxy-fuel combustion, whereas SO(3) was undetectable during pyrolysis and gasification. The trend of the experimental results was qualitatively matched by thermodynamic predictions. The residence time had little effect on the release of those species. The release of sulfur oxides, in particular both SO(2) and SO(3), is considerably high during oxy-fuel combustion even though the sulfur content in Morwell coal is only 0.80%. Therefore, for Morwell coal utilization during oxy-fuel combustion, additional sulfur removal, or polishing systems will be required in order to avoid corrosion in the boiler and in the CO(2) separation units of the CO(2) capture systems.

  8. Simulation of vibration modes of the fuel rod damaged due to the grid-to-rod fretting wear

    International Nuclear Information System (INIS)

    Kim, Kyu Tae; Kim, Kyeong Koo; Jang, Young Ki; Lee, Kyou Seok

    1997-01-01

    The flow-induced fuel fretting wear observed in some PWRs mainly proceeds in the grid-to-rod contact positions. The grid-to-rod fretting wear in the PWR fuel assembly depends on grid-to-rod gap size, its axial profile and flow-induced vibration. This paper describes the GRIDFORCE program which generates the axially dependent grid-to-rod gap size as a function of burnup. The axially dependent grid-to-rod gap profiles are employed to predict the fuel rod vibration mode shapes by the ANSYS code. With the help of the Paidousis empirical formula, this paper also calculates the fuel rod vibration amplitudes under various supporting conditions, which indicates that the increase of the number of unsupported mid-grids will increase the fuel rod vibration amplitude. On the other hand, the comparison of the predicted vibration mode shapes and the observed mid-grid fretting wear pattern indicates that the 1st and 6th vibration mode shapes under the supporting inactive condition at the mid-grids can simulate the observed mid-grid fretting wear profile. This paper also proposes design guidelines against the grid-to-rod fretting wear. (author). 3 refs., 8 figs

  9. 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

  10. Design retrofit to prevent damage due to heat transport pump operation under conditions of significant void

    International Nuclear Information System (INIS)

    Lam, K.F.

    1991-01-01

    The purpose of this paper is to provide a general review of certain key design areas which address the safety concerns of HT pump operation under conditions of significant void. To illustrate the challenges confronting designers and analysts, some of the highlights during the design of a protective system to prevent damage to HT piping and pump supports at Bruce NGS 'A' are outlined. The effects of this protective system on reactor safety are also discussed. HI pump operation under conditions of significant void offers a major challenge to designers and analysts to ensure that pump induced vibration and its effects on pump and piping are addressed. For an in-service station the search for a practical solution is often limited by existing. station equipment design and Layout. The diversity of design verification process requires a major commitment of engineering resources to ensure all. safety aspects meet the requirements of regulatory body. Work currently undertaken at Ontario Hydro Research Pump Test Complex on two-phase flow in pumps and piping may provide better prediction of vibration characteristics so that inherent conservativeness in fatigue Life prediction of HI system components can be reduced

  11. Design retrofit to prevent damage due to heat transport pump operation under conditions of significant void

    Energy Technology Data Exchange (ETDEWEB)

    Lam, K F [Bruce Engineering Department, In-Service Nuclear Projects, Ontario Hydro, North York, ON (Canada)

    1991-04-01

    The purpose of this paper is to provide a general review of certain key design areas which address the safety concerns of HT pump operation under conditions of significant void. To illustrate the challenges confronting designers and analysts, some of the highlights during the design of a protective system to prevent damage to HT piping and pump supports at Bruce NGS 'A' are outlined. The effects of this protective system on reactor safety are also discussed. HI pump operation under conditions of significant void offers a major challenge to designers and analysts to ensure that pump induced vibration and its effects on pump and piping are addressed. For an in-service station the search for a practical solution is often limited by existing. station equipment design and Layout. The diversity of design verification process requires a major commitment of engineering resources to ensure all. safety aspects meet the requirements of regulatory body. Work currently undertaken at Ontario Hydro Research Pump Test Complex on two-phase flow in pumps and piping may provide better prediction of vibration characteristics so that inherent conservativeness in fatigue Life prediction of HI system components can be reduced.

  12. Damage of natural stone tablets exposed to exhaust gas under laboratory conditions

    Science.gov (United States)

    Farkas, Orsolya; Szabados, György; Török, Ákos

    2016-04-01

    Natural stone tablets were exposed to exhaust gas under laboratory conditions to assess urban stone damage. Cylindrical test specimens (3 cm in diameter) were made from travertine, non-porous limestone, porous limestone, rhyolite tuff, sandstone, andesite, granite and marble. The samples were exposed to exhaust gas that was generated from diesel engine combustion (engine type: RÁBA D10 UTSLL 160, EURO II). The operating condition of the internal combustion engine was: 1300 r/m (app 50%). The exhaust gas was diverted into a pipe system where the samples were placed perpendicular to main flow for 1, 2, 4, 8 and 10 hours, respectively. The exhaust emission was measured by using AVL particulate measurement technology; filter paper method (AVL 415). The stone samples were documented and selective parameters were measured prior to and after exhaust gas exposure. Density, volume, ultrasonic pulse velocity, mineral composition and penetration depth of emission related particulate matter were recorded. The first results indicate that after 10 hours of exposure significant amount of particulate matter deposited on the stone surface independently from the surface properties and porosity. The black soot particles uniformly covered all types of stones, making hard to differentiate the specimens.

  13. Comprehensive study of biodiesel fuel for HSDI engines in conventional and low temperature combustion conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tormos, Bernardo; Novella, Ricardo; Garcia, Antonio; Gargar, Kevin [CMT-Motores Termicos, Universidad Politecnica de Valencia, Valencia, ES, Campus de Vera, s/n, Edificio 6D. Camino de Vera s/n, 46022 Valencia (Spain)

    2010-02-15

    In this research, an experimental investigation has been performed to give insight into the potential of biodiesel as an alternative fuel for High Speed Direct Injection (HSDI) diesel engines. The scope of this work has been broadened by comparing the combustion characteristics of diesel and biodiesel fuels in a wide range of engine loads and EGR conditions, including the high EGR rates expected for future diesel engines operating in the low temperature combustion (LTC) regime. The experimental work has been carried out in a single-cylinder engine running alternatively with diesel and biodiesel fuels. Conventional diesel fuel and neat biodiesel have been compared in terms of their combustion performance through a new methodology designed for isolating the actual effects of each fuel on diesel combustion, aside from their intrinsic differences in chemical composition. The analysis of the results has been sequentially divided into two progressive and complementary steps. Initially, the overall combustion performance of each fuel has been critically evaluated based on a set of parameters used as tracers of the combustion quality, such as the combustion duration or the indicated efficiency. With the knowledge obtained from this previous overview, the analysis focuses on the detailed influence of biodiesel on the different diesel combustion stages known ignition delay, premixed combustion and mixing controlled combustion, considering also the impact on CO and UHC (unburn-hydrocarbons) pollutant emissions. The results of this research explain why the biodiesel fuel accelerates the diesel combustion process in all engine loads and EGR rates, even in those corresponding with LTC conditions, increasing its possibilities as alternative fuel for future DI diesel engines. (author)

  14. Spent fuel stability under repository conditions - final report of the european project

    International Nuclear Information System (INIS)

    Poinssot, Ch.; Ferry, C.; Kelm, M.; Cavedon, J.M.; Corbel, C.; Jegou, Ch.; Lovera, P.; Miserque, F.; Poulesquen, A.; Grambow, B.; Andriambololona, Z.; Martinez-Esparza, A.; Kelm, M.; Loida, A.; Rondinella, V.; Wegen, D.; Spahiu, K.; Johnson, L.; Cachoir, Ch.; Lemmens, K.; Quinones, J.; Bruno, J.; Christensen, H.; Grambow, B.; Pablo, J. de

    2005-01-01

    This report is the final report of the European Project 'Spent Fuel Stability under Repository Conditions' (FIKW-CT-2001-00192 SFS) funded by the European Commission from Nov.2000 to Oct.2004. Gathering the work performed by 13 partners from 6 countries, it aims to specifically focus on the spent nuclear fuel long term alteration in deep repository and the subsequent radionuclides release rate as a function of time. This report synthesised the wide experimental work performed within this project and enlightens the major outcomes, which can be summarised as follow: - A new model for defining the Instant Release Fraction was developed in order to consider the potential fuel evolution before the water penetrates the canister. Quantitative assessment has been produced and shows a significant contribution to the long term dose; - Based on new experimental data, kinetic radiolytic scheme have been upgraded and are used to determine the amount of oxidants produced at the fuel/water interface; - The existence of a dose threshold below which the water radiolysis does not influence the fuel alteration has been demonstrated and occurs between 3.5 and 33 MBq.g UO21. Above the threshold, the fuel alteration rates is directly related to the dose rate. - Hydrogen was experimentally demonstrated to be an efficient oxidants scavenger preventing therefore the fuel oxidation. Molecular mechanism still need to be understood. - Finally, a new Matrix Alteration Model integrating most of the SFS results (apart of the hydrogen effect) has been developed and used to assess the fuel long tern stability in representative conditions of deep repository in salt, clay-rock and granite. The breadth of the results and the significance of the conclusions testify of the success of the collaboration within the project. (authors)

  15. Developing RCM Strategy for Hydrogen Fuel Cells Utilizing On Line E-Condition Monitoring

    International Nuclear Information System (INIS)

    Baglee, D; Knowles, M J

    2012-01-01

    Fuel cell vehicles are considered to be a viable solution to problems such as carbon emissions and fuel shortages for road transport. Proton Exchange Membrane (PEM) Fuel Cells are mainly used in this purpose because they can run at low temperatures and have a simple structure. Yet high maintenance costs and the inherent dangers of maintaining equipment using hydrogen are two main issues which need to be addressed. The development of appropriate and efficient strategies is currently lacking with regard to fuel cell maintenance. A Reliability Centered Maintenance (RCM) approach offers considerable benefit to the management of fuel cell maintenance since it includes an identification and consideration of the impact of critical components. Technological developments in e-maintenance systems, radio-frequency identification (RFID) and personal digital assistants (PDAs) have proven to satisfy the increasing demand for improved reliability, efficiency and safety. RFID technology is used to store and remotely retrieve electronic maintenance data in order to provide instant access to up-to-date, accurate and detailed information. The aim is to support fuel cell maintenance decisions by developing and applying a blend of leading-edge communications and sensor technology including RFID. The purpose of this paper is to review and present the state of the art in fuel cell condition monitoring and maintenance utilizing RCM and RFID technologies. Using an RCM analysis critical components and fault modes are identified. RFID tags are used to store the critical information, possible faults and their cause and effect. The relationship between causes, faults, symptoms and long term implications of fault conditions are summarized. Finally conclusions are drawn regarding suggested maintenance strategies and the optimal structure for an integrated, cost effective condition monitoring and maintenance management system.

  16. Developing RCM Strategy for Hydrogen Fuel Cells Utilizing On Line E-Condition Monitoring

    Science.gov (United States)

    Baglee, D.; Knowles, M. J.

    2012-05-01

    Fuel cell vehicles are considered to be a viable solution to problems such as carbon emissions and fuel shortages for road transport. Proton Exchange Membrane (PEM) Fuel Cells are mainly used in this purpose because they can run at low temperatures and have a simple structure. Yet high maintenance costs and the inherent dangers of maintaining equipment using hydrogen are two main issues which need to be addressed. The development of appropriate and efficient strategies is currently lacking with regard to fuel cell maintenance. A Reliability Centered Maintenance (RCM) approach offers considerable benefit to the management of fuel cell maintenance since it includes an identification and consideration of the impact of critical components. Technological developments in e-maintenance systems, radio-frequency identification (RFID) and personal digital assistants (PDAs) have proven to satisfy the increasing demand for improved reliability, efficiency and safety. RFID technology is used to store and remotely retrieve electronic maintenance data in order to provide instant access to up-to-date, accurate and detailed information. The aim is to support fuel cell maintenance decisions by developing and applying a blend of leading-edge communications and sensor technology including RFID. The purpose of this paper is to review and present the state of the art in fuel cell condition monitoring and maintenance utilizing RCM and RFID technologies. Using an RCM analysis critical components and fault modes are identified. RFID tags are used to store the critical information, possible faults and their cause and effect. The relationship between causes, faults, symptoms and long term implications of fault conditions are summarized. Finally conclusions are drawn regarding suggested maintenance strategies and the optimal structure for an integrated, cost effective condition monitoring and maintenance management system.

  17. Estimation of the core-wide fuel rod damage during a LWR LOCA

    International Nuclear Information System (INIS)

    Mattila, L.; Sairanen, R.; Stengaard, J.-O.

    1975-01-01

    The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

  18. Assessment of an accidental fuel radionuclide release data from the damaged Chernobyl NPP unit 4

    International Nuclear Information System (INIS)

    Mikhajlov, O.V.; Doroshenko, A.O.

    2015-01-01

    A procedure and results of assessment of fuel temperature dynamics during the formation of lava-like fuel containing materials (LFCM) in room 305/2 are presented. The assessment of the overheated fuel temperature carried out using mathematical type codes CORSOR's type from the known radionuclide release data in the period from 26.04 to 11.05.86. It is shown that the main LFCM's accumulations could be formed at a moderate value of temperatures than previously estimated. The obtained data were used to verify the ''blast furnace'' version of LFCM formation and formation of FCM with high uranium concentration and temperature of the core fragment's charge

  19. Global Combustion Mechanisms for Use in CFD Modeling under Oxy-Fuel Conditions

    DEFF Research Database (Denmark)

    Andersen, Jimmy; Rasmussen, Christian Lund; Giselsson, Trine

    2009-01-01

    Two global multistep schemes, the two-step mechanism of Westbrook and Dryer (WD) and the four-step mechanism of Jones and Lindstedt (JL), have been refined for oxy-fuel conditions. Reference calculations were conducted with a detailed chemical kinetic mechanism, validated for oxy-fuel combustion...... conditions. In the modification approach, the initiating reactions involving hydrocarbon and oxygen were retained, while modifying the H-2-CO-CO2 reactions in order to improve prediction of major species concentrations. The main attention has been to capture the trend and level of CO predicted...... by the detailed mechanism as well as the correct equilibrium concentration. A CFD analysis of a propane oxy-fuel flame has been performed using both the original and modified mechanisms. Compared to the original schemes, the modified WD mechanism improved the prediction of the temperature field and of CO...

  20. Criticality safety studies involved in actions to improve conditions for storing 'RA' research reactor spent fuel

    International Nuclear Information System (INIS)

    Matausek, M.; Marinkovic, N.

    1998-01-01

    A project has recently been initiated by the VINCA Institute of Nuclear Sciences to improve conditions in the spent fuel storage pool at the 6.5 MW research reactor RA, as well as to consider transferring this spent fuel into a new dry storage facility built for the purpose. Since quantity and contents of fissile material in the spent fuel storage at the RA reactor are such that possibility of criticality accident can not be a priori excluded, according to standards and regulations for handling fissile material outside a reactor, before any action is undertaken subcriticality should be proven under normal, as well as under credible abnormal conditions. To perform this task, comprehensive nuclear criticality safety studies had to be performed. (author)

  1. Infrared survey of 50 buildings constructed during 100 years: thermal performances and damage conditions

    Science.gov (United States)

    Ljungberg, Sven-Ake

    1995-03-01

    Different building constructions and craftsmanship give rise to different thermal performance and damage conditions. The building stock of most industrial countries consists of buildings of various age, and constructions, from old historic buildings with heavy stone or wooden construction, to new buildings with heavy or light concrete construction, or modern steel or wooden construction. In this paper the result from a detailed infrared survey of 50 buildings from six Swedish military camps is presented. The presentation is limited to a comparison of thermal performance and damage conditions of buildings of various ages, functions, and constructions, of a building period of more than 100 years. The result is expected to be relevant even to civilian buildings. Infrared surveys were performed during 1992-1993, with airborne, and mobile short- and longwave infrared systems, out- and indoor thermography. Interpretation and analysis of infrared data was performed with interactive image and analyzing systems. Field inspections were carried out with fiber optics system, and by ocular inspections. Air-exchange rate was measured in order to quantify air leakages through the building envelope, indicated in thermograms. The objects studied were single-family houses, barracks, office-, service-, school- and exercise buildings, military hotels and restaurants, aircraft hangars, and ship factory buildings. The main conclusions from this study are that most buildings from 1880 - 1940 have a solid construction with a high quality of craftsmanship, relatively good thermal performance, due to extremely thick walls, and adding insulation at the attic floor. From about 1940 - 1960 the quality of construction, thermal performance and craftsmanship seem to vary a lot. Buildings constructed during the period of 1960 - 1990 have in general the best thermal performance due to a better insulation capacity, however, also one finds here the greatest variety of problems. The result from this

  2. Combustion and emissions characteristics of diesel engine fueled by biodiesel at partial load conditions

    International Nuclear Information System (INIS)

    An, H.; Yang, W.M.; Chou, S.K.; Chua, K.J.

    2012-01-01

    Highlights: ► Impact of engine load on engine’s performance, combustion and emission characteristics. ► The brake specific fuel consumption (BSFC) increases significantly at partial load conditions. ► The brake thermal efficiency (BTE) drops at lower engine loads, and increases at higher loads. ► The partial load also influences the trend of CO emissions. -- Abstract: This paper investigated the performance, combustion and emission characteristics of diesel engine fueled by biodiesel at partial load conditions. Experiments were conducted on a common-rail fuel injection diesel engine using ultra low sulfur diesel, biodiesel (B100) and their blend fuels of 10%, 20%, 50% (denoted as B10, B20 and B50 respectively) under various loads. The results show that biodiesel/blend fuels have significant impacts on the engine’s brake specific fuel consumption (BSFC) and brake thermal efficiency (BTE) at partial load conditions. The increase in BSFC for B100 is faster than that of pure diesel with the decrease of engine load. A largest increase of 28.1% in BSFC is found at 10% load. Whereas for BTE, the results show that the use of biodiesel results in a reduced thermal efficiency at lower engine loads and improved thermal efficiency at higher engine loads. Furthermore, the characteristics of carbon monoxide (CO) emissions are also changed at partial load conditions. When running at lower engine loads, the CO emission increases with the increase of biodiesel blend ratio and the decrease of engine speed. However, at higher engine loads, an opposite trend is obtained.

  3. Feasibility Study for Monitoring Actinide Elements in Process Materials Using FO-LIBS at Advanced spent fuel Conditioning Process Facility

    Energy Technology Data Exchange (ETDEWEB)

    Han, Bo-Young; Choi, Daewoong; Park, Se Hwan; Kim, Ho-Dong [Nonproliferation System Research Division, Korea Atomic Energy Research Institute, Daejeon, 305-353 (Korea, Republic of); Dae, Dongsun [Department of Chemistry, Mokpo National University, Jeonnam 534-729 (Korea, Republic of); Whitehouse, Andrew I. [Applied Photonics Ltd., Unit 8 Carleton Business Park, Skipton, North Yorkshire BD23 2DE (United Kingdom)

    2015-07-01

    Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in the target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)

  4. Feasibility Study for Monitoring Actinide Elements in Process Materials Using FO-LIBS at Advanced spent fuel Conditioning Process Facility

    International Nuclear Information System (INIS)

    Han, Bo-Young; Choi, Daewoong; Park, Se Hwan; Kim, Ho-Dong; Dae, Dongsun; Whitehouse, Andrew I.

    2015-01-01

    Korea Atomic Energy Research Institute (KAERI) have been developing the design and deployment methodology of Laser- Induced Breakdown Spectroscopy (LIBS) instrument for safeguards application within the argon hot cell environment at Advanced spent fuel Conditioning Process Facility (ACPF), where ACPF is a facility being refurbished for the laboratory-scaled demonstration of advanced spent fuel conditioning process. LIBS is an analysis technology used to measure the emission spectra of excited elements in the local plasma of a target material induced by a laser. The spectra measured by LIBS are analyzed to verify the quality and quantity of the specific element in the target matrix. Recently LIBS has been recognized as a promising technology for safeguards purposes in terms of several advantages including a simple sample preparation and in-situ analysis capability. In particular, a feasibility study of LIBS to remotely monitor the nuclear material in a high radiation environment has been carried out for supporting the IAEA safeguards implementation. Fiber-Optic LIBS (FO-LIBS) deployment was proposed by Applied Photonics Ltd because the use of fiber optics had benefited applications of LIBS by delivering the laser energy to the target and by collecting the plasma light. The design of FO-LIBS instrument for the measurement of actinides in the spent fuel and high temperature molten salt at ACPF had been developed in cooperation with Applied Photonics Ltd. FO-LIBS has some advantages as followings: the detectable plasma light wavelength range is not limited by the optical properties of the thick lead-glass shield window and the potential risk of laser damage to the lead-glass shield window is not considered. The remote LIBS instrument had been installed at ACPF and then the feasibility study for monitoring actinide elements such as uranium, plutonium, and curium in process materials has been carried out. (authors)

  5. Heat conduction in a plate-type fuel element with time-dependent boundary conditions

    International Nuclear Information System (INIS)

    Faya, A.J.G.; Maiorino, J.R.

    1981-01-01

    A method for the solution of boundary-value problems with variable boundary conditions is applied to solve a heat conduction problem in a plate-type fuel element with time dependent film coefficient. The numerical results show the feasibility of the method in the solution of this class of problems. (Author) [pt

  6. Thermal behaviour of high burnup PWR fuel under different fill gas conditions

    International Nuclear Information System (INIS)

    Tverberg, T.

    2001-01-01

    During its more than 40 years of existence, a large number of experiments have been carried out at the Halden Reactor Project focusing on different aspects related to nuclear reactor fuel. During recent years, the fuels testing program has mainly been focusing on aspects related to high burnup, in particular in terms of fuel thermal performance and fission gas release, and often involving reinstrumentation of commercially irradiated fuel. The paper describes such an experiment where a PWR rod, previously irradiated in a commercial reactor to a burnup of ∼50 MWd/kgUO 2 , was reinstrumented with a fuel central oxide thermocouple and a cladding extensometer together with a high pressure gas flow line, allowing for different fill gas compositions and pressures to be applied. The paper focuses on the thermal behaviour of such LWR rods with emphasis on how different fill gas conditions influence the fuel temperatures and gap conductance. Rod growth rate was also monitored during the irradiation in the Halden reactor. (author)

  7. Dynamic characteristics of hydrocarbon fuel within the channel at supercritical and pyrolysis condition

    Science.gov (United States)

    Yu, Bin; Zhou, Weixing; Qin, Jiang; Bao, Wen

    2017-12-01

    Regenerative cooling with fuel as the coolant is used in the scramjet engine. In order to grasp the dynamic characteristics of engine fuel supply processes, this article studies the dynamic characteristics of hydrocarbon fuel within the channel. A one-dimensional dynamic model was proved, the thermal energy storage effect, fuel volume effect and chemical dynamic effect have been considered in the model, the ordinary differential equations were solved using a 4th order Runge-Kutta method. The precision of the model was validated by three groups of experimental data. The effects of input signal, working condition, tube size on the dynamic characteristics of pressure, flow rate, temperature have been simulated. It is found that cracking reaction increased the compressibility of the fuel pyrolysis mixture and lead to longer responding time of outlet flow. The responding time of outlet flow can reach 3s when tube is 5m long which will greatly influence the control performance of the engine thrust system. Meanwhile, when the inlet flow rate appears the step change, the inlet pressure leads to overshoot, the overshoot can reach as much as 100%, such highly transient impulse will result in detrimental effect on fuel pump.

  8. Inert materials for the GFR fuel. Characterizations, chemical interactions and irradiation damage

    International Nuclear Information System (INIS)

    Audubert, Fabienne; Carlot, Gaoelle; Lechelle, Jacques; David, Laurent; Gomes, Severine

    2005-01-01

    In the framework of an extensive R and D Program on GFR fuel, studies on inert materials have been performed at the French Atomic Energy Commission (CEA). The inert materials would be associated with the fuel with the aim of featuring an efficient barrier to radiotoxic species with regard to the cooling circuit of the reactor. Potential matrices identified for dispersion fuels or particles fuels are SiC, TiN, ZrN, ZrC, TiC. Physical microstructural and thermal properties have been determined in order to evaluate elaboration process effects. The evolution under irradiation of thermal properties (such as conductivity, diffusivity) of the materials has been studied using heavy ions to simulate fission product irradiation. After irradiation, scanning thermal microscopy is used to investigate the thermal degradation of the materials. Thermal conductivity variations were obtained on TiC irradiated with krypton ion at an energy of 86 MeV and a fluence of 5.10 15 ions.cm -2 . They are quantified at 19 W.m -1 .K -1 . On other materials such as SiC, ZrC, TiN, no thermal conductivity contrast was shown. Reactivity between the inert matrix (SiC or TiN) and the fuel (U, Pu)N have been evaluated on powders and on ceramic samples in contact by a thermal treatment under several atmospheres. It was shown that SiC reacts with (U, Pu)N in various atmospheres making secondary phases as PuSi 2 , USi 2 , U 20 Si 16 C 3 . TiN behaviour seems to be better: the only reactivity which may take place would be a variation of the nitrogen stoichiometry in TiN and (U, Pu)N at the interface. (author)

  9. Response of unirradiated and irradiated PWR fuel rods tested under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Quapp, W.J.; Martinson, Z.R.; McCardell, R.K.; Mehner, A.S.

    1978-01-01

    This report summarizes the results from the single-rod power-cooling-mismatch (PCM) and irradiation effects (IE) tests conducted to date in the Power Burst Facility (PBF) at the U.S. DOE Idaho National Engineering Laboratory. This work was performed for the U.S. NRC under contact to the Department of Energy. These tests are part of the NRC Fuel Behavior Program, which is designed to provide data for the development and verification of analytical fuel behavior models that are used to predict fuel response to abnormal or postulated accident conditions in commercial LWRs. The mechanical, chemical and thermal response of both previously unirradiated and previously irradiated LWR-type fuel rods tested under power-cooling-mismatch condition is discussed. A brief description of the test designs is presented. The results of the PCM thermal-hydraulic studies are summarized. Primary emphasis is placed on the behavior of the fuel and cladding during and after stable film boiling. (orig.) [de

  10. Visual investigation of transient fuel behavior under a rapid heating condition

    International Nuclear Information System (INIS)

    Saito, Shinzo

    1981-10-01

    An in-reactor experimental research on fuel behavior under reactivity initiated accident (RIA) conditions is being conducted in the Nuclear Safety Research Reactor (NSRR). The optical system in which a non-browning lens periscope is directly installed in the test section was successfully developed for photographing transient fuel behavior. Several phenomena which had never been revealed before were observed in the slow motion pictures taken in the NSRR experiments which were performed in the water and air environments. As for incipient failure mechanism for an unirradiated fuel rod under RIA conditions, brittle fracture of the cladding during quenching is dominant. However, a split cracking possibly occurs during even red hot state of the cladding. It is considered that the crack is generated by the local internal pressure increase at the specified region blocked up due to the melting of the cladding inner surface. The film boiling is unexpectablly violent specially in the early stage of the transient, and film thickness becomes 5 -- 6 mm at maximum. The observed thick vapor film can not be explained by the conventional theory, but the effect of hydrogen which is produced by Zircaloy-water reaction is reasonably explained to form thick film in the report. The molten fuel was expelled from the cladding in the experiment which was performed in an air environment. The expelled fuel fragmented due to possibly initial motion effect, not mechanical collision effect, because Weber number is smaller than the critical value. (author)

  11. Evaluation of gap heat transfer model in ELESTRES for CANDU fuel element under normal operating conditions

    International Nuclear Information System (INIS)

    Lee, Kang Moon; Ohn, Myung Ryong; Im, Hong Sik; Choi, Jong Hoh; Hwang, Soon Taek

    1995-01-01

    The gap conductance between the fuel and the sheath depends strongly on the gap width and has a significant influence on the amount of initial stored energy. The modified Ross and Stoute gap conductance model in ELESTRES is based on a simplified thermal deformation model for steady-state fuel temperature calculations. A review on a series of experiments reveals that fuel pellets crack, relocate, and are eccentrically positioned within the sheath rather than solid concentric cylinders. In this paper, the two recently-proposed gap conductance models (offset gap model and relocated gap model) are described and are applied to calculate the fuel-sheath gap conductances under experimental conditions and normal operating conditions in CANDU reactors. The good agreement between the experimentally-inferred and calculated gap conductance values demonstrates that the modified Ross and Stoute model was implemented correctly in ELESTRES. The predictions of the modified Ross and Stoute model provide conservative values for gap heat transfer and fuel surface temperature compared to the offset gap and relocated gap models for a limiting power envelope. 13 figs., 3 tabs., 16 refs. (Author)

  12. Safeguardability assessment on pilot-scale advanced spent fuel conditioning facility

    International Nuclear Information System (INIS)

    Lee, S.Y.; Li, T.K.; Pickett, S.E.; Miller, M.C.; Ko, W.I.; Kim, H.D.

    2006-01-01

    Full text: In South Korea, approximately 6,000 metric tons of spent nuclear fuel from commercial reactor operation has been accumulated with the expectation of more than 30,000 metric tons, three times the present storage capacity, by the end of 2040. To resolve these challenges in spent fuel management, the Korea Atomic Energy Research Institute (KAERI) has been developing a dry reprocessing technology called Advanced Spent Fuel Conditioning Process (ACP). This is an electrometallurgical treatment technique to convert oxide-type spent fuel into a metallic form, and the electrolytic reduction (ER) technology developed recently is known as a more efficient concept for spent fuel conditioning. The goal of the ACP study is to recover more than 99% of the actinide elements into a metallic form with minimizing the volume and heat load of spent fuel. The significant reduction of the volume and heat load of spent fuel is expected to lighten the burden of final disposal in terms of disposal size, safety, and economics. In the framework of R and D collaboration for the ACP safeguards, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and KAERI. The purpose of this study is to address the safeguardability of the ACP technology, through analysis of material flow and development of a proper safeguards system that meet IAEA's comprehensive safeguards objective. The sub-processes and material flow of the pilot-scale ACP facility were analyzed, and subsequently the relevant material balance area (MBA) and key measurement point (KMP) were designed for material accounting. The uncertainties in material accounting were also estimated with international target values, and design requirements for the material accounting systems were derived

  13. 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

  14. 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

  15. THEORETICAL INVESTIGATION OF MICROSTRUCTURE EVOLUTION AND DEFORMATION OF ZIRCONIUM UNDER CASCADE DAMAGE CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Barashev, Alexander V [ORNL; Golubov, Stanislav I [ORNL; Stoller, Roger E [ORNL

    2012-06-01

    This work is based on our reaction-diffusion model of radiation growth of Zr-based materials proposed recently in [1]. In [1], the equations for the strain rates in unloaded pure crystal under cascade damage conditions of, e.g., neutron or heavy-ion irradiation were derived as functions of dislocation densities, which include contributions from dislocation loops, and spatial distribution of their Burgers vectors. The model takes into account the intra-cascade clustering of self-interstitial atoms and their one-dimensional diffusion; explains the growth stages, including the break-away growth of pre-annealed samples; and accounts for some striking observations, such as of negative strain in prismatic direction, and co-existence of vacancy- and interstitial-type prismatic loops. In this report, the change of dislocation densities due to accumulation of sessile dislocation loops is taken into account explicitly to investigate the dose dependence of radiation growth. The dose dependence of climb rates of dislocations is calculated, which is important for the climb-induced glide model of radiation creep. The results of fitting the model to available experimental data and some numerical calculations of the strain behavior of Zr for different initial dislocation structures are presented and discussed. The computer code RIMD-ZR.V1 (Radiation Induced Microstructure and Deformation of Zr) developed is described and attached to this report.

  16. Standard test method for damage to contacting solid surfaces under fretting conditions

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This test method covers the studying or ranking the susceptibility of candidate materials to fretting corrosion or fretting wear for the purposes of material selection for applications where fretting corrosion or fretting wear can limit serviceability. 1.2 This test method uses a tribological bench test apparatus with a mechanism or device that will produce the necessary relative motion between a contacting hemispherical rider and a flat counterface. The rider is pressed against the flat counterface with a loading mass. The test method is intended for use in room temperature air, but future editions could include fretting in the presence of lubricants or other environments. 1.3 The purpose of this test method is to rub two solid surfaces together under controlled fretting conditions and to quantify the damage to both surfaces in units of volume loss for the test method. 1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.5...

  17. 40 CFR 80.530 - Under what conditions can 500 ppm motor vehicle diesel fuel be produced or imported after May 31...

    Science.gov (United States)

    2010-07-01

    ... motor vehicle diesel fuel be produced or imported after May 31, 2006? 80.530 Section 80.530 Protection... FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel Temporary Compliance Option § 80.530 Under what conditions can 500 ppm motor vehicle diesel...

  18. Optimization of combustion chamber geometry and operating conditions for compression ignition engine fueled with pre-blended gasoline-diesel fuel

    International Nuclear Information System (INIS)

    Lee, Seokhwon; Jeon, Joonho; Park, Sungwook

    2016-01-01

    Highlights: • Pre-blended gasoline-diesel fuel was used with direct injection system. • KIVA-CHEMKIN code modeled dual-fuel fuel spray and combustion processes with discrete multi-component model. • The characteristics of Combustion and emission on pre-blended fuel was investigated with various fuel reactivities. • Optimization of combustion chamber shape improved combustion performance of the gasoline-diesel blended fuel engine. - Abstract: In this study, experiments and numerical simulations were used to improve the fuel efficiency of compression ignition engine using a gasoline-diesel blended fuel and an optimization technology. The blended fuel is directly injected into the cylinder with various blending ratios. Combustion and emission characteristics were investigated to explore the effects of gasoline ratio on fuel blend. The present study showed that the advantages of gasoline-diesel blended fuel, high thermal efficiency and low emission, were maximized using the numerical optimization method. The ignition delay and maximum pressure rise rate increased with the proportion of gasoline. As the gasoline fraction increased, the combustion duration and the indicated mean effective pressure decreased. The homogeneity of the fuel-air mixture was improved due to longer ignition delay. Soot emission was significantly reduced up to 90% compared to that of conventional diesel. The nitrogen oxides emissions of the blended fuel increased slightly when the start of injection was retarded toward top dead center. For the numerical study, KIVA-CHEMKIN multi-dimensional CFD code was used to model the combustion and emission characteristics of gasoline-diesel blended fuel. The micro genetic algorithm coupled with the KIVA-CHEMKIN code were used to optimize the combustion chamber shape and operating conditions to improve the combustion performance of the blended fuel engine. The optimized chamber geometry enhanced the fuel efficiency, for a level of nitrogen oxides

  19. Some conditions and prospects of transition to closed fuel cycle in Russia

    International Nuclear Information System (INIS)

    Lependin, A.V.; Oussanov, V.I.; Lependina, E.V.; Ioughai, S.V.

    2001-01-01

    Nuclear policy of Russia is based on the necessity of closure of nuclear fuel cycle. But at the same time schedule of such a going is not defined. In this study some conditions and possible time-frames of going the nuclear fuel cycle of Russia to closure are discussed. Naturally, the main condition is revival of Russian economy wherein nuclear power will turn to be necessary in a number of Russian regions. But the question is whether closure of nuclear cycle strategy will be implemented in the near future or nuclear power will develop based on open fuel cycle over a long period of time? at present economic circumstances in Russia has formed in such a way that economics of current projects is not favourable to going to closure of cycle due to high capital investment cost and low fuel component of costs, due to low cost of natural uranium. Ecological analysis performed within the framework of external cost model also does not suggest that closed cycle has essential advantages at present, but also in sight. The authors have considered a model including not only external costs but also total resources expenditures with long-term power development. In the framework of such a method it can be demonstrated that closed fuel cycle has some important advantages taking into account not only tasks of immediate future, but power development strategy for the period of 30-50 years. Under conditions of nuclear capacities increase (to 30-50 GW) limitation of cheap uranium resources available in Russia will assume a new significance. Approach of prices at the back-end stages of nuclear fuel cycle to West Europe level also will favour to going to a closed fuel cycle. More severe ecological requirements answering to a sustainable development concept also will make a contribution. Closure of fuel cycle can be significantly accelerated in the case of implementation of weapon plutonium utilization program. The factors mentioned above facilitate evenly to going to a closed nuclear fuel

  20. 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.

  1. 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)

  2. Conceptual structure design of experimental facility for advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    Joo, J. S.; Koo, J. H.; Jung, W. M.; Jo, I. J.; Kook, D. H.; Yoo, K. S.

    2003-01-01

    A study on the advanced spent fuel conditioning process (ACP) is carring out for the effective management of spent fuels of domestic nuclear power plants. This study presents basic shielding design, modification of IMEF's reserve hot cell facility which reserved for future usage, conceptual and structural architecture design of ACP hot cell and its contents, etc. considering the characteristics of ACP. The results of this study will be used for the basic and detail design of ACP demonstration facility, and utilized as basic data for the safety evaluation as essential data for the licensing of the ACP facility

  3. The release of cesium and the actinides from spent fuel under unsaturated conditions

    International Nuclear Information System (INIS)

    Finn, P.A.; Hoh, J.C.; Wolf, S.F.; Slater, S.A.; Bates, J.K.

    1995-01-01

    Tests designed to be similar to the unsaturated and oxidizing conditions expected in the candidate repository at Yucca Mountain are in progress with spent fuel at 90 degree C. The similarities and the differences in release behavior for 137 Cs during the first 2.6 years and the actinides during the first 1.6 years of testing are presented for tests done with (1) water dripped on the fuel at a rate of 0.075 and 0.75 mL every 3.5 days and (2) in a saturated water vapor environment

  4. Comparison of Damage Models for Predicting the Non-Linear Response of Laminates Under Matrix Dominated Loading Conditions

    Science.gov (United States)

    Schuecker, Clara; Davila, Carlos G.; Rose, Cheryl A.

    2010-01-01

    Five models for matrix damage in fiber reinforced laminates are evaluated for matrix-dominated loading conditions under plane stress and are compared both qualitatively and quantitatively. The emphasis of this study is on a comparison of the response of embedded plies subjected to a homogeneous stress state. Three of the models are specifically designed for modeling the non-linear response due to distributed matrix cracking under homogeneous loading, and also account for non-linear (shear) behavior prior to the onset of cracking. The remaining two models are localized damage models intended for predicting local failure at stress concentrations. The modeling approaches of distributed vs. localized cracking as well as the different formulations of damage initiation and damage progression are compared and discussed.

  5. Current activities on improving storage conditions of the research reactor RA spent fuel - Part II

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kopecni, M.; Vukadin, Z.; Plecas, I.; Pavlovic, R.; Sotic, O.; Marinkovic, N.

    1998-01-01

    To minimize further corrosion and preserve integrity of aluminum barrels and the stainless steel channel-type containers that were found to contain leaking spent fuel, actions to improve conditions in the existing spent fuel storage pool at the RA research reactor were initiated. Technology was elaborated and equipment was produced and applied for removal of sludge and other debris from the bottom of the pool, filtration of the pool water, sludge conditioning in cement matrix and disposal at the low and medium waste repository at VINCA site. More sophisticated operations are to be performed together with foreign experts. Safety measures and precautions were determined. Subcriticality was proved under normal and/or possible abnormal conditions. (author)

  6. The influence of fast reactor emergency conditions upon fuel element performance

    International Nuclear Information System (INIS)

    Bagdasarov, Yu.E.; Buksha, Yu.K.; Zabudko, L.M.; Likhachev, Yu.I.

    1985-01-01

    Fuel-pin cladding is one of the most important protective barriers preventing the release and propagation of radioactive contamination. By now the calculated determination of fast-reactor fuel-element performance under stationary conditions has been considered in detail but the investigation of the influence of emergency conditions has been given less attention. Under emergency conditions of the fast reactor operation there arise short-duration excesses of rated parameters (temperature, energy release, etc.) which are confined within tolerable limits with the use of the safety system. Some features of the sodium-cooled fast reactors (small mean prompt-neutron lifetime, relatively weak reactivity feedback, etc.) complicate the work of safety systems. Therefore, the tolerable deviations of parameters should be carefully validated

  7. Calculation of the fuel temperature field under heat release and heat conductance transient conditions

    International Nuclear Information System (INIS)

    Kazakov, E.K.; Chernukhina, G.M.

    1974-01-01

    Results of calculation of the temperature distribution in an annular fuel element at transient thermal conductivity and heat release values are given. The calculation has been carried out by the mesh technique with the third-order boundary conditions for the inner surface assumed and with heat fluxes and temperatures at the zone boundaries to be equal. Three variants of solving the problem of a stationary temperature field are considered for failed fuel elements with clad flaking or cracks. The results obtained show the nonuniformity of the fuel element temperature field to depend strongly on the perturbation parameter at transient thermal conductivity and heat release values. In case of can flaking at a short length, the core temperature rises quickly after flaking. While evaluating superheating, one should take into account the symmetry of can flaking [ru

  8. Recent findings on the oxidation of UO2 fuel under nominally dry storage conditions

    International Nuclear Information System (INIS)

    Taylor, P.; McEachern, R.J.; Sunder, S.; Wasywich, K.M.; Miller, N.H.; Wood, D.D.

    1995-01-01

    This paper is an overview of fuel-storage demonstration experiments, supporting research on UO 2 oxidation, and associated model development, in progress at AECL's Whiteshell Laboratories. The work is being performed to determine the time/temperature limits for safe storage of irradiated CANDU fuel in dry air. The most significant recent experimental finding has been the detection of small quantities of U 3 O 8 , formed over periods of one to several years in a variety of experiments at 150-170 deg C. Another important trading is the slight suppression of U 3 O 8 formation in SIMFUEL and other doped U0 2 formulations. The development of a nucleation-and-growth model for U 3 O 8 formation is discussed, along with available activation energy data. These provide a basis for predicting U 3 O 8 formation rates under dry-storage conditions, and hence optimizing fuel storage strategies. (author)

  9. An experiment to examine the mechanistic behaviour of irradiated CANDU fuel stored under dry conditions

    International Nuclear Information System (INIS)

    Oldaker, I.E.; Crosthwaite, J.L.; Keltie, R.J.; Truss, K.J.

    1979-01-01

    A program has begun to use the Whiteshell Nuclear Research Establishment dry-storage canisters to store some selected CANDU irradiated fuel bundles in an 'easily retrievable basket.' The object of the experimental program is to study the long-term stability of the Zircaloy-sheathed UO 2 and UC fuel elements when stored in air. Bundles were loaded into a canister in October 1979 following detailed examination and removal of up to three complete elements from most bundles. These elements are currently being subjected to detailed destructive examinations, including metallography and scanning electron micrography, to fully characterize their pre-storage condition. After four years, and every five years thereafter, further elements will be examined similarly to study the effects of the storage environment on the stability of the Zircaloy sheathing, and on its continued ability to contain the fuel safely in an interim storage facility. (author)

  10. The ''THERMOST'' for analysing thermo-structural behaviour of LWR fuel rod under PCI conditions

    International Nuclear Information System (INIS)

    Nuno, H.; Ogawa, S.; Kobayashi, H.

    1983-01-01

    As one of the methods for evaluating the fuel rod performances under power ramping or load following operations, the combined ''FROST'' and ''THERMOST'' system has been developed and being brought into practical use. The former had already been presented at Blackpool Meeting in 1978, and the latter is going to be presented in this paper. The major purpose of the THERMOST is to analyse very detailed thermal and structural fuel behaviours in a rather localized part of fuel rod whereas the FROST deals with whole-rod-wide general performances. The code handles 2-dimensional thermal and structural analyses simultaneously by using finite element method, in axial section wide or in lateral section wide. It consists of a fundamental FEM system of generalized constitution and its surrounding subroutine system which characterizes fuel behaviours such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer elements (6 kinds) and structural analysis by axisymmetric ring and lateral plane elements (6 kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping condition is presented with some inpile test data. (author)

  11. 'THERMOST' for analysing thermo-structural behaviour of LWR fuel rods under PCI conditions

    International Nuclear Information System (INIS)

    Nuno, H.; Ogawa, S.; Kobayashi, H.

    1983-01-01

    As a method for evaluating fuel rod performance under power ramping or load following operations, the combined FROST/ THERMOST system has been developed and brought into practical use. FROST was presented at the IAEA Blackpool Meeting in 1978, and THERMOST is the subject of this paper. The major purpose of THERMOST is to analyse very detailed thermal and structural fuel behaviour in a rather localised part of the fuel rod whereas FROST deals with whole rod general performance. The code handles two-dimensional thermal and structural analyses simultaneously by using a finite element method, in axial section or in lateral section. It consists of a fundamental FEM system of generalised constitution, and a surrounding subroutine system which characterises fuel behaviour, such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer element (six kinds), and structural analysis by axisymmetric ring and lateral plane element (six kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping conditions is presented with some in-pile test data. (author)

  12. The prediction of the-circumferential fuel-temperature distribution under ballonian condition. Vol. 3

    Energy Technology Data Exchange (ETDEWEB)

    Abdallah, A M; El-Sherbiny, E M [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    Swelling and thermal distortion of nuclear fuel elements due to depressurization of reactor coolant may cause contracts in points or finite regions between adjacent fuel elements in square and triangle lattices. This is very probable in Advanced Pressurized Water Reactors where the clearance between fuel elements is about 1 mm. This results in partial blocking of the coolant flow and formation of hot spots in the contact regions. In these regions, absence of coolant results in nonuniform clad circumferential temperature distribution. This causes excessive thermal stresses which may produce local melting or clad failure. An accurate prediction of the clad circumferential temperature distribution during these severe incidents is very important. This problem was studied numerically during transient and steady state conditions. Recently, a semi analytical solution for the underlying problem was derived assuming the heat transfer coefficient to vary linearly with the circumferential distance measured from the cusp point, and the heat flux at the fuel-clad interface to be a constant quantity. In the present work, an approximate analytic solution is obtained. The accuracy is tested by solving the problem numerically. Also the problem is reanalyzed by considering the heat flux at the fuel-clad interface to be a power function of the angular distance along the clad surface. Moreover, the heat transfer coefficient is assumed to be a function of both the circumferential coordinate and temperature of the clad. Discussion of the analytical solution and the assumptions are rationalized in the text. 4 figs.

  13. A kinetic model for the stability of spent fuel matrix under oxic conditions

    International Nuclear Information System (INIS)

    Bruno, J.; Cera, E.; Duro, L.; Eriksen, T.E.

    1996-01-01

    A kinetic model for the UO 2 -spent fuel dissolution has been developed by integrating all the fundamental and experimental evidence about the redox buffer capacity of the UO 2 matrix itself within the methodological framework of heterogeneous redox reactions and dissolution kinetics. The purpose of the model is to define the geochemical stability of the spent fuel matrix and its resistance to internal and external disturbances. The model has been built in basis the reductive capacity (RDC) of the spent fuel/water system. A sensitivity analysis has been performed in order to identify the main parameters that affect the RDC of the system, the oxidant consumption and the radionuclide release. The number of surface co-ordination sites, the surface area to volume ratio, the kinetics of oxidants generation by radiolysis and the kinetics of oxidative dissolution of UO 2 , have been found to be the main parameters that can affect the reductive capacity of the spent fuel matrix. The model has been checked against some selected UO 2 and spent fuel dissolution data, performed under oxidizing conditions. The results are quite encouraging. (orig.)

  14. Examination of the damage and failure response of tantalum and copper under varied shock loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bronkhorst, Curt A [Los Alamos National Laboratory; Dennis - Koller, Darcie [Los Alamos National Laboratory; Cerreta, Ellen K [Los Alamos National Laboratory; Gray Ill, George T [Los Alamos National Laboratory; Bourne, Neil [AWE-ALDERMASTON

    2010-12-16

    A number of plate impact experiments have been conducted on high purity polycrystalline tantalum and copper samples using graded flyer plate configurations to alter the loading profile. These experiments are designed in a way so that a broad range of damage regimes are probed. The results show that the nucleation of damage primarily occurs at the grain boundaries of the materials. This affords us the opportunity to propose a porosity damage nucleation criterion which begins to account for the length scales of the microstructure (grain size distribution) and the mechanical response of the grain boundary regions (failure stress distribution). This is done in the context of a G-T-N type model for the ductile damage and failure response of both the materials examined. The role of micro-inertial effects on the porosity growth process is also considered.

  15. Transient response of a polymer electrolyte membrane fuel cell subjected to time-varying modulating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Noorani, S.; Shamim, T. [Michigan-Dearborn Univ., Dearborn, MI (United States). Dept. of Mechanical Engineering

    2009-07-01

    In order for fuel cells to compete with internal combustion engines, they must have significant advantages in terms of overall efficiency, weight, packaging, safety and cost. A key requirement is its ability to operate under highly transient conditions during start-up, acceleration, and deceleration with stable performance. Therefore, a better understanding of fuel cell dynamic behaviour is needed along with better water management and distributions inside the cell. Therefore, this study investigated the effect of transient conditions on water distribution inside a polymer electrolyte membrane (PEM) cell. A macroscopic single-fuel cell based, one-dimensional, isothermal mathematical model was used to study the effect of modulating cell voltage on the water distribution of anode, cathode, catalyst layers, and membrane. Compared to other existing models, this model did not rely on the non-physical assumption of the uptake curve equilibrium between the pore vapour and ionomer water in the catalyst layers. Instead, the transition between the two phases was modeled as a finite-rate equilibration process. The modulating conditions were simulated by forcing the temporal variations in fuel cell voltage. The results revealed that cell voltage modulations cause a departure in the cell behaviour from its steady behaviour, and the finite-rate equilibration between the catalyst vapour and liquid water can be a factor in determining the cell response. The cell response is also affected by the modulating frequency and amplitude. The peak cell response was observed at low frequencies. Keywords: fuel cell, water transport, dynamic behaviour, numerical simulations. 9 refs., 1 tab., 5 figs.

  16. Thermal analysis on NAC-STC spent fuel transport cask under different transport conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Yumei [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Yang, Jian, E-mail: zdhjkz@zju.edu.cn [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Xu, Chao; Wang, Weiping [Institute of Process Equipment, Zhejiang University, Hangzhou (China); Ma, Zhijun [Department of Material Engineering, South China University of Technology, Guangzhou (China)

    2013-12-15

    Highlights: • Spent fuel cask was investigated as a whole instead of fuel assembly alone. • The cask was successfully modeled and meshed after several simplifications. • Equivalence method was used to calculate the properties of parts. • Both the integral thermal field and peak values are captured to verify safety. • The temperature variations of key parts were also plotted. - Abstract: Transport casks used for conveying spent nuclear fuel are inseparably related to the safety of the whole reprocessing system for spent nuclear fuel. Thus they must be designed according to rigorous safety standards including thermal analysis. In this paper, for NAC-STC cask, a finite element model is established based on some proper simplifications on configurations and the heat transfer mechanisms. Considering the complex components and gaps, the equivalence method is presented to define their material properties. Then an equivalent convection coefficient is introduced to define boundary conditions. Finally, the temperature field is captured and analyzed under both normal and accident transport conditions by using ANSYS software. The validity of numerical calculation is given by comparing its results with theoretical calculation. Obtaining the integral distribution laws of temperature and peak temperature values of all vital components, the security of the cask can be evaluated and verified.

  17. Fuel temperature influence on diesel sprays in inert and reacting conditions

    International Nuclear Information System (INIS)

    Payri, Raul; García-Oliver, Jose M.; Bardi, Michele; Manin, Julien

    2012-01-01

    The detailed knowledge of the evaporation–combustion process of the Diesel spray is a key factor for the development of robust injection strategies able to reduce the pollutant emissions and keep or increase the combustion efficiency. In this work several typical measurement applied to the diesel spray diagnostic (liquid length, lift-off length and ignition delay) have been employed in a novel continuous flow test chamber that allows an accurate control on a wide range of thermodynamic test conditions (up to 1000 K and 15 MPa). A step forward in the control of the test boundary conditions has been done employing a special system to study the fuel temperature effect on the evaporation and combustion of the spray. The temperature of the injector body has been controlled with a thermostatic system and the relationship between injector body and fuel temperature has been observed experimentally. Imaging diagnostics have been employed to visualize the liquid phase penetration in evaporative/inert conditions and, lift-off length and ignition delay in reactive condition. The results underline a clear influence of the injector body temperature on both conditions, evaporative and, in a lesser degree, reactive; finally the physical models found in the literature have been compared with the results obtained experimentally. - Highlights: ► The effect of the fuel temperature is substantial on liquid length (up to 15%). ► Fuel temperature has low effect but still appreciable on LOL and ignition delay. ► Theoretical one dimensional spray models are able to reproduce the experimental results with good accuracy.

  18. 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

  19. Experimental study on the impact of operating conditions and fuel composition on PCCI combustion

    Energy Technology Data Exchange (ETDEWEB)

    Leermakers, C.A.J.

    2010-03-15

    Premixed Charge Compression Ignition (PCCI) is a combustion concept that holds the promise of combining emission levels of a spark-ignition (SI) engine with the efficiency of a compressionignition (CI) engine. In a short term scenario, PCCI combustion will be used in the low load part of the engine operating range only. This would guarantee low engine-out emission levels at operating conditions where exhaust temperatures are too low for effective NOx reduction through catalytic after treatment. At higher loads, the engine would run in conventional CI combustion mode, with emission requirements met through catalytic NOx reduction. Implicit with this scenario is that engine hardware design would be very close to that of current modern diesel engines. Compression ratio could be made load dependent through implementation of variable valve actuation. The PCCI experiments presented here have been performed using a modified 6 cylinder 12.6 liter heavy duty DI DAF XE 355 C engine. Experiments are conducted in one dedicated cylinder, which is equipped with a stand-alone fuel injection system, EGR circuit, and air compressor. For the low to medium load operating range the compression ratio has been lowered to 12:1 by means of a thicker head gasket. As engine hardware should - in the short term - preferably remain close to current diesel engines, optimizing operating conditions should focus on parameters like EGR level, intake temperature, intake pressure and injection timing. While past work in the Combustion Technology group has focused on low load PCCI combustion, in this report the effects on engine performance and emission behavior are investigated for both low and medium load PCCI combustion, up to 40% of full load. In the interpretation of experimental results, emphasis lies on the effect on combustion phasing and maximum pressure rise rate, which are inherent challenges to enable viable PCCI combustion. As in the short term scenario fuels will be used that are not too

  20. Damages and resource of locomotive wheels used under the north operating conditions

    Directory of Open Access Journals (Sweden)

    A. V. Grigorev

    2014-01-01

    Full Text Available In operating railway equipment, in particular the elements, such as a wheel and a rail there is damage accumulation of any kind, causing a premature equipment failure. Thus, an analysis of the mechanisms and modeling of damage accumulation and fracture both on the surface and in the bulk material remain a challenge.Data on the defective wheel sets to be subjected to facing has been collected and analyzed to assess the locomotive wheel sets damage of the locomotive fleet company of AK «Yakutia Railways», city of Aldan, The Republic of Sakha (Yakutia. For this purpose, three main locomotives have been examined.The object of research carried out in this paper, is a locomotive wheels tire, which is subjected to cyclic impact (dynamic loads during operation. In this regard, the need arises to determine both the strength of material in response to such shock loads and the quantitative calculation of damage accumulated therein.The accumulated fatigue damage has been attributed to one radial cross section of the wheel coming into contact with the rail once per revolution of the wheel. Consequently, in one revolution a wheel is under one loading cycle. As stated, the average mileage of locomotives to have the unacceptable damages formed on the tread surface is 12 thousand km.Test results establish that along with the high-cycle loading the shock-contact action on rail joints significantly affects the accumulation of damage in the locomotive wheels tire. The number of cycles to failure due to the formation of unacceptable damage in the locomotive wheels tire is N = 2,4×106 and 6×105 cycles, respectively, for fatigue and shock-contact loading.In general, we can say that the problem of higher intensity to form the surface damage is directly related to the operation of the locomotive wheel tire under abnormally low climatic temperatures. With decreasing ambient temperature, this element material rapidly looses its plastic properties, thereby accelerating

  1. Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-08-29

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U3Si2 at LWR conditions. The fission gas behavior of U3Si2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U3Si2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U3Si2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U3Si2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

  2. Visualization of Fuel Cell Water Transport and Performance Characterization under Freezing Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kandlikar, Satish G. [Rochester Inst. of Technology, Rochester, NY (United States); Lu, Zijie [Rochester Inst. of Technology, Rochester, NY (United States); Rao, Navalgund [Rochester Inst. of Technology, Rochester, NY (United States); Sergi, Jacqueline [Rochester Inst. of Technology, Rochester, NY (United States); Rath, Cody [Rochester Inst. of Technology, Rochester, NY (United States); McDade, Christopher [Rochester Inst. of Technology, Rochester, NY (United States); Trabold, Thomas [General Motors, Honeoye Falls, NY (United States); Owejan, Jon [General Motors, Honeoye Falls, NY (United States); Gagliardo, Jeffrey [General Motors, Honeoye Falls, NY (United States); Allen, Jeffrey [Michigan Technological Univ., Houghton, MI (United States); Yassar, Reza S. [Michigan Technological Univ., Houghton, MI (United States); Medici, Ezequiel [Michigan Technological Univ., Houghton, MI (United States); Herescu, Alexandru [Michigan Technological Univ., Houghton, MI (United States)

    2010-05-30

    In this program, Rochester Institute of Technology (RIT), General Motors (GM) and Michigan Technological University (MTU) have focused on fundamental studies that address water transport, accumulation and mitigation processes in the gas diffusion layer and flow field channels of the bipolar plate. These studies have been conducted with a particular emphasis on understanding the key transport phenomena which control fuel cell operation under freezing conditions.

  3. Effect of structural heterogeneity water-coal fuel conditions and characteristics of ignition

    Directory of Open Access Journals (Sweden)

    Syrodoy S.V.

    2015-01-01

    Full Text Available The problem of the particle ignition of coal-water fuel (CWF with a joint course of the main processes of a thermal (thermal conductivity, evaporation, filtration heat and mass transfer, thermal decomposition of the organic part has been solved. According to the results of numerical simulation ways of describing the extent of the influence of the thermophysical properties on the characteristics and conditions of ignition WCF have been set.

  4. Fuel consumption in an air blower for agricultural use under different operating conditions

    OpenAIRE

    Silva, Robson L. da

    2017-01-01

    ABSTRACT Evaluation of fuel consumption in internal combustion engines (ICE) of agricultural machinery and equipment is important in determining the performance under various operating conditions, especially when using biofuels. This study consisted of experimental evaluation of the gasoline (petrol)/ethanol consumption in a two-stroke 1-cylinder ICE, Otto cycle, functioning as an air blower for agriculture and related applications. A methodology for tests of non-automotive ICE, based on ABNT...

  5. Application of contact mechanics for fretting damage of fuel rod: part 1 influence functions and numerical method

    International Nuclear Information System (INIS)

    Kim, H. K.; Yoon, K. H.; Kang, H. S.; Song, G. N.

    1998-01-01

    For the analysis of the fretting problem of the fuel rods, present paper(Part I) shows the numerical method developed for evaluating the stresses on the contact surfaces between the fuel rods and the spacer grids. Theory of Contact Mechanics was incorporated. Contact area was regarded as a plane strain condition, so plane problem was taken into consideration. Normal stress profile on the contact surface was assumed to be Hertzian. As for the direction of the shear load, a closed load path, e.g. load increase in transverse increase in axial decrease in transverse decrease in axial increase in transverse increase in axial direction was considered for simulating the rod vibration in a reactor core. Partial slip problem was consulted. As for the numerical method, a triangular traction element was utilized and the corresponding influence functions were evaluated. Numerical program has been implemented for the present analysis, of which the validity was verified by comparing the Mindlin-Cattaneo solution

  6. Test on Similarity between the Flooded and Optimum Moderation Conditions of the Spent Fuel Storage Pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Gil Soo; Jang, Chang Sun; Woo, Sweng Woong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    In the criticality safety analysis, uncertainty and bias should be considered. The final multiplication factor including uncertainty and bias in addition to calculated k-eff should be below the administrative limit. The administrative limit of spent fuel pool is 0.95 with flooded condition (filled with unborated water), and 0.98 with optimum moderation condition (filled with foggy unborated water, usually occurs near 0.1g/cc water density) for new fuel storage. The bias is determined by comparing the calculation results of the critical experiments ever performed. It is important to choose 'good' experiments which have 'similar' condition with application. To obtain realistic bias, many experiments with similar conditions should be chosen and considered. In previous approach, same critical experiment set are used to determine bias of the flooded and optimum moderation conditions. It would be correct way if two conditions are similar. The similarity test on this paper was performed by TSUNAMI code included in SCALE5.1 package. TSUNAMI code produces sensitivity data for each nuclear reaction by using first order perturbation theory. TSUNAMI code performs forward and adjoint multigroup Monte Carlo calculation. Sensitivity data are obtained by forward and adjoint results. TSUNAMI also produces uncertainty data with sensitivity data and cross section covariance data. In this paper, similarity is determined by comparing energy of average lethargy of fission (EALF), uncertainty data, sensitivity data, and correlation coefficient which is also output of the TSUNAMI code.

  7. Failure Mechanisms and Damage Model of Ductile Cast Iron Under Low-Cycle Fatigue Conditions

    Science.gov (United States)

    Wu, Xijia; Quan, Guangchun; MacNeil, Ryan; Zhang, Zhong; Sloss, Clayton

    2014-10-01

    Strain-controlled low-cycle fatigue (LCF) tests were conducted on ductile cast iron (DCI) at strain rates of 0.02, 0.002, and 0.0002/s in the temperature range from room temperature to 1073 K (800 °C). A constitutive-damage model was developed within the integrated creep-fatigue theory (ICFT) framework on the premise of strain decomposition into rate-independent plasticity and time-dependent creep. Four major damage mechanisms: (i) plasticity-induced fatigue, (ii) intergranular embrittlement (IE), (iii) creep, and (iv) oxidation were considered in a nonlinear creep-fatigue interaction model which represents the overall damage accumulation process consisting of oxidation-assisted fatigue crack nucleation and propagation in coalescence with internally distributed damage ( e.g., IE and creep), leading to final fracture. The model was found to agree with the experimental observations of the complex DCI-LCF phenomena, for which the linear damage summation rule would fail.

  8. Study on Characteristics of Co-firing Ammonia/Methane Fuels under Oxygen Enriched Combustion Conditions

    Science.gov (United States)

    Xiao, Hua; Wang, Zhaolin; Valera-Medina, Agustin; Bowen, Philip J.

    2018-06-01

    Having a background of utilising ammonia as an alternative fuel for power generation, exploring the feasibility of co-firing ammonia with methane is proposed to use ammonia to substitute conventional natural gas. However, improvement of the combustion of such fuels can be achieved using conditions that enable an increase of oxygenation, thus fomenting the combustion process of a slower reactive molecule as ammonia. Therefore, the present study looks at oxygen enriched combustion technologies, a proposed concept to improve the performance of ammonia/methane combustion. To investigate the characteristics of ammonia/methane combustion under oxygen enriched conditions, adiabatic burning velocity and burner stabilized laminar flame emissions were studied. Simulation results show that the oxygen enriched method can help to significantly enhance the propagation of ammonia/methane combustion without changing the emission level, which would be quite promising for the design of systems using this fuel for practical applications. Furthermore, to produce low computational-cost flame chemistry for detailed numerical analyses for future combustion studies, three reduced combustion mechanisms of the well-known Konnov's mechanism were compared in ammonia/methane flame simulations under practical gas turbine combustor conditions. Results show that the reduced reaction mechanisms can provide good results for further analyses of oxygen enriched combustion of ammonia/methane. The results obtained in this study also allow gas turbine designers and modellers to choose the most suitable mechanism for further combustion studies and development.

  9. Durability study and lifetime prediction of baseline proton exchange membrane fuel cell under severe operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Marrony, M.; Quenet, S.; Aslanides, A. [European Institute for Energy Research, Emmy-Noether Strasse 11, 76131 Karlsruhe (Germany); Barrera, R.; Ginocchio, S.; Montelatici, L. [Edison, Via Giorgio La Pira 2, 10028 Trofarello (Italy)

    2008-08-01

    Comparative studies of mechanical and electrochemical properties of Nafion{sup registered} - and sulfonated polyetheretherketone polymer-type membranes are carried out under severe fuel cell conditions required by industrials, within stationary and cycling electric load profiles. These membranes are proposed to be used in PEM between 70 and 90 C as fluorinated or non-fluorinated baseline membranes, respectively. Thus, though the performance of both membranes remains suitable, Nafion{sup registered} backbone brought better mechanical properties and higher electrochemical stabilities than sulfonated polyetheretherketone backbone. The performance stability and the mechanical strength of the membrane-electrode assembly were shown to be influenced by several intrinsic properties of the membrane (e.g., thermal pre-treatment, thickness) and external conditions (fuel cell operating temperature, relative humidity). Finally, a lifetime prediction for membranes under stationary conditions is proposed depending on the operation temperature. At equivalent thicknesses (i.e. 50 {mu}m), Nafion{sup registered} membranes were estimated able to operate into the 80-90 C range while sulfonated polyetheretherketone would be limited into the 70-80 C range. This approach brings baseline information about the capability of these types of polymer electrolyte membrane under fuel cell critical operations. Finally, it is revealed as a potential tool for the selection of the most promising advanced polymers for the ensuing research phase. (author)

  10. 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

  11. Sulphation of calcium-based sorbents in circulating fluidised beds under oxy-fuel combustion conditions

    Energy Technology Data Exchange (ETDEWEB)

    Francisco Garcia-Labiano; Luis F. de Diego; Alberto Abad; Pilar Gayan; Margarita de las Obras-Loscertales; Aranzazu Rufas; Juan Adanez [Instituto de Carboquimica (CSIC), Zaragoza (Spain). Dept. Energy and Environment

    2009-07-01

    Sulphur Retention (SR) by calcium-based sorbents is a process highly dependent on the temperature and CO{sub 2} concentration. In circulating fluidised beds combustors (CFBC's) operating under oxy-fuel conditions, the sulphation process takes place in atmospheres enriched in CO{sub 2} with bed concentrations that can vary from 40 to 95%. Under so high CO{sub 2} concentrations, very different from that in conventional coal combustion atmosphere with air, the calcination and sulphation behaviour of the sorbent must be defined to optimise the SR process in the combustor. The objective of this work was to determine the SO{sub 2} retention capacity of a Spanish limestone at typical oxy-fuel conditions in CFBC's. Long term duration tests of sulphation (up to 24 h), to simulate the residence time of sorbents in CFBC's, were carried out by thermogravimetric analysis (TGA). Clear behaviour differences were found under calcining and non-calcining conditions. Especially relevant was the result obtained at calcining conditions but close to the thermodynamic temperature given for sorbent calcination. This situation must be avoided in CFBC's because the CO{sub 2} produced inside the particle during calcination can destroy the particles if a non-porous sulphate product layer has been formed around the particle. The effect of the main variables on the sorbent sulphation such as SO{sub 2} concentration, temperature, and particle size were analysed in the long term TGA tests. These data were also used to determine the kinetic parameters for the sulphation under oxy-fuel combustion conditions, which were able to adequately predict the sulphation conversion values in a wide range of operating conditions. 20 refs., 5 figs., 2 tabs.

  12. ILLEGAL ACTS - CONDITION OF LIABILITY FOR DAMAGES CAUSED IN EXERCISING LEGAL LABOR RELATIONS

    Directory of Open Access Journals (Sweden)

    Ştefania-Alina Dumitrache

    2014-11-01

    Full Text Available According to article 253 and 254 of Labor Code, both employers and employees are responsible under the rules and principles of contractual liability for damages to the other party of legal labor relationship and we emphasize that this is not purely civil liability, but a variety of it, determined by the specific peculiarities of legal labor relations. Thus, we highlight that labor law provisions which refer to liability for damages complement, unquestionably, with the common law relating to civil liability. The paper analyzes the objective basis of legal accountability, namely the illicit act causing damages committed in fulfilling labor duties or in connection tot hem, therewith the method detailed and comparative documentation of legislation in the field and relevant doctrine.

  13. Improvement of the vibration of the test fuel(Type-B) with a guide tube under operational condition

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong Seung; Yim, Jeong Sik; Lim, I. C. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-04-01

    The Type-B test fuel for the Hanaro has a flexible guide tube on top of the fuel to lead and guide the instrumentation wires. Depending on the flow condition in the reactor, the fuel is susceptible to vibration. During the test operation of the fuel, a fairly large amplitude vibration was observed and the possibility of flow tube contact with adjacent flow tubes, due to the excessive vibration of the fuel, and consequent wear or defect of the flow tubes were raised. Thus, to know the vibration characteristics as well as whether the flow tube contact each other, analyses of the Type-B fuel the dummy fuel were performed by BEVIRA and ANSYS. Besides the analyses, vibration tests using the dummy fuel in air and with Type-B fuel in the core at zero power under operational flow condition were executed. The results from the analyses were compared with those from tests to validate the analyses. From the deflection test of the dummy fuel in air to get the maximum displacement of the flow tube at the top, the flow tube were found to contact each other. For the prevention of the contact of the flow tubes caused by the excessive vibration of the guide tube, an additional support to the guide tube was proposed. With the additional support, analysis and in core vibration test under operational flow condition were conducted and there found to be no excessive vibration any more. 6 refs., 16 figs., 6 tabs. (Author)

  14. Burn-up Credit Criticality Safety Benchmark Phase III-C. Nuclide Composition and Neutron Multiplication Factor of a Boiling Water Reactor Spent Fuel Assembly for Burn-up Credit and Criticality Control of Damaged Nuclear Fuel

    International Nuclear Information System (INIS)

    Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.

    2016-01-01

    Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are

  15. Assessment of conditions of the spent nuclear fuel stored in the stainless steel channel-holders

    International Nuclear Information System (INIS)

    Pesic, M.; Sotic, O.; Cupac, S.; Maksin, T.; Dasic, N.

    2003-01-01

    The IAEA technical co-operation project 'Safe Removal of Spent Fuel of the Vinca RA Research Reactor' is carried out at the Vinca Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since January 2003. Present activities will provide up-to-date information on the conditions of the spent nuclear fuel, stored in the stainless steel channel-holders ('chekhols') and on the water quality in the storage basins. Water samples taken out from the chekhols and the basins are measured to determine their activity and chemical parameters. Until September 2003, about 1/3 of the chekhols containing spent fuel elements with initial enrichment of 2% and 80% of uranium were inspected. High activity of Cs-137 was found in several water samples taken out from chekhols. All water samples show very high electrical conductivity, while those taken from the basins show the presence of chlorides and aluminium ions, too. Information on established procedures and measuring results are given in this paper. The obtained results, so far, show that the spent nuclear fuel elements are leaking in about 10% of chekhols. (author)

  16. Performance Evaluation of the Neutron Coincidence Counter for the Advanced Spent Fuel Conditioning Process

    International Nuclear Information System (INIS)

    Lee, S.Y.; Li, T.K.; Menlove, Howard O.; Kim, H.D.; Ko, W.I.; Park, S.W.

    2005-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical dry reprocessing technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. With this technology, a significant reduction of the volume and heat load of spent fuel is expected, which could decrease the burden of safety and economics. In this study, MCNPX code calculations were carried out to estimate the performance of a neutron coincidence counter designed for measruement of the process materials in the pilot-scale ACP facility. To verify the design requirement, the singles and doubles counting rates of the detectors were simulated with the latest coincidence capability of the MCNPX code. Then, the precision of the coincidence measurements were evaluated on various process materials from the ACP. It was verified that the performance of the neutron coincidence counter could meet the design criteria for all samples in the ACP, and the material accounting system for the pilot-scale ACP facility could meet the IAEA safeguards goals.

  17. RODSWELL: a computer code for the thermomechanical analysis of fuel rods under LOCA conditions

    International Nuclear Information System (INIS)

    Casadei, F.; Laval, H.; Donea, J.; Jones, P.M.; Colombo, A.

    1984-01-01

    The code calculates the variation in space and time of all significant fuel rod variables, including fuel, gap and cladding temperature, fuel and cladding deformation, cladding oxidation and rod internal pressure. The code combines a transient 2-dimensional heat conduction code and a 1-dimensional mechanical model for the cladding deformation. The first sections of this report deal with the heat conduction model and the finite element discretization used for the thermal analysis. The mechanical deformation model is presented next: modelling of creep, phase change and oxidation of the zircaloy cladding is discussed in detail. A model describing the effect of oxidation and oxide cracking on the mechanical strength of the cladding is presented too. Next a mechanical restraint model, which allows the simulation of the presence of the neighbouring rods and is particularly important in assessing the amount of channel blockage during a transient, is presented. A description of the models used for the coolant conditions and for the power generation follows. The heat source can be placed either in the fuel or in the cladding, and direct or indirect clad heating by electrical power can be simulated. Then a section follows, dealing with the steady-state and transient types of calculation and with the automatic variable time step selection during the transient. The last sections deal with presentation of results, graphical output, test problems and an example of general application of the code

  18. Viscoelastic response of HTPB based solid fuel to horizontal and vertical storage slumping conditions and it's affect on service life

    International Nuclear Information System (INIS)

    Nawaz, Q.; Nizam, F.

    2011-01-01

    Frequent use of solid fuels as thrust generating energy source in modern day space vehicle systems has created a need to assess their serviceability for long term storage under various conditions. Solid fuel grain, the most important part of any solid fuel system, responds visco elastically to any loading condition. For the assessment of the service life of any solid fuel system, the solid fuel grain has to be structurally evaluated in applied storage conditions. Structural integrity of the grain is exceptionally significant to guarantee the successful operation of the solid fuel system. In this work, numerical simulations have been performed to assess the mechanical stresses and strains induced in an HTPB based solid fuel grain during service life employing ABAQUS standard FEA software using 4-node bilinear quadrilateral elements. For finite element analysis (FEA), typical 2-D and p/nth axisymmetric section of 5-point (n) star grain geometry is considered. Mechanical loads include the horizontal or vertical 1-g (solid fuel weight) storage condition. The simulation results are compared with the analytical results for the same grain geometry. Analytically measured slump deflections in grain segment at various storage times have been found in good relation with the FEA based simulation results. This proves the validity of the procedure adopted and is helpful in assessment of the service life of solid fuel systems. (author)

  19. Combined advanced finishing and UV laser conditioning process for producing damage resistant optics

    Science.gov (United States)

    Menapace, Joseph A.; Peterson, John E.; Penetrante, Bernardino M.; Miller, Philip E.; Parham, Thomas G.; Nichols, Michael A.

    2005-07-26

    A method for reducing the density of sites on the surface of fused silica optics that are prone to the initiation of laser-induced damage, resulting in optics which have far fewer catastrophic defects, and are better capable of resisting optical deterioration upon exposure to a high-power laser beam.

  20. Microstructural evolution adjacent to grain boundaries under cascade damage conditions and helium production

    DEFF Research Database (Denmark)

    Trinkaus, H.; Singh, B.N.; Victoria, M.

    1996-01-01

    the cascade damage is accompanied by a high helium production rate. It is shown that, in this case, the width of the peak zone is controlled by the (mostly invisible) bubble structure rather than by the (visible) void structure. The reduced swelling relative to that under neutron irradiation is attributed...

  1. Effect of oxygenate additive on diesel engine fuel consumption and emissions operating with biodiesel-diesel blend at idling conditions

    Science.gov (United States)

    Mahmudul, H. M.; Hagos, F. Y.; Mamat, R.; Noor, M. M.; Yusri, I. M.

    2017-10-01

    Biodiesel is promising alternative fuel to run the automotive engine but idling is the main problem to run the vehicles in a big city. Vehicles running with idling condition cause higher fuel supply and higher emission level due to being having fuel residues in the exhaust. The purpose of this study is to evaluate the impact of alcohol additive on fuel consumption and emissions parameters under idling conditions when a multicylinder diesel engine operates with the diesel-biodiesel blend. The study found that using 5% butanol as an additive with B5 (5% Palm biodiesel + 95% diesel) blends fuel lowers brake specific fuel consumption and CO emissions by 38% and 20% respectively. But the addition of butanol increases NOx and CO2 emissions. Based on the result it can be said that 5% butanol can be used in a diesel engine with B5 without any engine modifications to tackle the idling problem.

  2. Metallographic examination of (uth) O2 and UO2 fuel tested in power ramp conditions in triga reactor

    International Nuclear Information System (INIS)

    Ioncescu, M.; Uta, O.

    2015-01-01

    The purpose of this paper is to determine the behavior of two fuel experimental elements (EC1 and EC2), by destructive post-irradiation examination. The fuel elements were mounted inside a pattern port, one in extension of the other and irradiated in power ramp conditions in order to check their behavior. Fuel element 1 (EC1) contains (UTh)O''2 pellet, and other one (EC2) UO''2 pellet. The results of destructive post-irradiation examination are evidenced by metallographic and ceramographic analyses. The data obtained from the post-irradiation examinations are used, first to confirm the security, reliability and nuclear fuel performance, and second, for the development of CANDU fuel. The results obtained by destructive examinations regarding the integrity, sheath hydrating and oxidation as well as the structural modifications are typical for fuel elements tested in power ramp conditions. (authors)

  3. Premix fuels study applicable to duct burner conditions for a variable cycle engine

    Science.gov (United States)

    Venkataramani, K. S.

    1978-01-01

    Emission levels and performance of a premixing Jet-A/air duct burner were measured at reference conditions representative of take-off and cruise for a variable cycle engine. In a parametric variation sequence of tests, data were obtained at inlet temperatures of 400, 500 and 600K at equivalence ratios varying from 0.9 to the lean stability limit. Ignition was achieved at all the reference conditions although the CO levels were very high. Significant nonuniformity across the combustor was observed for the emissions at the take-off condition. At a reference Mach number of 0.117 and an inlet temperature of 600K, corresponding to a simulated cruise condition, the NOx emission level was approximately 1 gm/kg-fuel.

  4. 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)

  5. 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)

  6. 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

  7. Modelling disassembled fuel bundles using CATHENA MOD-3.5a under LOCA/LOECC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Lei, Q M; Sanderson, D B; Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-31

    CATHENA MOD-3.5a is a multipurpose thermalhydraulic computer code developed primarily to analyse postulated loss-of-coolant scenarios for CANDU nuclear reactors. The code contains a generalized heat transfer package that enables it to model the behaviour of a fuel channel in great detail. Throughout the development of the CATHENA code, considerable effort has been devoted to evaluating, validating and documenting its overall capability as a design and safety assessment tool. Specific attention has focused on its ability to predict fuel channel behaviour under postulated accident conditions. This paper describes an investigation of CATHENA`s ability to predict the thermal-chemical responses of a fuel channel in which the 37-element bundles were assumed to disassemble and rearrange into a closed-packed stack of elements at the bottom of the pressure tube. A representative disassembled bundle geometry was modelled during a simulated loss-of-coolant accident scenario using CATHENA MOD-3.5a/Rev 0, with superheated steam being the only coolant available. Thermal conduction in the radial and circumferential directions was calculated for individual fuel elements, the pressure tube, and the calandria tube. Radiation view factors for the intact and disassembled bundle geometries were calculated using a CATHENA utility program. Inter-element metal-to-metal contact was accounted for using the CATHENA solid-solid contact model. An offset pressure-tube configuration, representing a partially sagged pressure tube, and the effect of steam starvation on the exothermic zirconium-steam reaction, were included in the CATHENA model. The CATHENA-predicted results show a dramatic suppression of heat generation from the zirconium-steam reaction when bundle disassembly is initiated. The predicted results show a smaller temperature increase in the fuel sheaths and the pressure tube for the disassembled bundle geometry, compared to the temperature excursion for the intact bundle. (author

  8. Local area water removal analysis of a proton exchange membrane fuel cell under gas purge conditions.

    Science.gov (United States)

    Lee, Chi-Yuan; Lee, Yu-Ming; Lee, Shuo-Jen

    2012-01-01

    In this study, local area water content distribution under various gas purging conditions are experimentally analyzed for the first time. The local high frequency resistance (HFR) is measured using novel micro sensors. The results reveal that the liquid water removal rate in a membrane electrode assembly (MEA) is non-uniform. In the under-the-channel area, the removal of liquid water is governed by both convective and diffusive flux of the through-plane drying. Thus, almost all of the liquid water is removed within 30 s of purging with gas. However, liquid water that is stored in the under-the-rib area is not easy to remove during 1 min of gas purging. Therefore, the re-hydration of the membrane by internal diffusive flux is faster than that in the under-the-channel area. Consequently, local fuel starvation and membrane degradation can degrade the performance of a fuel cell that is started from cold.

  9. 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)

  10. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

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

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  11. Damage evaluation of Anastrepha fraterculus (Wiedemann) (Diptera: Tephritidae) on five apple cultivars under laboratory conditions

    International Nuclear Information System (INIS)

    Branco, E.S.; Vendramin, J.D.; Denardi, F.; Nora, I.

    1999-01-01

    The apple production losses in southern Brazil caused by the attack of the fruit fly Anastrepha fraterculus can reach up to 100% in some years. Its control demands intensive systematic sprays of insecticides, which increase production costs and affect environmental quality. In terms of integrated pest management, the use of resistant cultivars represents one of the most important alternatives to control this apple pest. With the objective of identifying sources of host plant resistance, apple fruits of different cultivars from the Clonal Germplasm Repository of the EPAGRI Research Station of Cacador were tested. The experiment consisted of 5 treatments (cultivars) with 5 replicates. Fruits at the harvest stage were used. The fruits were placed in boxes (40x110 cm), where they were exposed to oviposition by the fruit fly. After infestation, fruits were left on shelves at room temperature for 10 days in order to evaluate the damage level according to the following scale: 1 = fruit without attack; 2 = fruit with punctures and/or deformation without galleries; 3 = fruit with punctures and/or deformation and galleries; 4 = fruit with punctures and/or deformations, galleries and larvae. The Gala cultivar was the most susceptible, with an average damage level of 3.4, differing from the cultivars Fuji and Royal Red Delicious (damage levels of 1.6 and 1.2, respectively). The Belgolden and Sansa clones presented intermediate damage levels. A. fraterculus preferred to oviposit in the Golden Delicious group compared with the Delicious group. These studies suggest good possibilities for reduction of insecticide sprays to control the fruit fly in the cv. Fuji, as well as the incorporation of resistance factor in apple cultivars. (author)

  12. Condition-Based Maintenance Strategy for Production Systems Generating Environmental Damage

    Directory of Open Access Journals (Sweden)

    L. Tlili

    2015-01-01

    Full Text Available We consider production systems which generate damage to environment as they get older and degrade. The system is submitted to inspections to assess the generated environmental damage. The inspections can be periodic or nonperiodic. In case an inspection reveals that the environmental degradation level has exceeded the critical level U, the system is considered in an advanced deterioration state and will have generated significant environmental damage. A corrective maintenance action is then performed to renew the system and clean the environment and a penalty has to be paid. In order to prevent such an undesirable situation, a lower threshold level L is considered to trigger a preventive maintenance action to bring back the system to a state as good as new at a lower cost and without paying the penalty. Two inspection policies are considered (periodic and nonperiodic. For each one of them, a mathematical model and a numerical procedure are developed to determine simultaneously the preventive maintenance (PM threshold L∗ and the inspection sequence which minimize the average long-run cost per time unit. Numerical calculations are performed to illustrate the proposed maintenance policies and highlight their main characteristics with respect to relevant input parameters.

  13. Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

    International Nuclear Information System (INIS)

    Cruz, C.B.L.

    1990-01-01

    The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-30

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

  15. An Improved Gaussian Mixture Model for Damage Propagation Monitoring of an Aircraft Wing Spar under Changing Structural Boundary Conditions

    Science.gov (United States)

    Qiu, Lei; Yuan, Shenfang; Mei, Hanfei; Fang, Fang

    2016-01-01

    Structural Health Monitoring (SHM) technology is considered to be a key technology to reduce the maintenance cost and meanwhile ensure the operational safety of aircraft structures. It has gradually developed from theoretic and fundamental research to real-world engineering applications in recent decades. The problem of reliable damage monitoring under time-varying conditions is a main issue for the aerospace engineering applications of SHM technology. Among the existing SHM methods, Guided Wave (GW) and piezoelectric sensor-based SHM technique is a promising method due to its high damage sensitivity and long monitoring range. Nevertheless the reliability problem should be addressed. Several methods including environmental parameter compensation, baseline signal dependency reduction and data normalization, have been well studied but limitations remain. This paper proposes a damage propagation monitoring method based on an improved Gaussian Mixture Model (GMM). It can be used on-line without any structural mechanical model and a priori knowledge of damage and time-varying conditions. With this method, a baseline GMM is constructed first based on the GW features obtained under time-varying conditions when the structure under monitoring is in the healthy state. When a new GW feature is obtained during the on-line damage monitoring process, the GMM can be updated by an adaptive migration mechanism including dynamic learning and Gaussian components split-merge. The mixture probability distribution structure of the GMM and the number of Gaussian components can be optimized adaptively. Then an on-line GMM can be obtained. Finally, a best match based Kullback-Leibler (KL) divergence is studied to measure the migration degree between the baseline GMM and the on-line GMM to reveal the weak cumulative changes of the damage propagation mixed in the time-varying influence. A wing spar of an aircraft is used to validate the proposed method. The results indicate that the crack

  16. Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

    International Nuclear Information System (INIS)

    Kulikov, I.S.; Tverkovkin, B.E.

    1984-01-01

    The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

  17. Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

  18. Sensitivity Analysis of Heavy Fuel Oil Spray and Combustion under Low-Speed Marine Engine-Like Conditions

    Directory of Open Access Journals (Sweden)

    Lei Zhou

    2017-08-01

    Full Text Available On account of their high power, thermal efficiency, good reliability, safety, and durability, low-speed two-stroke marine diesel engines are used as the main drive devices for large fuel and cargo ships. Most marine engines use heavy fuel oil (HFO as the primary fuel, however, the physical and chemical characteristics of HFO are not clear because of its complex thermophysical properties. The present study was conducted to investigate the effects of fuel properties on the spray and combustion characteristics under two-stroke marine engine-like conditions via a sensitivity analysis. The sensitivity analysis of fuel properties for non-reacting and reacting simulations are conducted by comparing two fuels having different physical properties, such as fuel density, dynamic viscosity, critical temperature, and surface tension. The performances of the fuels are comprehensively studied under different ambient pressures, ambient temperatures, fuel temperatures, and swirl flow conditions. From the results of non-reacting simulations of HFO and diesel fuel properties in a constant volume combustion chamber, it can be found that the increase of the ambient pressure promotes fuel evaporation, resulting in a reduction in the steady liquid penetration of both diesel and HFO; however, the difference in the vapor penetrations of HFO and diesel reduces. Increasing the swirl flow significantly influences the atomization of both HFO and diesel, especially the liquid distribution of diesel. It is also found that the ambient temperature and fuel temperature have the negative effects on Sauter mean diameter (SMD distribution. For low-speed marine engines, the combustion performance of HFO is not sensitive to activation energy in a certain range of activation energy. At higher engine speed, the difference in the effects of different activation energies on the in-cylinder pressure increases. The swirl flow in the cylinder can significantly promote fuel evaporation and

  19. Study thermofluidynamic of the sub frame of fuel in the cell of discharge of the ATC; Estudio termofluidodinamico del bastidor auxiliar de combustible en la celda de descarga del ATC

    Energy Technology Data Exchange (ETDEWEB)

    Penalva, J.; Feria, F.; Herranz, L. E.

    2014-07-01

    The objective of this work was to determine the conditions that guarantee the maintenance of the State of the fuel during hypothetical stays in the discharge of a postulated ATC cell. The study includes three different conditions fuel element: intact, defective drawer of damaged fuel and defective without drawer of damaged fuel. (Author)

  20. 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

  1. 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.

  2. Motorcycle emissions and fuel consumption in urban and rural driving conditions.

    Science.gov (United States)

    Chen, K S; Wang, W C; Chen, H M; Lin, C F; Hsu, H C; Kao, J H; Hu, M T

    2003-08-01

    This work reports sampling of motorcycle on-road driving cycles in actual urban and rural environments and the development of representative driving cycles using the principle of least total variance in individual regions. Based on the representative driving cycles in individual regions, emission factors for carbon monoxide (CO), hydrocarbons (HC), nitrogen oxides (NO(x)=NO+NO(2)) and carbon dioxide (CO(2)), as well as fuel consumption, were determined using a chassis dynamometer. The measurement results show that the representative driving cycles are almost identical in the three largest cities in Taiwan, but they differ significantly from the rural driving cycle. Irrespective of driving conditions, emission factors differ insignificantly between the urban and rural regions at a 95% confidence level. However, the fuel consumption in urban centers is approximately 30% higher than in the rural regions, with driving conditions in the former usually poor compared to the latter. Two-stroke motorcycles generally have considerably higher HC emissions and quite lower NO(x) emissions than those of four-stroke motorcycles. Comparisons with other studies suggest that factors such as road characteristics, traffic volume, vehicle type, driving conditions and driver behavior may affect motorcycle emission levels in real traffic situations.

  3. Motorcycle emissions and fuel consumption in urban and rural driving conditions

    International Nuclear Information System (INIS)

    Chen, K.S.; Wang, W.C.; Chen, H.M.; Lin, C.F.; Hsu, H.C.; Kao, J.H.; Hu, M.T.

    2003-01-01

    This work reports sampling of motorcycle on-road driving cycles in actual urban and rural environments and the development of representative driving cycles using the principle of least total variance in individual regions. Based on the representative driving cycles in individual regions, emission factors for carbon monoxide (CO), hydrocarbons (HC), nitrogen oxides (NO x =NO+NO 2 ) and carbon dioxide (CO 2 ), as well as fuel consumption, were determined using a chassis dynamometer. The measurement results show that the representative driving cycles are almost identical in the three largest cities in Taiwan, but they differ significantly from the rural driving cycle. Irrespective of driving conditions, emission factors differ insignificantly between the urban and rural regions at a 95% confidence level. However, the fuel consumption in urban centers is approximately 30% higher than in the rural regions, with driving conditions in the former usually poor compared to the latter. Two-stroke motorcycles generally have considerably higher HC emissions and quite lower NO x emissions than those of four-stroke motorcycles. Comparisons with other studies suggest that factors such as road characteristics, traffic volume, vehicle type, driving conditions and driver behavior may affect motorcycle emission levels in real traffic situations

  4. Behavior of pre-irradiated fuel under a simulated RIA condition. Results of NSRR Test JM-5

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Tanzawa, Sadamitsu; Ishijima, Kiyomi; Kobayashi, Shinsho; Kamata, Hiroshi; Homma, Kozo; Sakai, Haruyuki.

    1995-11-01

    This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-5, conducted in the Nuclear Safety Research Reactor (NSRR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and interpretations and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 25.7 MWd/kgU with average linear heat rate of 33.4 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a desposited energy of 223 ± 7 cal/g·fuel (0.93 ± 0.03 kJ/g·fuel) and a peak fuel enthalpy of 167 ± 5 cal/g·fuel (0.70 ± 0.02 kJ/g·fuel) under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The Test JM-5 resulted in cladding failure. More than twenty small cracks were found in the post-test cladding, and most of the defects located in pre-existing locally hydrided region. The result indicates an occurrence of fuel failure by PCMI (pellet/cladding mechanical interaction) in combination with decreased integrity of hydrided cladding. (author)

  5. Safety evaluation report of hot cell facilities for demonstration of advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    You, Gil Sung; Choung, W. M.; Ku, J. H.; Cho, I. J.; Kook, D. H.; Park, S. W.; Bek, S. Y.; Lee, E. P.

    2004-10-01

    The advanced spent fuel conditioning process(ACP) proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. In the next phase(2004∼2006), the hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α- type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β- type will be refurbished to minimize construction expenditures of hot cell facility. Up to now, the detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. The design data were submitted for licensing which was necessary for construction and operation of hot cell facilities. The safety investigation of KINS on hot cell facilities was completed, and the license for construction and operation of hot cell facilities was acquired already from MOST. In this report, the safety analysis report submitted to KINS was summarized. And also, the questionnaires issued from KINS and answers of KAERI in process of safety investigation were described in detail

  6. Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

    2003-04-01

    In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed

  7. Preliminary assessment of safeguardability on the concepture design of advanced spent fuel conditioning process

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Yoon; Ha, Jang Ho; Ko, Won Il; Song, Dae Yong; Kim, Ho Dong

    2003-04-01

    In this report, a preliminary study on the safeguardability of ACP (Advanced spent fuel Conditioning Process) was conducted with Los Alamos National Laboratory. The proposed ACP concept is an electrometallurgical treatment technique to convert oxide-type spent nuclear fuels into metal forms, which can achieve significant reduction of the volume and heat load of spent fuel to be stored and disposed of. For the safeguardability analysis of the ACP facility, sub-processes and their KMPs (Key Measurement Points) were defined first, and then their material flows were analyzed. Finally, the standard deviation of the Inventory Difference (ID) value of the facility was estimated with assumption by assuming international target values for the uncertainty of measurement methods and their uncertainty. From the preliminary calculation, we concluded that if the assumptions regarding measurement instruments can be achieved in a safeguards system for the ACP facility, the safeguards goals of International Atomic Energy Agency (IAEA) could be met. In the second phase of this study, further study on sensitivity analyses considering various factors such as measurement errors, facility capacities, MBA periods etc. may be needed.

  8. The simulation of CANDU fuel channel behavior in thermal transient conditions

    International Nuclear Information System (INIS)

    Mihalache, M.; Roth, M.; Radu, V.; Dumitrescu, I.

    2005-01-01

    In certain LOCA conditions into the CANDU fuel channel, is possible the ballooning of the pressure tube and the contact with the calandria tube. After the contact moment, a radial heat transfer to the moderator through the contact area is occurs. When the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. Thus, the fuel channel could lose its integrity. This paper present a computer code, DELOCA, developed in INR, which simulate the transient thermo-mechanical behaviour of CANDU fuel channel before and after contact. The code contains few models: alloy creep, heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. It was verified step by step by Contact1 and Cathena codes. In this paper, the results obtained at different temperature increasing rates are presented. Also, the contact moment for a RIH 5% postulated accident was presented. The input data was furnished by the Cathena thermo-hydraulic code. (author)

  9. Effect of operating conditions on energy efficiency for a small passive direct methanol fuel cell

    International Nuclear Information System (INIS)

    Chu Deryn; Jiang Rongzhong

    2006-01-01

    Energy conversion efficiency was studied in a direct methanol fuel cell (DMFC) with an air-breathing cathode using Nafion 117 as electrolyte membrane. The effect of operating conditions, such as methanol concentration, discharge voltage and temperature, on Faradic and energy conversion efficiencies was analyzed under constant voltage discharge with quantitative amount of fuel. Both of Faradic and energy conversion efficiencies decrease significantly with increasing methanol concentration and environmental temperature. The Faradic conversion efficiency can be as high as 94.8%, and the energy conversion efficiency can be as high as 23.9% if the environmental temperature is low enough (10 deg. C) under constant voltage discharge at 0.6 V with 3 M methanol for a DMFC bi-cell. Although higher temperature and higher methanol concentration can achieve higher discharge power, it will result in considerable losses of Faradic and energy conversion efficiencies for using Nafion electrolyte membrane. Development of alternative highly conductive membranes with significantly lower methanol crossover is necessary to avoid loss of Faradic conversion efficiency with temperature and with fuel concentration

  10. Obtaining alternative fuel from sweet sorghum in the conditions of the Republic of Tatarstan

    Science.gov (United States)

    Kashapov, N. F.; Nafikov, M. M.; Gilmanshin, I. R.; Nigmatzyanov, A. R.

    2017-09-01

    In the agro-industrial complex of the Russian Federation the main types of energy resources is the FCM (fuel-lubricating materials), electricity, coal and gas. Priority energy is determined depending on the orientation of the activity of the agricultural enterprise. In the cost of getting products one of the key factors is its energy intensity. Under the energy intensity means the amount of energy expended per unit of finished product. Domestic manufacturers lag behind on this indicator from their foreign colleagues. Greatly influenced by the climatic conditions of production, which affects the amount of energy expended annually becoming more expensive. In the article, the authors address a topical issue of renewable(alternative) fuels from sweet sorghum in the stems of which contains from 14 to 21 % sugar. In the Republic of Tatarstan tested and introduced varieties of sweet sorghum. On the basis of literary data and carried out their own research given a set of equipment and presents non-waste production chain of biodiesel and fuel pellets from stems of sweet sorghum.

  11. Design information verification for spent fuel conditioning plants and for geological repositories

    International Nuclear Information System (INIS)

    Myatt, J.; Ward, M.D.

    1995-01-01

    The disposal of spent fuel is a major option for the back-end of the nuclear fuel cycle. It will require the construction, operation and eventual closure of conditioning plants and geological repositories. Consequently, a safeguards approach including Design Information Verification (DIV) must be developed for these facilities. DIV Is the examination of a completed facility to verify that it has been built to the design declared by the operator. Although DIV takes place chiefly before a plant begins routine operation, there is a continuing interest in ensuring that the plant remains as declared. That is, that the continuity of knowledge of design information is maintained during the operational phase of the plant and also post closure if necessary. A major problem with DIV of a repository is that there will be continuous structural changes during its operational life requiring advanced or special techniques for reverification. Some of these are briefly reviewed. Furthermore, since a disposal facility is expected to be operational for several decades, new mining technology may also have an impact on the DIV methods employed. Another factor in the safeguards supervision of a repository is that when the fuel has been backfilled and/or scaled in place a reassay will be a very costly exercise. The role of DIV in such novel circumstances must, therefore, be fully considered

  12. A mechanical deformation model of metallic fuel pin under steady state conditions

    International Nuclear Information System (INIS)

    Lee, D. W.; Lee, B. W.; Kim, Y. I.; Han, D. H.

    2004-01-01

    As a mechanical deformation model of the MACSIS code predicts the cladding deformation due to the simple thin shell theory, it is impossible to predict the FCMI(Fuel-Cladding Mechanical Interaction). Therefore, a mechanical deformation model used the generalized plane strain is developed. The DEFORM is a mechanical deformation routine which is used to analyze the stresses and strains in the fuel and cladding of a metallic fuel pin of LMRs. The accuracy of the program is demonstrated by comparison of the DEFORM predictions with the result of another code calculations or experimental results in literature. The stress/strain distributions of elastic part under free thermal expansion condition are completely matched with the results of ANSYS code. The swelling and creep solutions are reasonably well agreed with the simulations of ALFUS and LIFE-M codes, respectively. The predicted cladding strains are under estimated than experimental data at the range of high burnup. Therefore, it is recommended that the fine tuning of the DEFORM based on various range of experimental data

  13. DNS Study of the Ignition of n-Heptane Fuel Spray under HCCI Conditions

    Science.gov (United States)

    Wang, Yunliang; Rutland, Christopher J.

    2004-11-01

    Direct numerical simulations are carried out to investigate the mixing and auto-ignition processes of n-heptane fuel spray in a turbulent field using a skeletal chemistry mechanism with 44 species and 112 reactions. For the solution of the carrier gas fluid, we use the Eulerian method, while for the fuel spray, the Lagrangian method is used. We use an eighth-order finite difference scheme to calculate spacial derivatives and a fourth-order Runge-Kutta scheme for the time integration. The initial gas temperature is 926 K and the initial gas pressure is 30 atmospheres. The initial global equivalence ratio based on the fuel concentration is around 0.4. The initial droplet diameter is 60 macrons and the droplet temperature is 300 K. Evolutions of averaged temperature, species mass fraction, heat release and reaction rate are presented. Contours of temperature and species mass fractions are presented. The objective is to understand the mechanism of ignition under Homogeneous Charged Compression Ignition (HCCI) conditions, aiming at providing some useful information of HCCI combustion, which is one of the critical issues to be resolved.

  14. 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

  15. Mechanical behavior of irradiated fuel-pin cladding evaluated under transient heating and pressure conditions

    International Nuclear Information System (INIS)

    Hamilton, M.L.; Johnson, G.D.; Hunter, C.W.; Duncan, D.R.

    1982-11-01

    Fast breeder fuel-pin cladding has been tested under experimental conditions simulating the temperature and pressure history characteristic of anticipated transient events. Irradiation induces severe reductions in both strength and ductility. Ductility losses are independent of the rate of temperature increase and saturate by a fluence of approx. 2 x 10 22 n/cm 2 (E > 0.1 MeV). Losses in strength are dependent on the rate of temperature increase but saturate at a fluence of approx.5 x 10 22 n/cm 2 . Evidence is presented to show that fission products are probably responsible for the degradation in mechanical properties

  16. Modelling of WWER fuel rod during LOCA conditions using FEM code ANSYS

    International Nuclear Information System (INIS)

    Bogatyr, S. M.; Krupkin, A. V.; Kuznetsov, V. I.; Novikov, V. V.; Petrov, O. M.; Shestopalov, A. A.

    2013-01-01

    The report presents the results of the computer simulation of the IFA-650.6 experiment, the sixth test in Halden LOCA test project series, performed in May 18, 2007 with a pre-irradiated WWER-440 fuel with maximum burnup of 56 MWd/kgU. The thermo-mechanical analysis was fulfilled with the license finite element ANSYS code package.The calculation was carried out with the 2D axisymmetric and 3D problem definitions. Analysis of the calculational results shows that the ANSYS code can adequately simulate thermo-mechanical behavior of cladding under IFA-650.6 test conditions. (authors)

  17. Increased power generation from primary sludge by a submersible microbial fuel cell and optimum operational conditions

    DEFF Research Database (Denmark)

    Vologni, Valentina; Kakarla, Ramesh; Angelidaki, Irini

    2013-01-01

    Microbial fuel cells (MFCs) have received attention as a promising renewable energy technology for waste treatment and energy recovery. We tested a submersible MFC with an innovative design capable of generating a stable voltage of 0.250 ± 0.008 V (with a fixed 470 Ω resistor) directly from prima...... prolonged the current generation and increased the power density by 7 and 1.5 times, respectively, in comparison with raw primary sludge. These findings suggest that energy recovery from primary sludge can be maximized using an advanced MFC system with optimum conditions....

  18. Numerical studies of the heat-up-phase of Super-Sara 'severe fuel damage'. Boildown tests

    International Nuclear Information System (INIS)

    Eifler, W.; Shepherd, I.M.

    1983-01-01

    Calculations to investigate the heat-up phase of the Super-Sara 'severe fuel damage' test matrix have been performed using a simple computer code which models a typical pin. In particular the effect of the exothermic zirconium water reaction on the transient is considered. It is shown that it is possible to achieve the desired objectives of all the tests by a test procedure involving a constant power level a simple flow history. This flow history consists of an initial inlet flow, that has the water saturated at outlet. It is then linearly decreased in a time of the order of 200 seconds to a steady lower value. The clad temperature ramp rate is defined by the power and the peak clad temperature by the ratio of the power of the final steady inlet flow rate. If the final inlet flow rate for a particular power is below a certain critical value then the clad will reach melting temperature. The sensitivity of the results are discussed and a sample calculation is made for each test in the matrix

  19. Analysis of scenarios for the direct disposal of spent nuclear fuel disposal conditions as expected in Germany

    International Nuclear Information System (INIS)

    Ashton, P.; Mehling, O.; Mohn, R.; Wingender, H.J.

    1990-01-01

    This report contains an investigation of aspects of the waste management of spent light water reactor fuel by direct disposal in a deep geological formation on land. The areas covered are: interim dry storage of spent fuel with three options of pre-conditioning; conditioning of spent fuel for final disposal in a salt dome repository; disposal of spent fuel (heat-generating waste) in a salt dome repository; disposal of medium and low-level radioactive wastes in the Konrad mine. Dose commitments, effluent discharges and potential incidents were not found to vary significantly for the various conditioning options/salt dome repository types. Due to uncertainty in the cost estimates, in particular the disposal cost estimates, the variation between the three conditioning options examined is not considered as being significant. The specific total costs for the direct disposal strategy are estimated to lie in the range ECU 600 to 700 per kg hm (basis 1988)

  20. Production and processing of fuel by the forest industry - opportunities and conditions

    International Nuclear Information System (INIS)

    Magnusson, L.

    1991-01-01

    The purpose of this study was to illustrate the opportunities for the forest industry to establish a system of handling and processing biofuels in conjunction with their existing activities, and which would supply a future market for biofuels in, for example, electricity generation. The sawmills report that it is difficult today to find a market for fuel products, especially for sawmills at greater distances from larger biofuel-consuming plants. The sawmills show great interest in cogeneration in their own plants, but report difficulties in achieving profitability. The main problem is reported to be that the price of the surplus electricity delivered to the grid is too low, but also that the electricity prices today are so low that it is difficult to justify even generating electricity for the mill's own use. There is an interest in the paper and pulp industry for integrated methods and production of biofuels since the part-tree methods used, at least in some parts of Sweden, are considered to contribute also to an increase in the availability of pulp wood to the industry. A fundamental viewpoint is, however, that the plants are built for the primary purpose of producing pulp or paper. It is unlikely that the industry would give priority to investments for production of a new secondary product in the form of fuel products, particularly when the conditions today imply that there are few possibilities to achieve any particular profitability. The most probable solution is that the fuel is processed outside the industry by other parties, e.g., the forest divisions. In the long term, increased efficiency in the processes may lead to a primary heat surplus which could be used to produce processed fuels

  1. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Greg John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Engelhardt, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-05-01

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  2. Surrogate fuel assembly multi-axis shaker tests to simulate normal conditions of rail and truck transport

    Energy Technology Data Exchange (ETDEWEB)

    McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Koenig, Greg John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Uncapher, William Leonard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Grey, Carissa [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Engelhardt, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saltzstein, Sylvia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sorenson, Ken B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-05-12

    This report describes the third set of tests (the “DCLa shaker tests”) of an instrumented surrogate PWR fuel assembly. The purpose of this set of tests was to measure strains and accelerations on Zircaloy-4 fuel rods when the PWR assembly was subjected to rail and truck loadings simulating normal conditions of transport when affixed to a multi-axis shaker. This is the first set of tests of the assembly simulating rail normal conditions of transport.

  3. Secoisolariciresinol diglucoside abrogates oxidative stress-induced damage in cardiac iron overload condition.

    Directory of Open Access Journals (Sweden)

    Stephanie Puukila

    Full Text Available Cardiac iron overload is directly associated with cardiac dysfunction and can ultimately lead to heart failure. This study examined the effect of secoisolariciresinol diglucoside (SDG, a component of flaxseed, on iron overload induced cardiac damage by evaluating oxidative stress, inflammation and apoptosis in H9c2 cardiomyocytes. Cells were incubated with 50 μ5M iron for 24 hours and/or a 24 hour pre-treatment of 500 μ M SDG. Cardiac iron overload resulted in increased oxidative stress and gene expression of the inflammatory mediators tumor necrosis factor-α, interleukin-10 and interferon γ, as well as matrix metalloproteinases-2 and -9. Increased apoptosis was evident by increased active caspase 3/7 activity and increased protein expression of Forkhead box O3a, caspase 3 and Bax. Cardiac iron overload also resulted in increased protein expression of p70S6 Kinase 1 and decreased expression of AMP-activated protein kinase. Pre-treatment with SDG abrogated the iron-induced increases in oxidative stress, inflammation and apoptosis, as well as the increased p70S6 Kinase 1 and decreased AMP-activated protein kinase expression. The decrease in superoxide dismutase activity by iron treatment was prevented by pre-treatment with SDG in the presence of iron. Based on these findings we conclude that SDG was cytoprotective in an in vitro model of iron overload induced redox-inflammatory damage, suggesting a novel potential role for SDG in cardiac iron overload.

  4. Leaching Studies on ACR-1000{sup R} Fuel Under Reactor Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Sunder, S. [Atomic Energy of Canada Limited, Fuel and Fuel Channel Safety Branch, Chalk River, Ontario, K0J 1J0 (Canada)

    2009-06-15

    ACR-1000{sup R} is the latest nuclear power reactor being developed by AECL. The ACR-1000 fuel uses a modified CANFLEX{sup R} fuel bundle that contains low-enriched uranium and pellets of burnable neutron absorbers (BNA) in a central element. Dysprosium and gadolinium are used as the burnable neutron absorbers and are present as oxides in a 'fully-stabilized' zirconia matrix. The BNA material in the centre element is designed to limit the coolant void reactivity of the reactor core during postulated loss-of-coolant accidents. As part of the ACR-1000 fuel development, the stability of the BNA material under conditions associated with defects of the Zircaloy sheathing of the BNA central element has been investigated. The results of these tests can be used to demonstrate the phase stability and leaching behaviour of the ACR-1000 fuel under reactor operating conditions. The samples were disks, about 3-4 mm thick, obtained from BNA pellets and Candu fuel (natural uranium UO{sub 2}) pellets (the UO{sub 2} measurements provide a reference point). Leaching tests were carried out in light water at 325 deg. C, above the maximum coolant temperature in an ACR-1000 fuel channel during normal operating conditions (319 deg. C). This temperature also bounds the maximum operating temperature for the current Candu reactors (311 deg. C). The initial pH of the solution (measured at room temperature) used in the leaching tests was 10.3. The leach rates were determined by monitoring the amount of metals leached into solutions. Leaching tests were also carried out with BNA pellet samples in the presence of Zr-2.5%Nb pressure tube coupons to determine the effects, if any, of the presence of pressure tube material on leach rates. Other leaching tests with BNA pellet samples and UO{sub 2} pellets were conducted at 80 deg. C to study the effects of temperature on the leach rates. The temperature of 80 deg. C was selected as representative of typical shutdown temperatures

  5. Influence of radiolysis on UO2 fuel matrix dissolution under disposal conditions. Literature Study

    International Nuclear Information System (INIS)

    Ollila, K.

    2011-05-01

    The objective of this study was to examine the recent published literature on the influence of water radiolysis on UO 2 fuel matrix dissolution under the disposal conditions. The α radiation is considered to be dominating over the other types of radiations at times longer than 1000 years. The presence of the anaerobic corrosion products of iron, especially of hydrogen, has been observed to play an important role under radiolysis conditions. It is not possible to exclude gamma/beta radiolysis effects in the experiments with spent fuel, since there is not available a fuel over 100 years old. More direct measurements of α radiolysis effects have been conducted with α doped UO 2 materials. On the basis of the results of these experiments, a specific activity threshold to observe α radiolysis effects has been presented. The threshold is 1.8 x 10 7 to 3.3 x 10 7 Bq/g in anoxic 10 -3 M carbonate solution. It is dependent on the environmental conditions, such as the reducing buffer capacity of the conditions. The results of dissolution rate measurements at VTT with 233 U-doped UO 2 samples in 0.01 to 0.1 M NaCl solutions under anoxic conditions did not show any effect of α radiolysis with doping levels of 5 and 10% 233 U (3.2 x 10 7 and 6.3 x 10 7 Bq/g). Both Fe 2+ and hydrogen can act as reducing species and could react with oxidizing radiolytic species. Fe 2+ concentrations of the order of 10 -5 M can decrease the rate of H 2 O 2 production. Low dissolution rates, 2 x 10 -8 to 2 x 10 -7 /yr, have been measured in the presence of metallic Fe with 5 and 10% 233 U-doped UO 2 in 0.01 to 1 M NaCl solutions. The tests with isotope dilution method showed precipitation phenomena of U to occur during dissolution process. The concentrations of dissolved U were extremely low (≤ 8.4 x 10 -11 M). No effects of -radiolysis could be seen. It is difficult to distinguish the effects of metallic Fe, Fe 2+ or hydrogen in these tests. Hydrogen could also act as a reducing agent

  6. FARST: A computer code for the evaluation of FBR fuel rod behavior under steady-state/transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Sakagami, M.

    1984-01-01

    FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows: (I) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod. (II) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method. (III) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions. (IV) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as 'jump relocation model'. The code was successfully applied to analyses of the fuel rod irradiation data from pulse reactor for nuclear safety research in Cadarache (CABRI) and pulse reactor for nuclear safety research in Japan Atomic Energy Research Institute (NSRR). The code was further applied to stress analysis of a 1000 MW class large FBR plant fuel rod during transient conditions. The steady-state model which was used so far gave the conservative results for cladding stress during overpower transient, but underestimated the results for cladding stress during a rapid temperature decrease of coolant. (orig.)

  7. Analysis of a burning fuel on a water sublayer: conditions of triggering mechanism of superheated water explosion ('boilover')

    International Nuclear Information System (INIS)

    Jordan Y Hristov

    2005-01-01

    Full text of publication follows: The communication considers the burning of fuel on water sublayer that commonly occurs during tanks fires of combustible liquids. The main efforts are stressed on the qualitative assessments of the heat transfer mechanisms and the prediction of the boilover onset. Boilover is generally considered as one of the most dangerous fire phenomena. Fires in storage plants can and still do happen and cause severe damage and high losses. The boilover phenomenon is attractive from a fundamental point of view that address to better understanding of its mechanism and theoretical prediction of the critical condition of its onset. The analysis employed various data obtained by different research groups all over the world [1-5]. The evaluation of the suitable functional relationships predicting the pre-boilover time was done on the basis of dimensionless forms of two types of single layer heat conduction models: Surface absorption models [2,3,5] and In-depth absorption models [1,2,5]. Dimensional analysis of the models has detected several dimensionless numbers allowing easy buildup of similarity regression models predicting the pre-boilover time (critical Fourier number correlations) [ 5].The present work continues the study already started on the unified analysis of the boilover phenomenon [5] and the pre-explosion time prediction. The thermal conditions of the water sublayer are considered in order to evaluate the critical conditions for superheated water explosions. The latter have not been considered in the previous studies [1-5] due to both insufficient amount of data and incorrect interpretation of the phenomenon. REFERENCES: 1. Garo JP and Vantelon JP (1999) Thin layer boilover of pure or multicomponent fuel, in: Prevention of Hazardous Fires and Explosions. The transfer to Civil Applications of Military Experiences (Zarko V.E., Weiser V, Eisenreich N and Vasil'ev AA, Eds.), NATO Science Series, Series 1. Disarmament Technologies-vol. 26

  8. Bioavailability and biodegradation of weathered diesel fuel in aquifer material under denitrifying conditions

    International Nuclear Information System (INIS)

    Bregnard, T.P.A.; Hoehener, P.; Zeyer, J.

    1998-01-01

    During the in situ bioremediation of a diesel fuel-contaminated aquifer in Menziken, Switzerland, aquifer material containing weathered diesel fuel (WDF) and indigenous microorganisms was excavated. This material was used to identify factors limiting WDF biodegradation under denitrifying conditions. Incubations of this material for 360 to 390 d under denitrifying conditions resulted in degradation of 23% of the WDF with concomitant consumption of NO 3 - and production of inorganic carbon. The biodegradation of WDF and the rate of NO 3 - consumption was stimulated by agitation of the microcosms. Biodegradation was not stimulated by the addition of a biosurfactant (rhamnolipids) or a synthetic surfactant (Triton X-100) at concentrations above their critical micelle concentrations. The rhamnolipids were biodegraded preferentially to WDF, whereas Triton X-100 was not degraded. Both surfactants reduced the surface tension of the growth medium from 72 to <35 dynes/cm and enhanced the apparent aqueous solubility of the model hydrocarbon n-hexadecane by four orders of magnitude. Solvent-extracted WDF, added at a concentration equal to that already present in the aquifer material, was also biodegraded by the microcosms, but not at a higher rate than the WDF already present in the material. The results show that the denitrifying biodegradation of WDF is not necessarily limited by bioavailability but rather by the inherent recalcitrance of WDF

  9. 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

  10. Safety criterion for burnout of the plate-type fuel in pressurized conditions

    International Nuclear Information System (INIS)

    Komori, Y.; Kaminaga, M.; Sakurai, F.; Ando, H.; Sudo, Y.; Saito, M.; Futamura, Y.

    1992-01-01

    The reduced enrichment program for JMTR is now underway and the core conversion to LEU (Low Enrichment Uranium) is scheduled to be made in 1993. Consistent with the safety guide which have been recently developed for research and test reactors in Japan, the safety analysis for the JMTR LEU conversion was conducted. In the safety analysis, DNB (Departure from Nucleate Boiling) heat flux correlation for the JMTR downflow condition was reconsidered because recent studies on burnout show that DNB heat fluxes with thin rectangular channels under low flow rate and low pressure conditions are much lower than predicted values by conventional DNB correlations. Available DNB data, however, are very limited for the JMTR operation pressure range, so that DNB experiments were conducted simulating the JMTR fuel subchannel. Based mainly on the present experimental data, the DNB correlations scheme composed of three correlations was selected for the JMTR safety analysis. Errors of the correlations scheme with experimental data were evaluated in order to determine the allowable limit of the minimum DNB ratio for preventing fuel failure. (author)

  11. Methods for conditioning wastes from spent fuel cans and dissolver residues

    International Nuclear Information System (INIS)

    De Regge, P.; Loida, A.; Schmidt-Hansberg, T.; Sombret, C.

    1985-04-01

    Several methods for conditioning spent fuel decladding hulls or dissolver residues have been considered in various countries of the European Community. Five of these methods use embedding technique with or without prior compaction: they are based on incorporation in metallic alloys, glass, ceramics, cements and metals or graphite compounds. A sixth one consists in melting the decladding materials. The corresponding research programs have been pursued to varying states of progress with regard to demonstrating their feasibility on an industrial scale and the use of genuine wastes in bench scale experiments. The properties of the conditioned wastes have been investigated. Special attention has been paid to the corrosion resistance to various aqueous media as tap water, brine or clayey water. Although no categorical conclusion can be drawn from the initial results, the available finding provide a basis for assessing the different processes [fr

  12. Uraninite and its alternation at Palmottu - A possible natural analogue for spent fuel under reducing conditions

    International Nuclear Information System (INIS)

    Ruskeeniemi, T.; Blomqvist, R.; Ahonen, L.

    1994-01-01

    Uraninite is the major uraniferous mineral in the Precambrian U-Th rich pegmatites at Palmottu. Most euhedral uraninite grains were partially altered by silica-rich hydrothermal solutions during the late stage pegmatitic crystallization. The dominant secondary mineral is uranium silicate, with a chemical composition similar to that of coffinite (USiO 4 * Nh 2 O). The simultaneous formation of galena and other sulfides with the uranium silicate indicates that the alteration took place under reducing conditions. Hence, uranium occurs predominantly in the uranous (U 4+ ) state. Preliminary mass balance calculations imply that significant amounts of U, Th, and Pb were released during the replacement process. As the Palmottu U-Th deposit extends from ground level to distinctly reduced parts of the bedrock, it affords the opportinity of studying the stability and alteration of uraninite as an analogue for spent nuclear fuel under various redox conditions. (orig.) (28 refs., 5 figs., 1 tab.)

  13. Antioxidant capacity contributes to protection of ketone bodies against oxidative damage induced during hypoglycemic conditions.

    Science.gov (United States)

    Haces, María L; Hernández-Fonseca, Karla; Medina-Campos, Omar N; Montiel, Teresa; Pedraza-Chaverri, José; Massieu, Lourdes

    2008-05-01

    Ketone bodies play a key role in mammalian energy metabolism during the suckling period. Normally ketone bodies' blood concentration during adulthood is very low, although it can rise during starvation, an exogenous infusion or a ketogenic diet. Whenever ketone bodies' levels increase, their oxidation in the brain rises. For this reason they have been used as protective molecules against refractory epilepsy and in experimental models of ischemia and excitotoxicity. The mechanisms underlying the protective effect of these compounds are not completely understood. Here, we studied a possible antioxidant capacity of ketone bodies and whether it contributes to the protection against oxidative damage induced during hypoglycemia. We report for the first time the scavenging capacity of the ketone bodies, acetoacetate (AcAc) and both the physiological and non-physiological isomers of beta-hydroxybutyrate (D- and L-BHB, respectively), for diverse reactive oxygen species (ROS). Hydroxyl radicals (.OH) were effectively scavenged by D- and L-BHB. In addition, the three ketone bodies were able to reduce cell death and ROS production induced by the glycolysis inhibitor, iodoacetate (IOA), while only D-BHB and AcAc prevented neuronal ATP decline. Finally, in an in vivo model of insulin-induced hypoglycemia, the administration of D- or L-BHB, but not of AcAc, was able to prevent the hypoglycemia-induced increase in lipid peroxidation in the rat hippocampus. Our data suggest that the antioxidant capacity contributes to protection of ketone bodies against oxidative damage in in vitro and in vivo models associated with free radical production and energy impairment.

  14. Theoretical models to predict the transient heat transfer performance of HIFAR fuel elements under non-forced convective conditions

    International Nuclear Information System (INIS)

    Green, W.J.

    1987-04-01

    Simple theoretical models have been developed which are suitable for predicting the thermal responses of irradiated research fuel elements of markedly different geometries when they are subjected to loss-of-coolant accident conditions. These models have been used to calculate temperature responses corresponding to various non-forced convective conditions. Comparisons between experimentally observed temperatures and calculated values have shown that a suitable value for surface thermal emissivity is 0.35; modelling of the fuel element beyond the region of the fuel plate needs to be included since these areas account for approximately 25 per cent of the thermal power dissipated; general agreement between calculated and experimental temperatures for both transient and steady-state conditions is good - the maximum discrepancy between calculated and experimental temperatures for a HIFAR Mark IV/V fuel element is ∼ 70 deg C, and for an Oak Ridge Reactor (ORR) box-type fuel element ∼ 30 deg C; and axial power distribution does not significantly affect thermal responses for the conditions investigated. Overall, the comparisons have shown that the models evolved can reproduce experimental data to a level of accuracy that provides confidence in the modelling technique and the postulated heat dissipation mechanisms, and that these models can be used to predict thermal responses of fuel elements in accident conditions that are not easily investigated experimentally

  15. 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)

  16. Fretting Wear Damage Mechanism of Uranium under Various Atmosphere and Vacuum Conditions

    Directory of Open Access Journals (Sweden)

    Zhengyang Li

    2018-04-01

    Full Text Available A fretting wear experiment with uranium has been performed on a linear reciprocating tribometer with ball-on-disk contact. This study focused on the fretting behavior of the uranium under different atmospheres (Ar, Air (21% O2 + 78% N2, and O2 and vacuum conditions (1.05 and 1 × 10−4 Pa. Evolution of friction was assessed by coefficient of friction (COF and friction-dissipated energy. The oxide of the wear surface was evaluated by Raman spectroscopy. The result shows that fretting wear behavior presents strong atmosphere and vacuum condition dependence. With increasing oxygen content, the COF decreases due to abrasive wear and formation of oxide film. The COF in the oxygen condition is at least 0.335, and it has a maximum wear volume of about 1.48 × 107 μm3. However, the COF in a high vacuum condition is maximum about 1.104, and the wear volume is 1.64 × 106 μm3. The COF in the low vacuum condition is very different: it firstly increased and then decreased rapidly to a steady value. It is caused by slight abrasive wear and the formation of tribofilm after thousands of cycles.

  17. Improving the monitoring of quantitative conditions of peacetime fuel stocks at pumping stations

    Directory of Open Access Journals (Sweden)

    Slaviša M. Ilić

    2011-04-01

    human resources. Optimization of quantitative monitoring of peacetime supplies of fuel at gas stations should aim at reducing the impact of the human factor, introducing automated quantitative monitoring of fuel condition with modern equipment for handling as well as applying technology for fast reading and dissemination of information and reports. Civilian pumping stations have been modernized gradually with new digital pump machines, systems for automated production and automated systems for measuring the fuel level in buried tanks. The objectives and criteria of the optimization of model monitoring In order to solve the problem of multi-criteria nature, the methods of operational research have been applied and the formalization of problem solving has been carried out. Models have been identified, criteria and subcriteria have been defined as well as respective criteria values, sub-criteria and weight coefficients for chosen variants in order to rank the alternatives - models. On the basis of the defined objectives and optimization approaches, the task of optimization to be solved is to choose one optimal model of monitoring the quantitative condition of peacetime stocks of fuels at gas stations, out of three variations or alternative models. Application of expert assessment and methods of analytical hierarchy process The problem was solved first 'manually', by using MS Excell, and after that by using the Expert Choice software package. The Expert Choice software package is based on the application of the method of analytical hierarchy process and combines the benefits that this method offers with the speed and visibility of computerized calculations and their result display. The purpose of the AHP method is to rank alternative decisions by their importance and to select the most acceptable alternative on the basis of a defined set of criteria and alternatives. The problem of determining the weight of criteria has been determined by applying the method of expert

  18. Irradiated fuel behavior under accident heating conditions and correlation with fission gas release and swelling model (Chicago)

    International Nuclear Information System (INIS)

    Kryger, B.; Ducamp, F.; Combette, P.

    1981-08-01

    We analyse the mixed oxide fast fuel response to off normal conditions obtained by means of an out-of-pile transient simulation apparatus designed to provide direct observations (temperature, pressure, fuel motion) of fuel fission gas phenomena that might occur during the transients. The results are concerning fast transient tests (0,1 to 1 second) obtained with high gas concentration irradiated fuel (4 to 7 at % burn up, 0,4 cm 3 Xe + Kr /g.UPuO 2 ). The kinetics of fission gas release during the transients have been directly measured and then compared with the calculated results issued of the Chicago model. This model agrees, quite well, with other experiments done in the silene prompt reactor. Other gases than xenon and krypton (such as hydrogen and carbon monoxide) do not play any role in fuel behavior, since they have been completely ruled out

  19. Corrosion of titanium and titanium alloys in spent fuel repository conditions - literature review

    International Nuclear Information System (INIS)

    Aho-Mantila, I.; Haenninen, H.; Aaltonen, P.; Taehtinen, S.

    1985-03-01

    The spent nuclear fuel is planned to be disposed in Finnish bedrock. The canister of spent fuel in waste repository is one barrier to the release of radionuclides. It is possible to choose a canister material with a known, measurable corrosion rate and to make it with thickness allowing corrosion to occur. The other possibility is to use a material which is nearly immune to general corrosion. In this second category there are titanium and titanium alloys which exhibit a very high degree of resistance to general corrosion. In this literature study the corrosion properties of unalloyed titanium, titanium alloyed with palladium and titanium alloyed with molybdenum and nickel are reviewed. The two titanium alloys own in addition to the excellent general corrosion properties outstanding properties against localized corrosion like pitting or crevice corrosion. Stress corrosion cracking and corrosion fatique of titanium seem not to be a problem in the repository conditions, but the possibilities of delayed cracking caused by hydrogen should be carefully appreciated. (author)

  20. Chemical and mineralogical aspects of water-bentonite interaction in nuclear fuel disposal conditions

    International Nuclear Information System (INIS)

    Melamed, A.; Pitkaenen, P.

    1996-01-01

    In the field of nuclear fuel disposal, bentonite has been selected as the principal sealing and buffer material for placement around waste canisters, forming both a mechanical and chemical barrier between the radioactive waste and the surrounding ground water. Ion exchange and mineral alteration processes were investigated in a laboratory study of the long-term interaction between compacted Na-bentonite (Volclay MX-80) and ground water solutions, conducted under simulated nuclear fuel disposal conditions. The possible alteration of montmorillonite into illite has been a major object of the mineralogical study. However, no analytical evidence was found, that would indicate the formation of this non-expandable clay type. Apparently, the change of montmorillonite from Na- to Ca-rich was found to be the major alteration process in bentonite. In the water, a concentration decrease in Ca, Mg, and K, and an increase in Na, HCO 3 and SO 4 were recorded. The amount of calcium ions available in the water was considered insufficient to account for the recorded formation of Ca-montmorillonite. It is therefore assumed that the accessory Ca-bearing minerals in bentonite provide the fundamental source of these cations, which exchange with sodium during the alteration process. (38 refs.)

  1. Antecedent conditions control carbon loss and downstream water quality from shallow, damaged peatlands.

    Science.gov (United States)

    Grand-Clement, E; Luscombe, D J; Anderson, K; Gatis, N; Benaud, P; Brazier, R E

    2014-09-15

    Losses of dissolved organic carbon (DOC) from drained peatlands are of concern, due to the effects this has on the delivery of ecosystem services, and especially on the long-term store of carbon and the provision of drinking water. Most studies have looked at the effect of drainage in deep peat; comparatively, little is known about the behaviour of shallow, climatically marginal peatlands. This study examines water quality (DOC, Abs(400), pH, E4/E6 and C/C) during rainfall events from such environments in the south west UK, in order to both quantify DOC losses, and understand their potential for restoration. Water samples were taken over a 19 month period from a range of drains within two different experimental catchments in Exmoor National Park; data were analysed on an event basis. DOC concentrations ranging between 4 and 21 mg L(-1) are substantially lower than measurements in deep peat, but remain problematic for the water treatment process. Dryness plays a critical role in controlling DOC concentrations and water quality, as observed through spatial and seasonal differences. Long-term changes in depth to water table (30 days before the event) are likely to impact on DOC production, whereas discharge becomes the main control over DOC transport at the time scale of the rainfall/runoff event. The role of temperature during events is attributed to an increase in the diffusion of DOC, and therefore its transport. Humification ratios (E4/E6) consistently below 5 indicate a predominance of complex humic acids, but increased decomposition during warmer summer months leads to a comparatively higher losses of fulvic acids. This work represents a significant contribution to the scientific understanding of the behaviour and functioning of shallow damaged peatlands in climatically marginal locations. The findings also provide a sound baseline knowledge to support research into the effects of landscape restoration in the future. Crown Copyright © 2014. Published by

  2. Investigation of Spiral Bevel Gear Condition Indicator Validation Via AC-29-2C Using Damage Progression Tests

    Science.gov (United States)

    Dempsey, Paula J.

    2014-01-01

    This report documents the results of spiral bevel gear rig tests performed under a NASA Space Act Agreement with the Federal Aviation Administration (FAA) to support validation and demonstration of rotorcraft Health and Usage Monitoring Systems (HUMS) for maintenance credits via FAA Advisory Circular (AC) 29-2C, Section MG-15, Airworthiness Approval of Rotorcraft (HUMS) (Ref. 1). The overarching goal of this work was to determine a method to validate condition indicators in the lab that better represent their response to faults in the field. Using existing in-service helicopter HUMS flight data from faulted spiral bevel gears as a "Case Study," to better understand the differences between both systems, and the availability of the NASA Glenn Spiral Bevel Gear Fatigue Rig, a plan was put in place to design, fabricate and test comparable gear sets with comparable failure modes within the constraints of the test rig. The research objectives of the rig tests were to evaluate the capability of detecting gear surface pitting fatigue and other generated failure modes on spiral bevel gear teeth using gear condition indicators currently used in fielded HUMS. Nineteen final design gear sets were tested. Tables were generated for each test, summarizing the failure modes observed on the gear teeth for each test during each inspection interval and color coded based on damage mode per inspection photos. Gear condition indicators (CI) Figure of Merit 4 (FM4), Root Mean Square (RMS), +/- 1 Sideband Index (SI1) and +/- 3 Sideband Index (SI3) were plotted along with rig operational parameters. Statistical tables of the means and standard deviations were calculated within inspection intervals for each CI. As testing progressed, it became clear that certain condition indicators were more sensitive to a specific component and failure mode. These tests were clustered together for further analysis. Maintenance actions during testing were also documented. Correlation coefficients were

  3. Housing conditions influence motor functions and exploratory behavior following focal damage of the rat brain.

    Science.gov (United States)

    Gornicka-Pawlak, Elzbieta; Jabłońska, Anna; Chyliński, Andrzej; Domańska-Janik, Krystyna

    2009-01-01

    The present study investigated influence of housing conditions on motor functions recovery and exploratory behavior following ouabain focal brain lesion in the rat. During 30 days post-surgery period rats were housed individually in standard cages (IS) or in groups in enriched environment (EE) and behaviorally tested. The EE lesioned rats showed enhanced recovery from motor impairments in walking beam task, comparing with IS animals. Contrarily, in the open field IS rats (both lesioned and control) traveled a longer distance, showed less habituation and spent less time resting at the home base than the EE animals. Unlike the EE lesioned animals, the lesioned IS rats, presented a tendency to hyperactivity in postinjury period. Turning tendency was significantly affected by unilateral brain lesion only in the EE rats. We can conclude that housing conditions distinctly affected the rat's behavior in classical laboratory tests.

  4. Fuel Performance Characterisation under Various PWR Conditions: Description of the Annealing Test Facilities available at the LECA-STAR laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Pontillon, Y.; Cornu, B.; Clement, S.; Ferroud-Plattet, M.P.; Malgouyres, P.P. [Commissariat a l' Energie Atomique, CEA/DEN/DEC/SA3C - Centre d' Etudes de Cadarache, BP1, 13108 Saint Paul Lez Durance (France)

    2008-07-01

    The aim to improve LWR fuel behaviour has led Cea to improve its post-irradiation examination capacities in term of test facilities and characterization techniques in the shielded hot cells of the LECA-STAR facility, located in Cadarache Cea center. as far as the annealing test facilities are concerned, fuel qualification and improvement of knowledge require a set of furnaces which are already used or will be used. The main characteristics of these furnaces strongly depend on the experimental objectives. The aim of this paper is to review the main aspects of these specific experiments concerning: (i) fission gas release from high burn up fuel, (ii) global fission product release in severe-accident conditions and (iii) fuel microstructural changes, potential cladding failure, radionuclide source terms... under conditions representative of long term dry storage and geological disposal. (authors)

  5. Experimental studies of the influence of fuel properties and operational conditions on stoking when combusting fuels in a fixed-bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Arias, Fabiana; Kolb, T.; Seifert, H.; Gehrmann, Hans-Joachim [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. for Technical Chemistry (ITC)

    2013-09-01

    Besides from knowledge about pollutant emission, knowledge of the combustion behavior of fuels plays a major role in the operation and optimization of combustion plants for waste and biomass. If the fuel is exchanged partly or completely in existing or newly designed grate-type combustion plants, adaptation of technical parameters is usually based on purely empirical studies. In the KLEAA fixed-bed reactor of KIT, Institute for Technical Chemistry (ITC), quantitative data on the combustion behavior can be determined from experimental investigations on the laboratory scale. Based on the characteristics obtained, the combustion behavior on a continuous grate can be estimated, This estimation is based on the assumption that no back mixing of the fuel occurs on the grate. Depending on the type of grate, however, stoking and back mixing play an important role. To improve the quality of the characteristics determined in KLEAA and enhance their transferability to the continuous process, it is necessary to determine the influence of fuel properties and operation conditions on stoking. Work is aimed at further developing the characteristics model taking into account a stoking factor describing the combustion behavior of a non-stoked fixed bed compared to a stoked fixed bed. The main task is to make a systematic study of the major parameters influencing stoking (e.g. stroke length, stroke frequency, geometry of the stoking unit, and fuel properties) in a fixed-bed reactor. The results shall be presented in the form of a semi-empirical equation. It is recommended to first study a model fuel, whose fuel properties are defined exactly and can be adjusted variably. Then, a stoking factor shall be derived from the studies. Possibly, a dimension analysis may be helpful. Finally, the results obtained are to be verified for residue-derived fuel. (orig.)

  6. Modeling hydrogen starvation conditions in proton-exchange membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Ohs, Jan Hendrik; Sauter, Ulrich; Maass, Sebastian [Robert Bosch GmbH, Robert-Bosch-Platz 1, 70839 Gerlingen-Schillerhoehe (Germany); Stolten, Detlef [Forschungszentrum Juelich GmbH, IEF-3: Fuel Cells, 52425 Juelich (Germany)

    2011-01-01

    In this study, a steady state and isothermal 2D-PEM fuel cell model is presented. By simulation of a single cell along the channel and in through-plane direction, its behaviour under hydrogen starvation due to nitrogen dilution is analysed. Under these conditions, carbon corrosion and water electrolysis are observed on the cathode side. This phenomenon, causing severe cell degradation, is known as reverse current decay mechanism in literature. Butler-Volmer equations are used to model the electrochemical reactions. In addition, we account for permeation of gases through the membrane and for the local water content within the membrane. The results show that the membrane potential locally drops in areas starved from hydrogen. This leads to potential gradients >1.2 V between electrode and membrane on the cathode side resulting in significant carbon corrosion and electrolysis reaction rates. The model enables the analysis of sub-stoichiometric states occurring during anode gas recirculation or load transients. (author)

  7. Conditioning of high activity solid waste: fuel claddings and dissolution residues

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    This chapter reports on experimental studies of embedding into matrix material, the melting and conversion of zircaloy, and waste properties and characterization. Methods are developed for embedding the waste scrap into a solid and resistant matrix material in order to confine the radioactivity and to prevent it from dispersion. The matrix materials investigated are lead alloys, ceramics and compacted graphite or aluminium powder. The treatment of fuel hulls by melting or chemical conversion is described. Cemented hulls are characterized and the pyrophoricity of zircaloy fines is investigated. Topics considered include the embedding of hulls into graphite and aluminium, the embedding of hulls and dissolution residues into alumino-ceramics, the solidification of alpha-bearing wastes into a ceramic matrix, and the conditioning of cladding waste by eutectoidic melting and by embedding in glass

  8. Performance of diagonal control structures at different operating conditions for polymer electrolyte membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Serra, Maria; Husar, Attila; Feroldi, Diego; Riera, Jordi [Institut de Robotica i Informatica Industrial, Universitat Politecnica de Catalunya, Consejo Superior de Investigaciones Cientificas, C. Llorens i Artigas 4, 08028 Barcelona (Spain)

    2006-08-25

    This work is focused on the selection of operating conditions in polymer electrolyte membrane fuel cells. It analyses efficiency and controllability aspects, which change from one operating point to another. Specifically, several operating points that deliver the same amount of net power are compared, and the comparison is done at different net power levels. The study is based on a complex non-linear model, which has been linearised at the selected operating points. Different linear analysis tools are applied to the linear models and results show important controllability differences between operating points. The performance of diagonal control structures with PI controllers at different operating points is also studied. A method for the tuning of the controllers is proposed and applied. The behaviour of the controlled system is simulated with the non-linear model. Conclusions indicate a possible trade-off between controllability and optimisation of hydrogen consumption. (author)

  9. The effects of fuel characteristics and engine operating conditions on the elemental composition of emissions from heavy duty diesel buses

    Energy Technology Data Exchange (ETDEWEB)

    M.C.H. Lim; G.A. Ayoko; L. Morawska; Z.D. Ristovski; E.R. Jayaratne [Queensland University of Technology, Brisbane, Qld. (Australia). International Laboratory for Air Quality and Health, School of Physical and Chemical Sciences

    2007-08-15

    The effects of fuel characteristics and engine operating conditions on elemental composition of emissions from twelve heavy duty diesel buses have been investigated. Two types of diesel fuels - low sulfur diesel (LSD) and ultra low sulfur diesel (ULSD) fuels with 500 ppm and 50 ppm sulfur contents respectively and 3 driving modes corresponding to 25%, 50% and 100% power were used. Elements present in the tailpipe emissions were quantified by inductively coupled plasma mass spectrometry (ICPMS) and those found in measurable quantities included Mg, Ca, Cr, Fe, Cu, Zn, Ti, Ni, Pb, Be, P, Se, Ti and Ge. Multivariate analyses using multi-criteria decision making methods (MCDM), principal component analysis (PCA) and partial least squares (PLS) facilitated the extraction of information about the structure of the data. MCDM showed that the emissions of the elements were strongly influenced by the engine driving conditions while the PCA loadings plots showed that the emission factors of the elements were correlated with those of other pollutants such as particle number, total suspended particles, CO, CO{sub 2} and NOx. Partial least square analysis revealed that the emission factors of the elements were strongly dependent on the fuel parameters such as the fuel sulfur content, fuel density, distillation point and cetane index. Strong correlations were also observed between these pollutants and the engine power or exhaust temperature. The study provides insights into the possible role of fuel sulfur content in the emission of inorganic elements from heavy duty diesel vehicles. 39 refs., 1 fig., 4 tabs.

  10. Speciation, behaviour, and fate of mercury under oxy-fuel combustion conditions

    International Nuclear Information System (INIS)

    Córdoba, Patricia; Maroto-Valer, M.; Delgado, Miguel Angel; Diego, Ruth; Font, Oriol; Querol, Xavier

    2016-01-01

    The work presented here reports the first study in which the speciation, behaviour and fate of mercury (Hg) have been evaluated under oxy-fuel combustion at the largest oxy-Pulverised Coal Combustion (oxy-PCC) demonstration plant to date during routine operating conditions and partial exhaust flue gas re-circulation to the boiler. The effect of the CO 2 -rich flue gas re-circulation on Hg has also been evaluated. Results reveal that oxy-PCC operational conditions play a significant role on Hg partitioning and fate because of the continuous CO 2 -rich flue gas re-circulations to the boiler. Mercury escapes from the cyclone in a gaseous form as Hg 2+ (68%) and it is the prevalent form in the CO 2 -rich exhaust flue gas (99%) with lower proportions of Hg 0 (1.3%). The overall retention rate for gaseous Hg is around 12%; Hg 0 is more prone to be retained (95%) while Hg 2+ shows a negative efficiency capture for the whole installation. The negative Hg 2+ capture efficiencies are due to the continuous CO 2 -rich exhaust flue gas recirculation to the boiler with enhanced Hg contents. Calculations revealed that 44 mg of Hg were re-circulated to the boiler as a result of 2183 re-circulations of CO 2 -rich flue gas. Especial attention must be paid to the role of the CO 2 -rich exhaust flue gas re-circulation to the boiler on the Hg enrichment in Fly Ashes (FAs). - Highlights: • The fate of gaseous Hg has been evaluated under oxy-fuel combustion. • The Hg oxidation process is enhanced in CO 2 -rich flue gas recirculation. • Hg 2+ is the prevalent gas species in the CO 2 -rich exhaust flue gas. • Hg 2+ (g) shows a negative efficiency capture for the whole installation. • Especial attention must be paid to the Hg enrichment in Fly Ashes.

  11. SSYST, Modular System for Transient Fuel Rod Behaviour Under Accident Condition

    International Nuclear Information System (INIS)

    Gulden, W.; Meyder, R.; Borgwaldt, H.

    1987-01-01

    1 - Description of problem or function: SSYST is a code system for analyzing transient fuel rod behaviour under off-normal conditions, developed jointly by the Institut fuer Kernenergetik und Energie-systeme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract for the Projekt Nukleare Sicherheit (PNS) at KfK. Main differences versus codes with similar applications are: (1) an open-ended modular code organisation; (2) a preference for simple models, wherever possible. While feature (1) makes SSYST a very flexible tool, easily adapted to changing requirements, feature (2) leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 minutes CPU time on IBM 3033, so that extensive parametric studies are feasible. Main differences between SSYST-3 and previous versions are related to a general clean-up of the code system, which reduces the implementation effort: - advanced modules for cladding deformation and oxidation and reflooding conditions are included; - an input processor thoroughly checks all input data

  12. Effects of boundary conditions on thermomechanical calculations: Spent fuel test - climax

    International Nuclear Information System (INIS)

    Butkovich, T.R.

    1982-10-01

    The effects of varying certain boundary conditions on the results of finite-element calculations were studied in relation to the Spent Fuel Test - Climax. The study employed a thermomechanical model with the ADINA structural analysis. Nodal temperature histories were generated with the compatible ADINAT heat flow codes. The boundary conditions studied included: (1) The effect of boundary loading on three progressively larger meshes. (2) Plane strain vs plane stress conditions. (3) The effect of isothermal boundaries on a small mesh and on a significantly larger mesh. The results showed that different mesh sizes had an insignificant effect on isothermal boundaries up to 5 y, while on the smallest and largest mesh, the maximum temperature difference in the mesh was 0 C. In the corresponding ADINA calculation, these different mesh sizes produce insignificant changes in the stress field and displacements in the region of interest near the heat sources and excavations. On the other hand, plane stress produces horizontal and vertical stress differences approx. 9% higher than does plane strain

  13. Reaction progress pathways for glass and spent fuel under unsaturated conditions

    International Nuclear Information System (INIS)

    Bates, J.; Finn, P.; Bourcier, W.; Stout, R.

    1994-10-01

    The source term for the release of radionuclides from a nuclear waste repository is the waste form. In order to assess the performance of the repository and the engineered barrier system (EBS) compared to regulations established by the Nuclear Regulatory Commission and the Environmental Protection Agency it is necessary (1) to use available data to place bounding limits on release rates from the EBS, and (2) to develop a mechanistic predictive model of the radionuclide release and validate the model against tests done under a variety of different potential reaction conditions. The problem with (1) is that there is little experience to use when evaluating waste form reaction under unsaturated conditions such that errors in applying expert judgment to the problem may be significant. The second approach, to test and model the waste form reaction, is a more defensible means of providing input to the prediction of radionuclide release. In this approach, information related to the source term has a technical basis and provides a starting point to make reasonable assumptions for long-term behavior. Key aspects of this approach are an understanding of the reaction progress mechanism and the ability to model the tests using a geochemical code such as EQ3/6. Current knowledge of glass, UO 2 , and spent fuel reactions under different conditions are described below

  14. Fuel pin behavior under slow ramp-type transient-overpower conditions in the CABRI-FAST experiments

    International Nuclear Information System (INIS)

    Fukano, Yoshitaka; Onoda, Yuichi; Sato, Ikken; Charpenel, Jean

    2009-01-01

    In the CABRI-FAST experimental program, four in-pile tests were performed with slow power-ramp-type transient-overpower conditions (called hereafter as 'slow TOP') to study transient fuel pin behavior under inadvertent control rod withdrawal events in liquid metal cooled fast breeder reactors. Annular-pellet fuel pins were used in three tests, while a solid-pellet fuel pin was used in the other test. All of these pins were pre-irradiated in Phenix. The slow TOP test with a solid-pellet fuel pin was realized as a comparatory test against an existing test (E12) in the CABRI-2 program. In the CABRI-FAST test (BCF1), a power ramp rate of 3% Po/s was applied, while in the CABRI-2 test, 1% Po/s was adopted. Moreover, overpower condition was maintained for a few seconds beyond the observed pin failure in the BCF1 test. In spite of the different power ramp rates, evaluated fuel thermal conditions at the observed failure time are quite similar. The continued overpower condition in the BCF1 test resulted in gradual degradation of the pin structure providing information effective for evaluation of various accident scenarios. Three slow TOP tests with the annular fuel in the CABRI-FAST program resulted in no pin failure showing high failure threshold. Based on post-test examination data and a theoretical evaluation, it was concluded that intra-pin free spaces, such as central hole, macroscopic cracks and fuel-cladding gap effectively mitigated fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in case of very large amount of fuel melting. Furthermore, such cavity pressurization was effectively mitigated by a molten-fuel squirting into the upper blanket region pushing the blanket pellets upward. These CABRI FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database with various fuel and transient conditions. (author)

  15. Kinetic and thermodynamic bases to resolve issues regarding conditioning of uranium metal fuels

    International Nuclear Information System (INIS)

    Johnson, A.B.; Ballinger, R.G.; Simpson, K.A.

    1994-12-01

    Numerous uranium - bearing fuels are corroding in fuel storage pools in several countries. At facilities where reprocessing is no longer available, dry storage is being evaluated to preclude aqueous corrosion that is ongoing. It is essential that thermodynamic and kinetic factors are accounted for in transitions of corroding uranium-bearing fuels to dry storage. This paper addresses a process that has been proposed to move Hanford N-Reactor fuel from wet storage to dry storage

  16. 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

  17. A Status of Art-Report on the Fission Products Behavior Released from Spent Fuel at High Temperature Conditions

    International Nuclear Information System (INIS)

    Park, Geun Il; Kim, J. H.; Lee, J. W.

    2003-04-01

    The experiments on the fission products release behavior from spent fuel at high temperature assuming reactor accident conditions have been carried out at Oak Ridge Nation Laboratory of USA in HI/VI tests, CEA of France in HEVA/VERCOS tests, AEA of England and CRNL of Canada in HOX test. The VEGA program to study the fission product release behavior from LWR irradiated fuel was recently initiated at JAERI. The key parameter affecting the fission product(FP) release behavior is temperature. In addition, other parameters such as fuel oxidation, burnup, pre-transient conditions are found to affect the FP releases considerably in the earlier tests. The atmosphere conditions such as oxidizing atmosphere (steam or air) or reducing atmosphere (hydrogen) can cause significant change of FPs release and transport behavior due to chemical forms of the reactive FPs which is dependent on the oxidation potential. The effect of fuel burnup on the Kr-85 or Cs-137 release showed that the release rates of these radionuclides increased with the increase of burnup, meaning that release rates are dominated by the atomic diffusions in the grains and they are primarily a function of temperature. However, the data on FPs release behavior using higher burnups above 50,000 MWD/MTU are not so many reported up to now. This report summarizes the test results of FPs release behavior in reactor accident conditions produced from other countries mentioned above. This review and analysis on earlier studies would be useful for predicting the release characteristics of FPs from domestic spent fuel. The release rates of fission gas or FPs from spent fuel at high temperature conditions during fabrication process of dry recycling fuel were also analyzed using many data obtained from earlier tests

  18. Tutorial review of spent-fuel degradation mechanisms under dry-storage conditions

    International Nuclear Information System (INIS)

    Einziger, R.E.

    1983-02-01

    This tutorial reviews our present understanding of fuel-rod degradation over a range of possible dry-storage environments. Three areas are covered: (1) why study fuel-rod degradation; (2) cladding-degradation mechanisms; and (3) the status of fuel-oxidation studies

  19. Fuel-disruption experiments under high-ramp-rate heating conditions

    International Nuclear Information System (INIS)

    Wright, S.A.; Worledge, D.H.; Cano, G.L.; Mast, P.K.; Briscoe, F.

    1983-10-01

    This topical report presents the preliminary results and analysis of the High Ramp Rate fuel-disruption experiment series. These experiments were performed in the Annular Core Research Reactor at Sandia National Laboratories to investigate the timing and mode of fuel disruption during the prompt-burst phase of a loss-of-flow accident. High-speed cinematography was used to observe the timing and mode of the fuel disruption in a stack of five fuel pellets. Of the four experiments discussed, one used fresh mixed-oxide fuel, and three used irradiated mixed-oxide fuel. Analysis of the experiments indicates that in all cases, the observed disruption occurred well before fuel-vapor pressure was high enough to cause the disruption. The disruption appeared as a rapid spray-like expansion and occurred near the onset of fuel melting in the irradiated-fuel experiments and near the time of complete fuel melting in the fresh-fuel experiment. This early occurrence of fuel disruption is significant because it can potentially lower the work-energy release resulting from a prompt-burst disassembly accident

  20. Pump Coupling & Motor bearing damage detection using Condition Monitoring at DTPS

    Science.gov (United States)

    Bari, H. M.; Deshpande, A. A.; Jalkote, P. S.; Patil, S. S.

    2012-05-01

    This paper shares a success story out of the implementation of Co-ordinated Condition Monitoring techniques at DTPS, wherein imminent Mis-alignment of HT auxiliary BFP - 1B and Motor bearing failure of ID FAN - 1B was diagnosed. On 30/12/2010, Booster Pump DE horizontal reading increased from 4.8 to 5.1 and then upto 5.9 mm/sec. It was suspected that Booster pump was mis-aligned with Motor. To confirm misalignment, Phase Analysis was also done which showed that Coupling phase difference was 180 Degrees. Vibration & Phase Analysis helped in diagnosing the exact root cause of abnormity in advance, saving plant from huge losses which could have caused total cost of £ 104,071. On 06/01/2011, ID fan 1B Motor NDE & DE horizontal vibration readings deviated from 0.5 to 0.8 and 0.6 to 0.8 mm/sec (RMS) respectively. Noise level increased from 99.1 to 101.9 db. It was suspected that Motor bearings had loosened over the shaft. Meanwhile, after opening of Motor, Inner race of NDE side was found cracked and loosened over the shaft. Vibration Analysis & Noise Monitoring helped in diagnosing the exact root cause of abnormity in advance, saving plant from huge losses which could have caused total cost of £ 308,857.

  1. An external peer review of the U.S. Department of Energy's assessment of ''damages and benefits of the fuel cycles: Estimation methods, impacts, and values''

    International Nuclear Information System (INIS)

    1993-01-01

    The need for better assessments of the ''external'' benefits and costs of environmental effects of various fuel cycles was identified during the development of the National Energy Strategy. The growing importance of this issue was emphasized by US Department of Energy (DOE) management because over half of the states were already pursuing some form of social costing in electricity regulation and a well-established technical basis for such decisions was lacking. This issue was identified as a major area of controversy--both scientifically and politically--in developing energy policies at the state and national level. In 1989, the DOE's Office of Domestic and International Energy Policy commissioned a study of the external environmental damages and benefits of the major fuel cycles involved in electric power generation. Over the next 3-year period, Oak Ridge National Laboratory and Resources for the Future conducted the study and produced a series of documents (fuel cycle documents) evaluating the costs of environmental damages of the coal, oil, natural gas, biomass, hydroelectric, and nuclear fuel cycles, as well as the Background Document on methodological issues. These documents described work that took almost 3 years and $2.5 million to complete and whose implications could be far reaching. In 1992, the Secretary of Energy sought advice on the overall concepts underlying the studies and the means employed to estimate environmental externalities. He asked the Secretary of Energy's Advisory Board to undertake a peer review of the fuel cycle studies and encouraged the Board to turn to outside expertise, as needed

  2. Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

    1979-05-01

    The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

  3. Flow behaviour in a CANDU horizontal fuel channel from stagnant subcooled initial conditions

    International Nuclear Information System (INIS)

    Caplan, M.Z.; Gulshani, P.; Holmes, R.W.; Wright, A.C.D.

    1984-01-01

    The flow behaviour in a CANDU primary system with horizontal fuel channels is described following a small inlet header break. With the primary pumps running, emergency coolant injection is in the forward direction so that the channel outlet feeders remain warmer than the inlet thereby promoting forward natural circulation. However, the break force opposes the forward driving force. Should the primary pumps run down after the circuit has refilled, there is a break size for which the natural circulation force is balanced by the break force and channels could, theoretically, stagnate. Result of visualization and of full-size channel tests on channel flow behaviour from an initially stagnant channel condition are discussed. After a channel stagnation, the decay power heats the coolant to saturation. Steam is then formed and the coolant stratifies. The steam expands into the subcooled water in the end fitting in a chugging type of flow regime due to steam condensation. After the end fitting reaches the saturation temperature, steam is able to penetrate into the vertical feeder thereby initiating a large buoyancy induced flow which refills the channel. The duration of stagnation is shown to be sensitive to small asymmetries in the initial conditions. A small initial flow can significantly shorten the occurrence and/or duration of boiling as has been confirmed by reactor experience. (author)

  4. 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

  5. DUPIC fuel compatibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

  6. DUPIC fuel compatibility assessment

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

  7. Effects of fabrication requirements on fuel performance in relation to operating conditions. The views of Electricite de France (EDF)

    International Nuclear Information System (INIS)

    Ponticq, M.; Richer, P.; Scribe, G.

    1979-01-01

    Because of the operational constraints relating to fuel behaviour, imperfect knowledge of the behaviour of a defective fuel assembly and, in the near future, the need to adapt reactor power to grid following (load following and remote control), EDF is aiming to reduce the present rate of fuel failure. While the phenomena affecting fuel behaviour have now been listed and analysed, the efforts at reducing their consequences have yet to be completed. This can be achieved, firstly, by reducing or eliminating fabrication defects, which are responsible for failure at the beginning of fuel life, through establishment of a good quality assurance organization and the search for still higher efficiency of quality control and fabrication equipment, and secondly, by developing fabrication techniques minimizing in particular cladding-pellet interactions and the stress corrosion of the cladding, which are responsible for fuel failure as from mid-life. However, the reactor operating conditions likely to apply in the near future may lead, for a given fuel configuration, to a re-evaluation of the fabrication parameters of cladding and UO 2 pellets. (author)

  8. Radionuclide release from spent fuel under geologic disposal conditions: An overview of experimental and theoretical work through 1985

    Energy Technology Data Exchange (ETDEWEB)

    Reimus, P.W.; Simonson, S.A.

    1988-04-01

    This report presents an overview of experimental and theoretical work on radionuclide release from spent fuel and uranium dioxide (UO/sub 2/) under geologic disposal conditions. The purpose of the report is to provide a source book of information that can be used to develop models that describe radionuclide release from spent fuel waste packages. Modeling activities of this nature will be conducted within the Waste Package Program (WPP) of the Department of Energy's Salt Repository Project (SRP). The topics discussed include experimental methods for investigating radionuclide release, how results have been reported from radionuclide release experiments, theoretical studies of UO/sub 2/ and actinide solubility, results of experimental studies of radionuclide release from spent fuel and UO/sub 2/ (i.e., the effects of different variables on radionuclide release), characteristics of spent fuel pertinent to radionuclide release, and status of modeling of radionuclide release from spent fuel. Appendix A presents tables of data from spent fuel radionuclide release experiments. These data have been digitized from graphs that appear in the literature. An annotated bibliography of literature on spent fuel characterization is provided in Appendix B.

  9. 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.

  10. The main conditions ensured problemless implementation of 235U high enriched fuel in Kozloduy NPP (Bulgaria) - WWER-1000 Units

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.; Minkova, K.; Michaylov, G.; Penev, P.; Gerchev, N.

    2009-01-01

    The collected water chemistry and radiochemistry data during the operation of the Kozloduy NPP Unit 5 for the period 2006-2009 (12-th, 13-th 14-th and 15-th fuel cycles) undoubtedly indicate for WWER-1000 Units (whose specific features are: Steam generators with austenitic stainless steel 08Cr18N10T tubing; Steam generators are with horizontal straight tubing and Fuel elements cladding material is Zr-1%Nb (Zr1Nb) alloy), that one realistic way for problemless implementation of 235 U high enriched fuel have been found. The main feature characteristics of this way are: Implementation of solid neutron burnable absorbers together with the dissolved in coolant neutron absorber - natural boric acid; Application of fuel cladding materials with enough corrosion resistance by the specific fuel cladding environment created by presence of SNB; Keeping of suitable coolant water chemistry which ensures low corrosion rates of core- and out-of-core- materials and limits in core (cladding) depositions and restricts out-of-core radioactivity buildup. The realization of this way in WWER-1000 Units in Kozloduy NPP was practically carried out through: 1) Implementation of Russian fuel assemblies TVSA which have as fuel cladding material E-110 alloy (Zr1Nb) with enough high corrosion resistance by presence of sub-cooled nucleate boiling (SNB) and use burnable absorber (Gd) integrated in the uranium-gadolinium (U-Gd 2 O 3 ) fuel (fuel rod with 5.0% Gd 2 O 3 ); 2) Development and implementation of water chemistry primary circuit guidelines, which require the relation between boric acid concentration and total alkalising agent concentrations to ensure coolant pH 300 = 7.0 - 7.2 values during the whole operation period. The above mentioned conditions by the passing of WWER-1000 Units in NPP Kozloduy to uranium fuel with 4.4% 235 U (TVSA fuel assemblies) practically ensured avoidance of the creation of the necessary conditions for AOA onset. The operational experience (2006-2009) of the

  11. Effect of zinc injection on BWR fuel cladding corrosion. Pt. 1. Study on an accelerated corrosion condition to evaluate corrosion resistance of zircaloy-2 fuel cladding

    International Nuclear Information System (INIS)

    Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

    2002-01-01

    Japanese BWR utilities have a plan to apply zinc injection to the primary coolant in order to reduce radioactivity accumulation on the structure. Prior to applying the zinc injection to BWR plants, it is necessary to evaluate the effect of zinc injection on corrosion resistance of fuel cladding. The objective of this report was to examine the accelerated corrosion condition for evaluation of BWR fuel cladding corrosion resistance under non-irradiated conditions, as the first step of a zinc injection evaluation study. A heat transfer corrosion test facility, in which a two phase flow condition could be achieved, was designed and constructed. The effects of heat flux, void fraction and solution temperature on BWR fuel cladding corrosion resistance were quantitatively investigated. The main findings were as follows. (1) In situ measurements using high speed camera and a void sensor together with one dimensional two phase flow analysis results showed that a two phase flow simulated BWR core condition can be obtained in the corrosion test facility. (2) The heat transfer corrosion test results showed that the thickness of the zirconium oxide layer increased with increasing solution temperature and was independent of heat flux and void fraction. The corrosion accelerating factor was about 2.5 times in the case of a temperature increase from 288degC to 350degC. (author)

  12. Modeling of a Membrane Based Humidifier for Fuel Cell Applications Subject to End-Of-Life Conditions

    DEFF Research Database (Denmark)

    Nielsen, Mads Pagh; Olesen, Anders Christian; Menard, Alan

    2014-01-01

    applications. For instance for automotive applications and various backup power systems substituting batteries. Humidification of the inlet air of PEM fuel cell stacks is essential to obtain optimum proton conductivity. Operational humidities of the anode and cathode streams having dew points close to the fuel......Proton Exchange Membrane (PEM) Fuel Cell Stacks efficiently convert the chemical energy in hydrogen to electricity through electrochemical reactions occurring on either side of a proton conducting electrolyte. This is a promising and very robust energy conversion process which can be used in many...... cell operating temperature are required. These conditions must be met at the Beginning-Of-Life (BOL) as well as at the End-Of-Life (EOL) of the fuel cell system. This paper presents results of a numerical 1D model of the heat- and mass transport phenomena in a membrane humidifier with a Nafion...

  13. Damage detection methodology under variable load conditions based on strain field pattern recognition using FBGs, nonlinear principal component analysis, and clustering techniques

    Science.gov (United States)

    Sierra-Pérez, Julián; Torres-Arredondo, M.-A.; Alvarez-Montoya, Joham

    2018-01-01

    Structural health monitoring consists of using sensors integrated within structures together with algorithms to perform load monitoring, damage detection, damage location, damage size and severity, and prognosis. One possibility is to use strain sensors to infer structural integrity by comparing patterns in the strain field between the pristine and damaged conditions. In previous works, the authors have demonstrated that it is possible to detect small defects based on strain field pattern recognition by using robust machine learning techniques. They have focused on methodologies based on principal component analysis (PCA) and on the development of several unfolding and standardization techniques, which allow dealing with multiple load conditions. However, before a real implementation of this approach in engineering structures, changes in the strain field due to conditions different from damage occurrence need to be isolated. Since load conditions may vary in most engineering structures and promote significant changes in the strain field, it is necessary to implement novel techniques for uncoupling such changes from those produced by damage occurrence. A damage detection methodology based on optimal baseline selection (OBS) by means of clustering techniques is presented. The methodology includes the use of hierarchical nonlinear PCA as a nonlinear modeling technique in conjunction with Q and nonlinear-T 2 damage indices. The methodology is experimentally validated using strain measurements obtained by 32 fiber Bragg grating sensors bonded to an aluminum beam under dynamic bending loads and simultaneously submitted to variations in its pitch angle. The results demonstrated the capability of the methodology for clustering data according to 13 different load conditions (pitch angles), performing the OBS and detecting six different damages induced in a cumulative way. The proposed methodology showed a true positive rate of 100% and a false positive rate of 1.28% for a

  14. Assessment of the State of the Art of Integrated Vehicle Health Management Technologies as Applicable to Damage Conditions

    Science.gov (United States)

    Reveley, Mary S.; Kurtoglu, Tolga; Leone, Karen M.; Briggs, Jeffrey L.; Withrow, Colleen A.

    2010-01-01

    A survey of literature from academia, industry, and other Government agencies assessed the state of the art in current integrated vehicle health management (IVHM) aircraft technologies. These are the technologies that are used for assessing vehicle health at the system and subsystem level. This study reports on how these technologies are employed by major military and commercial platforms for detection, diagnosis, prognosis, and mitigation. Over 200 papers from five conferences from the time period of 2004 to 2009 were reviewed. Over 30 of these IVHM technologies are then mapped into the 17 different adverse event damage conditions identified in a previous study. This study illustrates existing gaps and opportunities for additional research by the NASA IVHM Project.

  15. Influence of implantation conditions of He+ ions on the structure of a damaged layer in GaAs(001)

    International Nuclear Information System (INIS)

    Shcherbachev, Kirill; Bailey, Melanie J.

    2011-01-01

    An investigation into the influence of implantation conditions (dose, energy, and target temperature) of He + ions on the damage structure of GaAs (100) substrates was performed by HRXRD, scanning electron microscopy, and Nomarski microscopy. Blistering is shown to become apparent as characteristic features of isolines in RSMs. We propose that the formation of the defects yielding a characteristic XRDS is defined by the behavior of implanted atoms in the GaAs matrix, depending on two competing processes: (1) formation of the gas-filled bubbles; (2) diffusion of the He atoms from the bubbles toward the surface and deep into the GaAs substrate. We conclude that the gas-filled bubbles change the structure of the irradiated layer, resulting in the formation of strained crystalline areas of the GaAs matrix. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  16. Effect of Fuel on Performance of a Single Combustor of an I-16 Turbojet Engine at Simulated Altitude Conditions

    Science.gov (United States)

    Zettle, Eugene V; Bolz, Ray E; Dittrich, R T

    1947-01-01

    As part of a study of the effects of fuel composition on the combustor performance of a turbojet engine, an investigation was made in a single I-16 combustor with the standard I-16 injection nozzle, supplied by the engine manufacturer, at simulated altitude conditions. The 10 fuels investigated included hydrocarbons of the paraffin olefin, naphthene, and aromatic classes having a boiling range from 113 degrees to 655 degrees F. They were hot-acid octane, diisobutylene, methylcyclohexane, benzene, xylene, 62-octane gasoline, kerosene, solvent 2, and Diesel fuel oil. The fuels were tested at combustor conditions simulating I-16 turbojet operation at an altitude of 45,000 feet and at a rotor speed of 12,200 rpm. At these conditions the combustor-inlet air temperature, static pressure, and velocity were 60 degrees F., 12.3 inches of mercury absolute, and 112 feet per second respectively, and were held approximately constant for the investigation. The reproducibility of the data is shown by check runs taken each day during the investigation. The combustion in the exhaust elbow was visually observed for each fuel investigated.

  17. Speciation, behaviour, and fate of mercury under oxy-fuel combustion conditions

    Energy Technology Data Exchange (ETDEWEB)

    Córdoba, Patricia, E-mail: pc247@hw.ac.uk [Centre for Innovation on Carbon Capture and Storage (CICCS), Institute of Mechanical, Process and Energy Engineering (IMPEE), Heriot-Watt University, EH14 4AS (United Kingdom); Maroto-Valer, M. [Centre for Innovation on Carbon Capture and Storage (CICCS), Institute of Mechanical, Process and Energy Engineering (IMPEE), Heriot-Watt University, EH14 4AS (United Kingdom); Delgado, Miguel Angel; Diego,