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Sample records for ex-vessel corium coolability

  1. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

    2008-03-15

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

  2. Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

    International Nuclear Information System (INIS)

    Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R.

    2008-03-01

    This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to the

  3. Ex-vessel corium coolability sensitivity study with the CORQUENCH code

    International Nuclear Information System (INIS)

    Robb, Kevin; Corradini, Michael

    2009-01-01

    An unresolved safety issue for light water reactor beyond design basis accidents is the coolability and stabilization of ex-vessel core melt debris by top flooding. Several experimental programs, including the OECD MACE, MCCI-1, and the current MCCI-2 program, have investigated core-concrete interactions and debris cooling of ex-vessel core melts. As part of the OECD programs, the CORQUENCH computer model was developed based on phenomena identified from the experiments. Predictions by CORQUENCH have previously been compared against experiments and have also been extrapolated to reactor scale. The current study applied statistical techniques to investigate the importance of initial system parameters and cooling phenomena in CORQUENCH 3.01 on the accident progression of ex-vessel core melts. The purpose of this sensitivity study is to identify parameters that are of major importance, any code peculiarities over the range of inputs, and where modeling improvements may produce the most gain in prediction accuracy. The sensitivity studies were carried out over a range of input conditions, in 1-D and 2-D geometries, and for two concrete compositions. In terms of initial system parameters, the melt height had the most importance on concrete ablation and melt coolability. With respect to cooling phenomena, the amount of melt entrainment through the crust had the most importance on concrete ablation and melt coolability. (author)

  4. State-of-the-Art Report on Molten Corium Concrete Interaction and Ex-Vessel Molten Core Coolability

    International Nuclear Information System (INIS)

    Bonnet, Jean-Michel; Cranga, Michel; Vola, Didier; Marchetto, Cathy; Kissane, Martin; ); Robledo, Fernando; Farmer, Mitchel T.; Spengler, Claus; Basu, Sudhamay; Atkhen, Kresna; Fargette, Andre; Fisher, Manfred; Foit, Jerzi; Hotta, Akitoshi; Morita, Akinobu; Journeau, Christophe; Moiseenko, Evgeny; Polidoro, Franco; Zhou, Quan

    2017-01-01

    Activities carried out over the last three decades in relation to core-concrete interactions and melt coolability, as well as related containment failure modes, have significantly increased the level of understanding in this area. In a severe accident with little or no cooling of the reactor core, the residual decay heat in the fuel can cause the core materials to melt. One of the challenges in such cases is to determine the consequences of molten core materials causing a failure of the reactor pressure vessel. Molten corium will interact, for example, with structural concrete below the vessel. The reaction between corium and concrete, commonly referred to as MCCI (molten core concrete interaction), can be extensive and can release combustible gases. The cooling behaviour of ex-vessel melts through sprays or flooding is also complex. This report summarises the current state of the art on MCCI and melt coolability, and thus should be useful to specialists seeking to predict the consequences of severe accidents, to model developers for severe-accident computer codes and to designers of mitigation measures

  5. Status of the corquench model for calculation of ex-vessel corium coolability by an overlying water layer

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.

    2000-01-01

    The results of melt attack and coolability experiment (MACE) tests have identified several heat transfer mechanisms which could potentially lead to long term corium coolability. Based on physical observations from these tests, an integrated model of corium quenching (CORQUENCH) behavior is being developed. Aside from modeling of the primary physical processes observed in the tests, considerable effort has also been devoted to modeling of test occurrences which deviate from the behavior expected at reactor scale. In this manner, extrapolation of the models validated against the test data to the reactor case can be done with increased confidence. The integrated model currently addresses early bulk cooling and incipient crust formation heat transfer phases, as well as a follow-on water ingression phase which leads to development of a sustained quench front progressing downwards through the debris. In terms of experiment distortions, the model is also able to mechanistically calculate crust anchoring to the test section sidewalls, as well as the subsequent melt/crust separation phase which arises due to concrete densification upon melting. In this paper, the status of the model development and validation activities are described. In addition, representative calculations for PWR plant conditions are provided in order to illustrate the potential benefits of overlying water on mitigation of the accident sequence. (orig.)

  6. Analyses on ex-vessel debris formation and coolability in SARNET frame

    International Nuclear Information System (INIS)

    Pohlner, G.; Buck, M.; Meignen, R.; Kudinov, P.; Ma, W.; Polidoro, F.; Takasuo, E.

    2014-01-01

    Highlights: • Melt outflow varies from dripping melt outflow to molten corium jets of variable size. • Experiments show clear trend of producing particles in size range 2-4 mm. • Code calculations show complete solidification of particles, yielding formation of fragmented debris beds. • Limits of debris bed cooling and coolability margins are analysed. - Abstract: The major aim of work in the SARNET2 European project on ex-vessel debris formation and coolability was to get an overall perspective on coolability of melt released from a failed reactor pressure vessel and falling into a water-filled cavity. Especially, accident management concepts for BWRs, dealing with deep water pools below the reactor vessel, are addressed, but also shallower pools in existing PWRs, with questions about partial cooling and time delay of molten corium concrete interaction. The subject can be divided into three main topics: (i) Debris bed formation by breakup of melt, (ii) Coolability of debris and (iii) Coupled treatment of the processes. Accompanied by joint collaborations of the partners, the performed work comprises theoretical, experimental and modelling activities. Theoretical work was done by KTH on the melt outflow conditions from a RPV and on the quantification of the probability of yielding a non-coolable ex-vessel bed by use of probabilistic assessment. IKE introduced a theoretical concept to improve debris bed coolability. A large amount of experimental work was done by partners (KTH, VTT, IKE) on the coolability of debris beds using different bed geometries, particles, heating methods and water feeds, yielding a valuable base for code validation. Modelling work was mainly done by IKE, IRSN, RSE and VTT concerning jet breakup and/or debris bed formation and cooling in 2D and 3D geometries. A benchmark for the DEFOR-A experiment of KTH was performed. Important progress was reached for several tasks and aspects and important insights are given, enabling to focus the

  7. Importance of the in and ex-vessel corium coolability in case of severe accident for the French PWRs. Some views from L2 PSA and perspectives

    International Nuclear Information System (INIS)

    Raimond, E.; Caroli, C.; Meignen, R.; Rahni, N.; Laurent, B.

    2011-01-01

    In the case of a severe accident on a NPP leading to core degradation after a default in the core cooling as during the accident of Three Mile Island (TMI2), the most efficient way to stop the accident progression would be the in-vessel water injection if a specific mean is available. The TMI2 accident has shown that the accident can be stopped and that the corium, even if highly degraded, can be cooled, but no one can generalize the TMI2 accident termination to all situations. The present paper aims at presenting the situation for the French operated PWRs and is mainly based on the IRSN experience in level 2 probabilistic safety assessment (L2 PSA) development for this type of reactor. It tries to highlight the benefit that could be obtained from a better understanding of the corium cooling phenomenology, including both possible positive and negative effects. Three main negative effects of in-vessel flooding have to be taken into account in a L2 PSA for a PWR: an increase of the hydrogen production rate, a risk of in-vessel pressure increase and the development of conditions for steam explosion. L2 PSAs in France have now reached a certain maturity allowing raising some more precise issues, but for the issues presented in this paper, some progress from the research-development and the simulation tools (mainly the ASTEC integral code) are still necessary to support decision-making

  8. Physics of coolability of top flooded molten corium

    International Nuclear Information System (INIS)

    Kulkarni, P.P.; Singh, R.K.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    During a postulated severe accident in a nuclear reactor in case of ex-vessel scenario the molten corium can be relocated in the containment cavity forming a melt pool. In order to arrest further progression of severe accident, complete quenching of the molten corium pool is necessary. Most common way to deal with ex-vessel scenario is to flood the melt pool with large quantity of water. However, the mechanism of coolability is much more complex involving multi-component, multiphase heat, mass and momentum transfer. In this paper, a mechanistic model has been presented for the corium coolability under top flooding conditions. The model has been validated with the experimental data of COMECO test facility available in literature. Simulations have been carried out using the model to explore the physics behind the corium coolability with MCCI under top flooding condition. Variations in the thermo-physical properties as a result of MCCI have been considered and its effect on coolability has been studied. (author)

  9. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Dinh, T.N. [Royal Institute of Technology (Sweden)

    2007-04-15

    The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

  10. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

    International Nuclear Information System (INIS)

    Park, H.S.; Dinh, T.N.

    2007-04-01

    The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

  11. Multiphase flow in ex-vessel coolability: development of an innovative concept

    International Nuclear Information System (INIS)

    Corradini, Michael L.

    2006-01-01

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific Advanced Light Water Reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability

  12. Internal structure of an ex-vessel corium debris bed during severe accidents of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Park, Jin Ho; Moriyama, Kiyofumi; Park, Hyun Sun [POSTECH, Daejeon (Korea, Republic of)

    2015-10-15

    In the aspect of the coolability assessment the configuration of the debris bed, including internal and external characteristics, has significant importance as boundary conditions for simulations, however, relatively little investigation of the sedimentation process. For the development of a debris bed, recently there have been several studies that focused on thermal characteristics of corium particles. Yakush et al. performed simulation studies and showed that two phase natural convection affects the particle settling trajectory and changes the final arrival location of particles to result more flattened bed. Those simulation results have been supported by the experimental studies of Kim et al. using simulant particles and air bubble injection. For the internal structure of a debris bed, there have been several simulation and experimental studies, which investigated the effect of internal structure on debris bed coolability. Magallon has reported the particle size distribution at three elevations of the debris bed of FARO L-31 case, where the mean particle size was bigger for the lower elevation. However, there is a lack of detailed information on the characteristics of the debris bed, including the local structure and porosity. In this study, we investigated the internal structure of the debris bed using a mixture of stainless steel particles and air bubble injection. Local particle sedimentation quantity, particle size distribution change in radial direction and axial direction, and bed porosity was measured to investigate a relationship between the internal structure and the accident condition. An experimental investigation was carried out for the internal structure of ex-vessel corium debris bed in the flooded cavity during sever accident. Moderate corium discharge in high flooding level was assumed for full fragmentation of melt jet. The test particle mixture was prepared by following an empirical correlation, which reflects the particle size distribution of

  13. Internal structure of an ex-vessel corium debris bed during severe accidents of LWRs

    International Nuclear Information System (INIS)

    Kim, Eunho; Park, Jin Ho; Moriyama, Kiyofumi; Park, Hyun Sun

    2015-01-01

    In the aspect of the coolability assessment the configuration of the debris bed, including internal and external characteristics, has significant importance as boundary conditions for simulations, however, relatively little investigation of the sedimentation process. For the development of a debris bed, recently there have been several studies that focused on thermal characteristics of corium particles. Yakush et al. performed simulation studies and showed that two phase natural convection affects the particle settling trajectory and changes the final arrival location of particles to result more flattened bed. Those simulation results have been supported by the experimental studies of Kim et al. using simulant particles and air bubble injection. For the internal structure of a debris bed, there have been several simulation and experimental studies, which investigated the effect of internal structure on debris bed coolability. Magallon has reported the particle size distribution at three elevations of the debris bed of FARO L-31 case, where the mean particle size was bigger for the lower elevation. However, there is a lack of detailed information on the characteristics of the debris bed, including the local structure and porosity. In this study, we investigated the internal structure of the debris bed using a mixture of stainless steel particles and air bubble injection. Local particle sedimentation quantity, particle size distribution change in radial direction and axial direction, and bed porosity was measured to investigate a relationship between the internal structure and the accident condition. An experimental investigation was carried out for the internal structure of ex-vessel corium debris bed in the flooded cavity during sever accident. Moderate corium discharge in high flooding level was assumed for full fragmentation of melt jet. The test particle mixture was prepared by following an empirical correlation, which reflects the particle size distribution of

  14. Draft paper: On the analysis of diffusive mass transfer in ex-vessel corium pools

    International Nuclear Information System (INIS)

    Frolov, Kyrill N.

    2003-01-01

    In case of a severe accident at a nuclear power plant (NPP) involving the reactor pressure vessel (RPV) melt-through, confident solidification of ex-vessel corium is the imperative condition of its safe retention within the plant containment. The rate-determining process for solidification of ex-vessel coriums in the long-term is the chemical diffusion in the liquid phase at the solid-liquid interface. The process of chemical diffusion in the diffusive boundary layer can evolve taking on different rates, depending on the boundary conditions and the melt composition. Nonetheless, the chemical diffusion rates would entwine the self-diffusivities of corium constituents, which in turn would depend on the melt chemical composition. This work looks at some aspects of analytical and experimental determination of self-diffusivities of corium constituents. Following the corium-concrete interaction, an ex-vessel corium melt would contain several chemical components, including a fraction of silica. Accordingly, ex-vessel corium is considered in this paper as a silicate melts. In the realm of the geological and glass sciences, where silicate melts are most often discussed, the diffusive transport and viscous flow are conceived interrelated from a phenomenological point of view. Though the viscous and diffusive mass transfer mechanisms are not identical for different species even in the same melt, a combination of semi-empirical models can still provide an estimation of the diffusion thresholds in ex-vessel corium melts. Thus, the first part of this paper presents an analysis of the applicability of such empirical models for simple silicate melts based on the published data. This is followed by an estimation of diffusivities in melt compositions typical of ex-vessel coriums. Alternatively, although the general trend towards a coupled description of the viscous flow and diffusion for ex-vessel corium melts seems promising, it is limited to published data on self-diffusivities of

  15. Status Report on Ex-Vessel Coolability and Water Management

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Robb, K. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-15

    Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell vent path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.

  16. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors - EXCOOLSE project report 2004

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Nayak, A.K.; Hansson, R.C.; Sehgal, B.R. [Royal Inst. of Technology, Div. of Nuclear Power Safety (Sweden)

    2005-10-01

    Beyond-the-design-basis accidents, i.e. severe accidents, involve melting of the nuclear reactor core and release of radioactivity. Intensive research has been performed for years to evaluate the consequence of the postulated severe accidents. Severe accidents posed, to the reactor researchers, a most interesting and most difficult set of phenomena to understand, and to predict the consequences, for the various scenarios that could be contemplated. The complexity of the interactions, occurring at such high temperatures ({approx} 2500 deg. C), between different materials, which are changing phases and undergoing chemical reactions, is simply indescribable with the accuracy that one may desire. Thus, it is a wise approach to pursue research on SA phenomena until the remaining uncertainty in the predicted consequence, or the residual risk, can be tolerated. In the PRE-DELI-MELT project at NKS, several critical issues on the core melt loadings in the BWR and PWR reactor containments were identified. Many of Nordic nuclear power plants, particularly in boiling water reactors, adopted the Severe Accident Management Strategy (SAMS) which employed the deep subcooled water pool in lower dry-well. The success of this SAMS largely depends on the issues of steam explosions and formation of debris bed and its coolability. From the suggestions of the PRE-DELI-MELT project, a series of research plan was proposed to investigate the remaining issues specifically on the ex-vessel coolability of corium during severe accidents; (a) ex-vessel coolability of the melt or particulate debris, and (b) energetics and debris characteristics of fuel-coolant interactions endangering the integrity of the reactor containments. (au)

  17. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors - EXCOOLSE project report 2004

    International Nuclear Information System (INIS)

    Park, H.S.; Nayak, A.K.; Hansson, R.C.; Sehgal, B.R.

    2005-10-01

    Beyond-the-design-basis accidents, i.e. severe accidents, involve melting of the nuclear reactor core and release of radioactivity. Intensive research has been performed for years to evaluate the consequence of the postulated severe accidents. Severe accidents posed, to the reactor researchers, a most interesting and most difficult set of phenomena to understand, and to predict the consequences, for the various scenarios that could be contemplated. The complexity of the interactions, occurring at such high temperatures (∼ 2500 deg. C), between different materials, which are changing phases and undergoing chemical reactions, is simply indescribable with the accuracy that one may desire. Thus, it is a wise approach to pursue research on SA phenomena until the remaining uncertainty in the predicted consequence, or the residual risk, can be tolerated. In the PRE-DELI-MELT project at NKS, several critical issues on the core melt loadings in the BWR and PWR reactor containments were identified. Many of Nordic nuclear power plants, particularly in boiling water reactors, adopted the Severe Accident Management Strategy (SAMS) which employed the deep subcooled water pool in lower dry-well. The success of this SAMS largely depends on the issues of steam explosions and formation of debris bed and its coolability. From the suggestions of the PRE-DELI-MELT project, a series of research plan was proposed to investigate the remaining issues specifically on the ex-vessel coolability of corium during severe accidents; (a) ex-vessel coolability of the melt or particulate debris, and (b) energetics and debris characteristics of fuel-coolant interactions endangering the integrity of the reactor containments. (au)

  18. Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

    1984-01-01

    Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity

  19. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    International Nuclear Information System (INIS)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun; Moriyama, Kiyofumi; Park, Jin Ho

    2016-01-01

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  20. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Moriyama, Kiyofumi; Park, Jin Ho

    2016-03-15

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  1. Analysis of ex-vessel melt jet breakup and coolability. Part 1: Sensitivity on model parameters and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kiyofumi; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Hwang, Byoungcheol; Jung, Woo Hyun

    2016-06-15

    Highlights: • Application of JASMINE code to melt jet breakup and coolability in APR1400 condition. • Coolability indexes for quasi steady state breakup and cooling process. • Typical case in complete breakup/solidification, film boiling quench not reached. • Significant impact of water depth and melt jet size; weak impact of model parameters. - Abstract: The breakup of a melt jet falling in a water pool and the coolability of the melt particles produced by such jet breakup are important phenomena in terms of the mitigation of severe accident consequences in light water reactors, because the molten and relocated core material is the primary heat source that governs the accident progression. We applied a modified version of the fuel–coolant interaction simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA) to a plant scale simulation of melt jet breakup and cooling assuming an ex-vessel condition in the APR1400, a Korean advanced pressurized water reactor. Also, we examined the sensitivity on seven model parameters and five initial/boundary condition variables. The results showed that the melt cooling performance of a 6 m deep water pool in the reactor cavity is enough for removing the initial melt enthalpy for solidification, for a melt jet of 0.2 m initial diameter. The impacts of the model parameters were relatively weak and that of some of the initial/boundary condition variables, namely the water depth and melt jet diameter, were very strong. The present model indicated that a significant fraction of the melt jet is not broken up and forms a continuous melt pool on the containment floor in cases with a large melt jet diameter, 0.5 m, or a shallow water pool depth, ≤3 m.

  2. Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

    1984-01-01

    The results of reactor material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address exvessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debrids characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity. (orig.)

  3. Ex-vessel debris coolability test during severe accident (COTELS project)

    International Nuclear Information System (INIS)

    Ogasawara, H.

    1998-01-01

    The objectives of the COTELS project are for severe accident management, to investigate phenomena of ex-vessel fuel-coolant interactions after reactor pressure vessel (RPV) failure and to investigate molten core-concrete interaction when coolant is injected onto molten debris. The project has being cooperated with the National Nuclear Center in the Republic of Kazakstan from 1994 to 1997 under the sponsorship of the Ministry of International Trade and Industry of Japan. Total programs are composed with the following tests. (1) Test 01 was meant to observe flow mode of falling debris. (2) Test A was meant to investigate phenomena of fuel-coolant interactions when molten debris falls into a coolant pool. (3) Test B/C investigated fuel coolant interactions and molten core-concrete interaction when coolant is injected onto debris. Detail data evaluation is underway. The following results were thus for obtained: (1) It was confirmed in Test 01 series that about 60 kg of UO 2 mixture was completely melted and fallen as a continuous jet. (2) No energetic fuel-coolant interaction was observed both in Test A and B series. (3) Debris in which decay heat was simulated was cooled by water injection in Test C series

  4. Ex-vessel corium spreading: results from the VULCANO spreading tests

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Christophe E-mail: christophe.journeau@cea.fr; Boccaccio, Eric E-mail: eric.boccaccio@cea.fr; Brayer, Claude; Cognet, Gerard E-mail: gerard.cognet@cea.fr; Haquet, Jean-Francois E-mail: haquet@eloise.cad.cea.fr; Jegou, Claude E-mail: claude.jegou@cea.fr; Piluso, Pascal E-mail: pascal.piluso@cea.fr; Monerris, Jose E-mail: jose.monerris@cea.fr

    2003-07-01

    In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture, called corium, of highly refractory oxides (UO{sub 2}, ZrO{sub 2}) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the basemat decomposition products (generally oxides such as SiO{sub 2}, Al{sub 2}O{sub 3}, CaO, Fe{sub 2}O{sub 3}, ...). For some years, the French Atomic Energy Commission (CEA) has launched an R and D program which aimed at providing the tools for improving the mastering of severe accidents. Within this program, the VULCANO experimental facility is operated to perform experiments with prototypic corium (corium of realistic chemical composition including depleted UO{sub 2}). This is coupled with the use of specific high-temperature instrumentation requiring in situ cross calibration. This paper is devoted to the 'spreading experiments' performed in the VULCANO facility, in which the effects of flow and solidification are studied. Due to the complex behavior of corium in the solidification range, an interdisciplinary approach has been used combining thermodynamics of multicomponent mixtures, rheological models of silicic semisolid materials, heat transfer at high temperatures, free-surface flow of a fluid with temperature-dependant properties. Twelve high-temperature spreading tests have been performed and analyzed. The main experimental results are the good spreadability of corium-concrete mixtures having large solidification ranges even with viscous silicic melts, the change of microstructure due to cooling rates, the occurrence of a large thermal contact resistance at the corium-substrate interface, the presence of a steep viscosity gradient at the surface, the transient concrete ablation. Furthermore, the experiments showed the presence of the gaseous inclusions in the melt even without concrete substrate. This gas release is linked to the local oxygen content in the melt which is

  5. Two-Phase Flow Effect on the Ex-Vessel Corium Debris Bed Formation in Severe Accident

    International Nuclear Information System (INIS)

    Kim, Eunho; Park, Jin Ho; Kim, Moo Hwan; Park, Hyun Sun; Ma, Weimin; Bechta, Sevostian V.

    2014-01-01

    In Korean IVR-ERVC(In-Vessel Retention of molten corium through External Reactor Vessel Cooling) strategy, if the situation degenerates into insufficient external vessel cooling, the molten core mixture can directly erupt into the flooded cavity pool from the weakest point of the vessel. Then, FCI (molten Fuel Coolant Interaction) will fragment the corium jet into small particulates settling down to make porous debris bed on the cavity basemat. To secure the containment integrity against the MCCI (Molten Core - Concrete Interaction), cooling of the heat generating porous corium debris bed is essential and it depends on the characteristics of the bed itself. For the characteristics of corium debris bed, many previous experimental studies with simulant melts reported the heap-like shape mostly. There were also following experiments to develop the correlation for the heap-like shaped debris bed. However, recent studies started to consider the effect of the decay heat and reported some noticeable results with the two-phase flow effect on the debris bed formation. The Kyushu University and JAEA group reported the experimental studies on the 'self-leveling' effect which is the flattening effect of the particulate bed by the inside gas generation. The DECOSIM simulation study of RIT (Royal Institute of Technology, Sweden) with Russian researchers showed the 'large cavity pool convection' effect, which is driven by the up-rising gas bubble flow from the pre-settled debris bed, on the particle settling trajectories and ultimately final bed shape. The objective of this study is verification of the two-phase flow effect on the ex-vessel corium debris bed formation in the severe accident. From the analysis on the test movie and resultant particle beds, the two-phase flow effect on the debris bed formation, which has been reported in the previous studies, was verified and the additional findings were also suggested. For the first, in quiescent pool the

  6. Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

    International Nuclear Information System (INIS)

    Jasiulevicius, Audrius

    2003-01-01

    This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

  7. PIV Visualization of Bubble Induced Flow Circulation in 2-D Rectangular Pool for Ex-Vessel Debris Bed Coolability

    Energy Technology Data Exchange (ETDEWEB)

    Han, Teayang; Kim, Eunho; Park, Hyun Sun; Moriyama, Kiyofumi [POSTECH, Pohang (Korea, Republic of)

    2015-10-15

    The previous research works demonstrated the debris bed formation on the flooded cavity floor in experiments. Even in the cases the core melt is once solidified, the debris bed can be re-melted due to the decay heat. If the debris bed is not cooled enough by the coolant, the re-melted debris bed will react with the concrete base mat. This situation is called the molten core-concrete interaction (MCCI) which threatens the integrity of the containment by generated gases which pressurize the containment. Therefore securing the long term coolability of the debris bed in the cavity is crucial. According to the previous research works, the natural convection driven by the rising bubbles affects the coolability and the formation of the debris bed. Therefore, clarification of the natural convection characteristics in and around the debris bed is important for evaluation of the coolability of the debris bed. In this study, two-phase flow around the debris bed in a 2D slice geometry is visualized by PIV method to obtain the velocity map of the flow. The DAVINCI-PIV was developed to investigate the flow around the debris bed. In order to simulate the boiling phenomena induced by the decay heat of the debris bed, the air was injected separately by the air chamber system which consists of the 14 air-flowmeters. The circulation flow developed by the rising bubbles was visualized by PIV method.

  8. Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Jasiulevicius, Audrius

    2003-07-01

    This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

  9. Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

  10. Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

    International Nuclear Information System (INIS)

    Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

    2015-01-01

    Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.

  11. Report of Task Group on Ex-Vessel Thermal-Hydraulics Corium/concrete interactions and combustible gas distribution in large dry containments

    International Nuclear Information System (INIS)

    1987-11-01

    The Task Group on Ex-Vessel Thermal-Hydraulics was established by the PWG 2 to address the physical processes that occur in the ex-vessel phase of severe accidents, to study their impact on containment loading and failure, and to assess the available calculation methods. This effort is part of an overall CSNI effort to come to an international understanding of the issues involved. The Task Group decided to focus its initial efforts on the Large Dry Containment used extensively to contain the consequences of postulated (design basis) accidents in Light Water Reactors (LWR). Although such containments have not been designed with explicit consideration of severe accidents, recent assessments indicate a substantial inherent capability for these accidents. The Task Group has examined the loads likely to challenge the integrity of the containment, and considered the calculation of the containment's response. This report is the outcome of this effort

  12. Problem of corium melt coolability in passive protection systems against severe accidents in the containment

    Directory of Open Access Journals (Sweden)

    Ali Kalvand

    2018-05-01

    Full Text Available Paper is devoted to the development of the mathematical model and analysis of the problem of corium melt interaction with low-temperature melting blocks in the passive protection systems against severe accidents at the NPP, which is of high importance for substantiation of the nuclear power safety, for building and successful op-erating of passive protection systems. In the third-generation reactors passive protection systems against severe accidents at the NPP are mandatory, therefore this paper is of importance for the nuclear power safety. A few configurations for the cooling blocks’ distribution have been considered and an analysis of the blocks’ melting and corium’s cooling in the pool under reactor vessel have been done, which can serve more effective for further improvement of the safety current systems and for the development of new ones. The ways for solution of the problems and the methods for their successful elaboration were discussed. The developed mathematical models and the analysis performed in the paper might be helpful for the design of passive protection systems of the cori-um melt retention inside the containment after corium melt eruption from the broken reactor vessel.

  13. Development and application of surrogate model for assessment of ex-vessel debris bed dryout probability - 15157

    International Nuclear Information System (INIS)

    Yakush, S.E.; Lubchenko, N.T.; Kudinov, P.

    2015-01-01

    In this work we consider a water-cooled power reactor severe accident scenario with pressure vessel failure and subsequent release of molten corium. A surrogate model for prediction of dryout heat flux for ex-vessels debris beds of different shapes is developed. Functional form of dryout heat flux dependence on problem parameters is developed by the analysis of coolability problem in non-dimensional variables. It is shown that for a flat debris bed the dryout heat flux can be represented in terms of three 1-dimensional functions for which approximating formulas are found. For two-dimensional debris beds (cylindrical, conical, Gaussian heap, mound-shaped), an additional function taking into account the bed shape geometry is obtained from numerical simulations using DECOSIM code as a full model. With the surrogate model in hand, risk analysis of debris bed coolability is carried out by Monte Carlo sampling of the input parameters within selected ranges, with assumed distribution functions

  14. Analysis of heat transfer mechanism on in-vessel corium coolability in severe accidents

    International Nuclear Information System (INIS)

    Park, Rae Joon; Jeong, Ji Whan; Kim, Sang Baik; Kang, Kyung Ho; Kim, Jong Whan

    1998-04-01

    When the molten core material relocates to the lower plenum of the reactor vessel, the cooling process of corium and the related heat transfer mechanism have been analyzed. The critical heat flux in gap (CHFG) test is being performed as a part of simulation of naturally arrested thermal attack in (SONATA-IV) project and the state of art on CHF has been reviewed. A series of complex heat transfer mechanism of molten pool formation, natural convection in the molten pool, solidification and remelting of the corium, conduction in the solidified crust, and boiling heat transfer to surroundings can be occurred in the lower plenum. Many studies are needed to investigate the complex heat transfer mechanism in the lower plenum, because these phenomena have not been clearly understand until now. The SONATA-IV/CHFG experiments are being carried out to develop CHF correlation in a hemispherical gap, which is the upper limit of heat transfer. There is no experimental or analytical CHF correlation applicable to a hemispherical gap. So lots of analytical and experimental correlations developed using the similar experimental condition were gathered and compared with each other. According to the experimental work that was carried out with pool boiling condition, CHF in a parallel gap was reduced by 1/30 compared with the value measured without gap. A basic form of a CHF correlation has been developed to correlate measurements that will be made in the SONATA-IV/CHFG experiments. That correlation is based on the fact that the CHF in a hemispherical gap is enhanced by CCFL and a Kutateladze type CCFL correlation develops CCFL date will in geometry like this. The experimental facility consists of a heater, a pressure vessel, a heat exchanger and lots of sensors. The heater capacity is 40 kw and the maximum heat flux at the surface is 100 kw/m 2 . The experiments will be carried out in the range of 1 to 10 atm and the gap size of 0.5, 1, 2 mm. The CHF will be detected using 66 type

  15. The VULCANO ex-vessel programme

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G.; Laffont, G.; Jegou, C.; Pierre, J.; Journeau, C.; Cranga, M.; Sudreau, F.; Ramacciotti, M. [CEA Cadarache, St. Paul lez Durance (France). Direction des Reacteurs Nucleaires

    2000-05-01

    Among the currently studied core-catcher projects, several concepts suppose corium spreading before cooling. In particular, the EPR (European pressurized reactor) core-catcher concept is based on mixing the corium with a special concrete, spreading the molten mixture on a large multi-layer surface cooled from the bottom and subsequently cooling by flooding with water. Therefore, melt spreading deserves intensive investigation in order to determine and quantify key phenomena which govern the spreading. In France, for some years now, the nuclear reactor division of the atomic energy commission (CEA/DRN) has undertaken a large program to improve knowledge on corium behaviour and coolability. This program is based on experimental and theoretical investigations which are finally gathered in scenario and mechanistic computer codes. Within this framework, the real material experimental programme, VULCANO, conducted in collaboration with European partners, is currently devoted to the study of corium spreading. Since 1997, several tests have been performed on dry corium spreading with various melt compositions. After a brief description of the general objectives and the facility, this paper will present the most important spreading results. (orig.)

  16. The VULCANO ex-vessel programme

    International Nuclear Information System (INIS)

    Cognet, G.; Laffont, G.; Jegou, C.; Pierre, J.; Journeau, C.; Cranga, M.; Sudreau, F.; Ramacciotti, M.

    2000-01-01

    Among the currently studied core-catcher projects, several concepts suppose corium spreading before cooling. In particular, the EPR (European pressurized reactor) core-catcher concept is based on mixing the corium with a special concrete, spreading the molten mixture on a large multi-layer surface cooled from the bottom and subsequently cooling by flooding with water. Therefore, melt spreading deserves intensive investigation in order to determine and quantify key phenomena which govern the spreading. In France, for some years now, the nuclear reactor division of the atomic energy commission (CEA/DRN) has undertaken a large program to improve knowledge on corium behaviour and coolability. This program is based on experimental and theoretical investigations which are finally gathered in scenario and mechanistic computer codes. Within this framework, the real material experimental programme, VULCANO, conducted in collaboration with European partners, is currently devoted to the study of corium spreading. Since 1997, several tests have been performed on dry corium spreading with various melt compositions. After a brief description of the general objectives and the facility, this paper will present the most important spreading results. (orig.)

  17. New sacrificial material for ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Komlev, Andrei A., E-mail: komlev@kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Nuclear Power Safety Division, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Almjashev, Vyacheslav I., E-mail: vac@mail.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Bechta, Sevostian V., E-mail: bechta@safety.sci.kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Khabensky, Vladimir B., E-mail: vladimirkhabensky@gmail.com [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Granovsky, Vladimir S., E-mail: gran@niti.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Gusarov, Victor V., E-mail: victor.v.gusarov@gmail.com [Ioffe Institute, 26 Polytekhnicheskaya Str., St. Petersburg, 194021 (Russian Federation)

    2015-12-15

    A new functional (sacrificial) material has been developed in the Fe{sub 2}O{sub 3}–SrO–Al{sub 2}O{sub 3}–CaO system based on strontium hexaferrite ceramic in concrete matrix. The method of producing SM has been advanced technologically; this technological effectiveness allows the SM to be used in ex-vessel core catchers with corium spreading as well as in crucible-type core catchers. Critical properties regarding the efficiency of SM in ex-vessel core catchers, such as porosity, pycnometric density, apparent density, solidus and liquidus temperatures, and water content have been measured. Suitable fractions of SrFe{sub 12}O{sub 19} and high alumina cement (HAC) were found in the SM based on thermodynamic analysis of the SM/corium interaction. The use of sacrificial steel as an additional heat adsorption component in the core catcher allowed us to increase the mass fraction range of SrFe{sub 12}O{sub 19} in the SM from 0.3−0.5 to 0.3–0.85. The activation temperature of the SM/corium interaction has been shown to correspond to the liquidus temperature of the local composition at the SM/corium interface. The calculated value of this temperature was 1716 °C. Analysis of phase transformations in the SrO–Fe{sub 2}O{sub 3} system revealed advantages of the SrFe{sub 12}O{sub 19}–based sacrificial material compared with the Fe{sub 2}O{sub 3}-contained material owing to the time proximity of SrFe{sub 12}O{sub 19} decomposition and corium interaction activation. - Highlights: • A sacrificial material (SM) was developed for ex-vessel core catcher. • Suitable proportions in the SrFe{sub 12}O{sub 19}–Al{sub 2}O{sub 3}·CaO–Fe system were determined. • Hydrogen release limitation was shown for ex-vessel corium retention with the SM. • Calculated temperature of the active initiation of corium/SM interaction is 1716 °C. • Functional properties of the SM were measured.

  18. PWG4 perspective on ex-vessel hydrogen sources

    International Nuclear Information System (INIS)

    2000-07-01

    The purpose of this perspective document is to identify the potential ex-vessel hydrogen sources and to address the question whether, considered the uncertainties associated to these sources, further investigations are required. The statement is established with reference to the needs for safety evaluation of nuclear reactors under severe accident conditions. It is recognised that the views could be different if one looks at these issues from another standpoint. Since the TMI-2 accident in 1979, there had been a large interest in the nuclear reactor safety community for studying the behaviour of hydrogen in case of a severe accident. As a result, different 'state of the art' reports were produced. Examples of these documents are NUREG/CR-1561 and EUR 14307. In particular, they identified potential hydrogen sources during accidents, including ex-vessel sources. Various ex-vessel hydrogen sources, covering a variety of physical and chemical processes, were identified. Although their precise quantification and relative importance is to be established on a case by case basis with respect to the specific reactor design of interest, general trends can be formulated. The sources to be considered are the followings: - radiolysis of water; - corrosion reactions, - reaction of urania with steam and water; - core-concrete interaction; - debris-atmosphere interaction. These sources are discussed successively. The PWG4 (CSNI's Principal Working Group on the Confinement of Accidental Radioactive Releases) perspective on Ex-vessel Hydrogen Sources can be summarised in the following statements: 1. The issue of hydrogen sources must be considered as a whole and cannot be separated into in-vessel and ex-vessel issues. For significant sources that may not be accommodated by mitigation means associated to DBA, the uncertainty is largely dominated by the unknown extent of Zr oxidation during the in-vessel phase. 2. PWG4 notes that hydrogen production during corium quenching by water is

  19. Prediction of corium debris characteristics in lower plenum of a nordic BWR in different accident scenarios using MELCOR code - 15367

    International Nuclear Information System (INIS)

    Phung, V.A.; Galushin, S.; Raub, S.; Goronovski, A.; Villanueva, W.; Koeoep, K; Grishchenko, D.; Kudinov, P.

    2015-01-01

    Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed

  20. FARO tests corium-melt cooling in water pool: Roles of melt superheat and sintering in sediment

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Gisuk [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States); Kaviany, Massoud [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Moriyama, Kiyofumi [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hwang, Byoungcheol; Lee, Mooneon; Kim, Eunho; Park, Jin Ho [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Nasersharifi, Yahya [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States)

    2016-08-15

    Highlights: • The numerical approach for FARO experimental data is suggested. • The cooling mechanism of ex-vessel corium is suggested. • The predicted minimum pool depth for no cake formation is suggested. - Abstract: The FARO tests have aimed at understanding an important severe accident mitigation action in a light water reactor when the accident progresses from the reactor pressure vessel boundary. These tests have aimed to measure the coolability of a molten core material (corium) gravity dispersed as jet into a water pool, quantifying the loose particle diameter distribution and fraction converted to cake under range of initial melt superheat and pool temperature and depth. Under complete hydrodynamic breakup of corium and consequent sedimentation in the pool, the initially superheated corium can result in debris bed consisting of discrete solid particles (loose debris) and/or a solid cake at the bottom of the pool. The success of the debris bed coolability requires cooling of the cake, and this is controlled by the large internal resistance. We postulate that the corium cake forms when there is a remelting part in the sediment. We show that even though a solid shell forms around the melt particles transiting in the water pool due to film-boiling heat transfer, the superheated melt allows remelting of the large particles in the sediment (depending on the water temperature and the transit time) using the COOLAP (Coolability Analysis with Parametric fuel-cooant interaction models) code. With this remelting and its liquid-phase sintering of the non-remelted particles, we predict the fraction of the melt particles converting to a cake through liquid sintering. Our predictions are in good agreement with the existing results of the FARO experiments. We address only those experiments with pool depths sufficient/exceeding the length required for complete breakup of the molten jet. Our analysis of the fate of molten corium aimed at devising the effective

  1. OECD/CSNI Workshop on In-Vessel Core Debris Retention and Coolability - Summary and Conclusions

    International Nuclear Information System (INIS)

    Behbahani, Ali-Reza; Drozd, Andrzej; Kim, Sang-Baik; Micaelli, Jean-Claude; Okkonen, Timo; Sugimoto, Jun; Trambauer, Klaus; Tuomisto, Harri

    1999-01-01

    In the spring of 1994 an OECD Workshop on Large Pool Heat transfer was held in Grenoble. The scope of this workshop was the investigation of (1) molten pool heat transfer, (2) heat transfer to the surrounding water, and (3) the feasibility of in-vessel core debris cooling through external cooling of the vessel. Since this time, experimental test series have been completed (e.g., COPO, ULPU, CORVIS) and new experimental programs (e.g., BALI, SONATA, RASPLAV, debris and gap heat transfer) have been established to consolidate and expand the data base for further model development and to improve the understanding of in-vessel debris retention and coolability in a nuclear power plant. Discussions within the CSNI's PWG-2 and the Task Group on Degraded Core Cooling (TG-DCC) have led to the conclusion that the time was ripe for organizing a new international Workshop with the objectives: - to review the results of experimental research that has been conducted in this area; - to exchange information on the results of member countries experiments and model development on in-vessel core debris retention and coolability; - to discuss areas where additional experimental research is needed in order to provide an adequate data base for analytical model development for core debris retention and coolability. The scope of this workshop was limited to the phenomena connected to in-vessel core debris retention and coolability and did not include steam explosion and fission product issues. The workshop was structured into the following sessions: Key note papers; Experiments and model development; Debris bed heat transfer; Corium properties, molten pool convection and crust formation; Gap formation and gap cooling; Creep behaviour of reactor pressure vessel lower head; Ex-vessel boiling and critical heat flux phenomena; Scaling to reactor severe accident conditions and reactor applications. Compared to the previous workshop held in Grenoble in 1994, large progress has been made in the

  2. Thermal Load Analysis of Multilayered Corium in the Lower Head of Reactor Pressure Vessel during Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Seok Won; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Hwang, Tae Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    In-Vessel Retention (IVR) is one of the severe accident management strategies to terminate or mitigate the severe accident which is also called 'core-melt accident'. The reactor vessel would be cooled by flooding the cavity with water. The molten core mixture is divided into two or three layers due to the density difference. Light metal layer which contains Fe and Zr is on the oxide layer which is consist of UO{sub 2} and ZrO{sub 2}. Heavy metal layer which contains U, Fe and Zr is located under the oxide layer. In oxide layer, the crust which is solidified material is formed along the boundary. The assessment of IVR for nuclear power plant has been conducted with lumped parameter method by Theofanous, Rempe and Esmaili. In this paper, the numerical analysis was performed and verified with the Esmaili's work to analyze thermal load of multilayered corium in pressurized reactor vessel and also to examine the condition of in-vessel corium characteristic before the vessel failure that lead to ex-vessel severe accident progression for example, ex-vessel debris bed cooling. The in-vessel coolability analysis for several scenarios is conducted for the plant which has higher power than AP1000. Two sensitivity analyses are conducted, the first is emissivity of light metal layer and the second is the heat transfer coefficient correlations of oxide layer. The effect of three layered system also investigated. In this paper, the numerical analysis was performed and verified with Esmaili's model to analyze thermal load of multilayered corium in pressurized reactor vessel. For two layered system, thermal load was analyzed according to the severe accident scenarios, emissivity of the light metal layer and heat transfer correlations of the.

  3. Proceedings of the Workshop on in-vessel core debris retention and coolability

    International Nuclear Information System (INIS)

    1999-01-01

    on in-vessel debris coolability through inherent cooling mechanisms, FOREVER experiments on thermal and mechanical behaviour of a reactor pressure vessel during a severe accident, Experimental studies of heat transfer in the slotted channels at the CTF facility, Experimental study on CHF in a hemispherical narrow gap, Experiments on heat removal in a gap between debris crust and RPV wall), sub-session 4 (Creep behaviour of reactor pressure vessel lower head: Experimental investigation of creep behaviour of RPV lower head, Lower head thermo-mechanical behaviour, Pressure vessel creep rupture analysis, Parametric studies on creep behavior of a reactor pressure vessel lower head, Study of RPV materials with respect to mechanical behaviour in case of complete core fusion), sub-session 5 (Ex-vessel boiling and critical heat flux phenomena: Natural convection boiling on the outer surface of a hemispherical vessel surrounded by a thermal insulation structure, Reactor vessel external cooling for corium retention SULTAN experimental program and modelling with CATHARE code), and session 3 (Scaling to reactor severe accident conditions and reactor applications: Potential for in-vessel retention through ex-vessel flooding, In-vessel core melt retention by RPV external cooling for high power PWR MAAP4 analysis on a LBLOCA scenario without SI, Coupled thermal-hydraulic analyses of the molten pool and pressure vessel during a severe accident, Studies on core melt behaviour in a BWR pressure vessel lower head, Analysis of reactor lower head penetration tube failure, Thermal hydraulic and mechanical aspects of in-vessel retention of core debris)

  4. A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

    International Nuclear Information System (INIS)

    Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

    2015-01-01

    A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

  5. Experimental study of in-and-ex-vessel melt cooling during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Baik; Yoo, K J; Park, C K; Seok, S D; Park, R J; Yi, S J; Kang, K H; Ham, Y S; Cho, Y R; Kim, J H; Jeong, J H; Shin, K Y; Cho, J S; Kim, D H

    1997-07-01

    After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into in-vessel and ex-vessel cooling depending on the location of molten core which is dependent on the timing of vessel failure. Since the cooling mechanism varies with the conditions of molten core and surroundings and related phenomena, it contains many phenomenological uncertainties so far. In this study, an experimental study for verification of in-vessel corium cooling and several separate effect experiments for ex-vessel cooling are carried out to verify in- and ex-vessel cooling phenomena and finally to develop the accident management strategy and improve engineered reactor design for the severe accidents. SONATA-IV (Simulation of Naturally Arrested Thermal Attack in Vessel) program is set up for in-vessel cooling and a progression of the verification experiment has been done, and an integral verification experiment of the containment integrity for ex-vessel cooling is planned to be carried out based on the separate effect experiments performed in the first phase. First phase study of SONATA-IV is proof of principle experiment and it is composed of LALA (Lower-plenum Arrested Vessel Attack) experiment to find the gap between melt and the lower plenum during melt relocation and to certify melt quenching and CHFG (Critical Heat Flux in Gap) experiment to certify heat transfer mechanism in an artificial gap. As separate effect experiments for ex-vessel cooling, high pressure melt ejection experiment related to the initial condition for debris layer formation in the reactor cavity, crust formation and heat transfer experiment in the molten pool and molten core concrete interaction experiment are performed. (author). 150 refs., 24 tabs., 127 figs.

  6. Étude thermodynamique du corium en cuve - Application à l'interaction corium/béton

    OpenAIRE

    Quaini , Andrea

    2015-01-01

    During a severe accident in a pressurised water reactor, the nuclear fuel can interact with the Zircaloy cladding, the neutronic absorber and the surrounding metallic structure forming a partially or completely molten mixture. The molten core can then interact with the reactor steel vessel forming a mixture called in-vessel corium. In the worst case, this mixture can pierce the vessel and pour onto the concrete underneath the reactor, leading the formation of the ex-vessel corium. Furthermore...

  7. First evaluations of ex-vessel fuel-coolant interaction with MC3D

    International Nuclear Information System (INIS)

    Meignen, R.; Dupas, J.; Chaumont, B.

    2003-01-01

    In the frame of severe accident nuclear safety studies, we evaluate for French PWR's the potential of Steam Explosion in the reactor pit, consecutively to a vessel failure and to the mixing of the corium with the water that might be present. The evaluations are made with MC3D. This thermalhydraulic multiphasic code has firstly been qualified and its main parameters chosen so that a sufficient validation is obtained with regards to reactor situations. The safety study for ex-vessel situations is a step-by-step procedure that leads to a progressive process of hypotheses relaxations. We find that it is important to adequately model the corium ejection from the RPV. The rapid transition of the flow at the breach towards 2-phase dispersed flow leads to an important mixing of corium and water. The vessel pressurization is a very important parameter and strong pressure cases lead to a fine fragmentation and thus a high voiding. The small pressure cases are more dangerous for two reasons: the corium is dispersed in larger drops, and some important interactions (in the premixing sense) are reported

  8. Experimental studies of oxidic molten corium-vessel steel interaction

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V.

    2001-01-01

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere

  9. Experimental studies of oxidic molten corium-vessel steel interaction

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. E-mail: niti-npc@sbor.net; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V

    2001-12-01

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

  10. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, K.I.; Kim, B.S.; Kim, D.H. [Korea Atomic Energy Research Inst., Thermal Hydraulic Safety Research, Taejon (Korea, Republic of)

    2001-07-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

  11. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    International Nuclear Information System (INIS)

    Ahn, K.I.; Kim, B.S.; Kim, D.H.

    2001-01-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

  12. Ex-vessel core catcher design requirements and preliminary concepts evaluation

    International Nuclear Information System (INIS)

    Friedland, A.J.; Tilbrook, R.W.

    1974-01-01

    As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant

  13. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary

  14. The evaluation of pressure effects on the ex-vessel cooling for KNGR with MELCOR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hwa; Park, Soo Yong; Kim, Dong Ha

    2001-03-01

    In this report, the effect of external vessel cooling on debris coolability and vessel integrity for the KNGR were examined from the two typical pressure range of high(170 bar) and low(5 bar)case using the lower plenum model in MELCOR1.8.4. As the conditions of these calculations, 80 ton of debris was relocated simultaneously into the lower vessel head and the debris relocation temperature from the core region was 2700 K. The decay heat has been assumed to be that of one hour after reactor shutdown. The creep failure of the vessel wall was simulated with 1-D model, which can consider the rapid temperature gradient over the wall thickness during the ex-vessel cooling. From the calculation results, both the coolant temperature and the total amount of coolant mass injected into the cavity are known to be the important factors in determining the time period to keep the external vessel cool. Therefore, a long-term strategy to keep the coolant temperature subcooled throughout the transient is suggested to sustain or prolong the effect of external vessel cooling. Also, it is expected that to keep the primary side at low pressure and to perform the ex-vessel flooding be the essential conditions to sustain the vessel integrity. From MELCOR, the penetration failure always occurs after relocation regardless of the RCS pressure or availability of the external vessel cooling. Therefore, It is expected that the improvement of the model for the penetration tube failure will be necessary.

  15. ITER diagnostics ex-vessel engineering services

    Energy Technology Data Exchange (ETDEWEB)

    Arumugam, A.P., E-mail: arun.prakash@iter.org; Walker, C.I.; Andrew, P.; Barnsley, R.; Beltran, D.; Bertalot, L.; Dammann, A.; Direz, M.F.; Drevon, J.M.; Encheva, A.; Giacomin, T.; Hourtoule, J.; Kuehn, I.; Lanza, R.; Levesy, B.; Maquet, P.; Patel, K.M.; Patisson, L.; Pitcher, C.S.; Portales, M.; and others

    2013-10-15

    Highlights: • This paper describes about the ITER diagnostics ex-vessel engineering services. • It describes various diagnostics systems, its location and its environment. • Diagnostics interfaces with other services such as the buildings, HVAC, electrical services, cooling water, vacuum, liquid and gas distribution. • All the interfaces with these services are identified and defined. • Buildings services for diagnostics, such as penetrations, local shielding, embedment and temperature control are discussed. -- Abstract: Extensive diagnostics systems will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include measurements of temperature, density, impurity concentration, and particle and energy confinement times. ITER diagnostic systems extend from the center of the Tokamak to the various diagnostic areas, where they are controlled and acquired data is processed. This mainly includes the areas such as ports, port cells, gallery, diagnostics enclosures and cubicle areas. The diagnostics port plugs encloses the front end of the diagnostic systems and the diagnostics building houses the diagnostics equipment, instrumentation and control cubicles. There are several systems providing services to diagnostics. These mainly include ITER buildings, electrical power services, cooling water services, Heating Ventilation and Air Conditioning (HVAC), vacuum services, liquid and gas distribution services, cable engineering, de-tritiation systems, control cubicles, etc. Requirements of these service systems have to be defined, even though many of the diagnostics are at an early stage of development. It is a real challenge to define and to design diagnostics systems considering the constraints imposed by these service systems. This paper summarizes the provision of these services to the individual diagnostics and diagnostics areas

  16. ITER diagnostics ex-vessel engineering services

    International Nuclear Information System (INIS)

    Arumugam, A.P.; Walker, C.I.; Andrew, P.; Barnsley, R.; Beltran, D.; Bertalot, L.; Dammann, A.; Direz, M.F.; Drevon, J.M.; Encheva, A.; Giacomin, T.; Hourtoule, J.; Kuehn, I.; Lanza, R.; Levesy, B.; Maquet, P.; Patel, K.M.; Patisson, L.; Pitcher, C.S.; Portales, M.

    2013-01-01

    Highlights: • This paper describes about the ITER diagnostics ex-vessel engineering services. • It describes various diagnostics systems, its location and its environment. • Diagnostics interfaces with other services such as the buildings, HVAC, electrical services, cooling water, vacuum, liquid and gas distribution. • All the interfaces with these services are identified and defined. • Buildings services for diagnostics, such as penetrations, local shielding, embedment and temperature control are discussed. -- Abstract: Extensive diagnostics systems will be installed on the ITER machine to provide the measurements necessary to control, evaluate and optimize plasma performance in ITER and to further the understanding of plasma physics. These include measurements of temperature, density, impurity concentration, and particle and energy confinement times. ITER diagnostic systems extend from the center of the Tokamak to the various diagnostic areas, where they are controlled and acquired data is processed. This mainly includes the areas such as ports, port cells, gallery, diagnostics enclosures and cubicle areas. The diagnostics port plugs encloses the front end of the diagnostic systems and the diagnostics building houses the diagnostics equipment, instrumentation and control cubicles. There are several systems providing services to diagnostics. These mainly include ITER buildings, electrical power services, cooling water services, Heating Ventilation and Air Conditioning (HVAC), vacuum services, liquid and gas distribution services, cable engineering, de-tritiation systems, control cubicles, etc. Requirements of these service systems have to be defined, even though many of the diagnostics are at an early stage of development. It is a real challenge to define and to design diagnostics systems considering the constraints imposed by these service systems. This paper summarizes the provision of these services to the individual diagnostics and diagnostics areas

  17. Corium spreading issue; Le corium et son etalement

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G.; Brayer, C.; Cranga, M.; Journeau, C.; Laffont, G.; Splinder, B.; Veteau, J.M. [CEA Grenoble, Dept. de Thermohydraulique et de Physique (DPT), 38 (France)

    1999-07-01

    Safety is one of the major issues for nuclear power plants; its improvement is a constant R and D axis for the CEA. In the event of a highly unlikely core melt-down accident in Light Water Reactors, the Safety Authorities of several EU countries have requested the industries and utilities to consider severe accidents with reactor pressure vessel failure for the design of the next generation of nuclear power plants. The objective is to preserve the integrity of the containment as the main barrier of fission product release to the environment. This can only be achieved if the core melt mixture (called corium, essentially composed of UO{sub 2}, ZrO{sub 2}, Zr, Fe and fission products) is stabilized before it can penetrate the basement. Consequently, various core-catcher concepts are under investigation for future reactors in order to prevent basement erosion, and to stabilize and control the corium within the containment. In particular, in the EPR (European Pressurized Reactor) core-catcher concept, the corium is mixed with a special concrete, and the molten mixture spread over a large multi-layer surface cooled from the bottom; with subsequent cooling by flooding with water. Therefore, melt spreading requires intensive investigation in order to determine and quantify the key phenomena, which govern the spreading. For some years now, the Nuclear Reactor Division of the Atomic Energy Commission (CEA/DRN) has been conducting a large program to improve knowledge on corium behaviour and coolability. This program is based on experimental (with simulant and prototypic materials) and theoretical investigations, which are finally gathered into scenario and mechanistic computer codes. Within this framework, a large part is currently devoted to the study of corium spreading. After a reminder of the general objectives and a description of the DRn approach and facilities, this paper presents the most important results. (authors)

  18. Melt coolability modeling and comparison to MACE test results

    International Nuclear Information System (INIS)

    Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.

    1992-01-01

    An important question in the assessment of severe accidents in light water nuclear reactors is the ability of water to quench a molten corium-concrete interaction and thereby terminate the accident progression. As part of the Melt Attack and Coolability Experiment (MACE) Program, phenomenological models of the corium quenching process are under development. The modeling approach considers both bulk cooldown and crust-limited heat transfer regimes, as well as criteria for the pool thermal hydraulic conditions which separate the two regimes. The model is then compared with results of the MACE experiments

  19. Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

    International Nuclear Information System (INIS)

    Okkonen, T.; Dinh, T.N.; Bui, V.A.; Sehgal, B.R.

    1995-07-01

    Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs

  20. Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T; Dinh, T N; Bui, V A; Sehgal, B R [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Energy Systems Technology

    1995-07-01

    Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs.

  1. MELCOR ex-vessel LOCA simulations for ITER+

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

    1995-01-01

    Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

  2. An assessment of ex-vessel fuel-coolant interaction energetics for advanced light water reactors

    International Nuclear Information System (INIS)

    Murphy, J.G.; Corradini, M.L.

    1997-01-01

    The occurrence of an energetic fuel/coolant interaction (FCI) below the reactor pressure vessel in the cavity of advanced light water reactors (ALWRs) are analyzed to determine the possible hazard to structural walls as a result of dynamic liquid phase pressures. Such analyses are important to demonstrate that these cavity walls will maintain their integrity so that ex-vessel core debris coolability is possible. Past studies that have examined this or related issues are reviewed, and a methodology is proposed to analyze the occurrence of this physical event using the IFCI and TEXAS models for the FCI as well as dynamic shock wave propagation estimates using hand calculations as well as the CTH hydro model. Scenarios for the ALWRs are reviewed, and one severe accident scenario is used as an example to demonstrate the methodology. Such methodologies are recommended for consideration in future safety studies. These methodologies should be verified with direct comparison to energetic FCI data such as that being produced in KROTOS at the Joint Research Centre, Ispra

  3. Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

    International Nuclear Information System (INIS)

    Cho, Dae H.; Page, Richard J.; Abdulla, Sherif H.; Anderson, Mark H.; Klockow, Helge B.; Corradini, Michael L.

    2006-01-01

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of our work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb-Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed

  4. Investigation on energetics of ex-vessel vapor explosion based on spontaneous nucleation fragmentation

    International Nuclear Information System (INIS)

    Liu, Jie; Koshizuka, Seiichi; Oka, Yoshiaki

    2002-01-01

    A computer code PROVER-I is developed for propagation phase of vapor explosion. A new thermal fragmentation model is proposed with three kinds of time scale for modeling instant fragmentation, spontaneous nucleation fragmentation and normal boiling fragmentation. The energetics of ex-vessel vapor explosion is investigated based on different fragmentation models. A higher pressure peak and a larger mechanical energy conversion ratio are obtained by spontaneous nucleation fragmentation. A smaller energy conversion ratio results from normal boiling fragmentation. When the delay time in thermal fragmentation model is near 0.0 ms, the pressure propagation behavior tends to be analogous with that in hydrodynamic fragmentation. If the delay time is longer, pressure attenuation occurs at the shock front. The high energy conversion ratio (>4%) is obtained in a small vapor volume fraction together with spontaneous nucleation fragmentation. These results are consistent with fuel-coolant interaction experiments with alumina melt. However, in larger vapor volume fraction conditions (α υ >0.3), the vapor explosion is weak. For corium melt, a coarse mixture with void fraction of more than 30% can be generated in the pre-mixing process because of its physical properties. In the mixture with such a high void fraction the energetic vapor explosion hardly takes place. (author)

  5. A feasibility experiment for assessing the efficacy of ex-vessel cooling through the external gap structure

    International Nuclear Information System (INIS)

    Kang, K. H.; Kim, J. H.; Park, L. J.; Kim, S. B.; Hwang, I. S.

    1999-01-01

    This paper presents the results of a feasibility experiment for assessing the efficacy of ex-vessel cooling through the external gap structure during a severe accident. In this study, a 1/8 linear scale mockup of a lower plenum was used with Al2O3/Fe thermite melt as a corium simulant. The results show that in dry case test conducted without cooling the outside of the vessel, after about thirty second from the thermite ignition the vessel was heated to cause a complete melt penetration at about 30 degree upper position from the bottom. Whereas in wet case test conducted cooling the outside of the vessel with 0.85 kg/s of water flow rate using 2.5 cm of uniform gap structure, the vessel effectively cooled down with 23.7 K/s of cooling rate by nucleate boiling at the surface of the vessel. The results of two-dimensional analyses using FLUENT code show a similar trend of vessel thermal behavior presented in the tests. Synthesized the results of the tests and analyses work, a natural convection of the melt pool could cause the formation of hot spot at the upper portion of the vessel, but the vessel could effectively cool down by heat removal with ex-vessel cooling

  6. Final synthesis of Sarnet (Phase 1) corium activities

    International Nuclear Information System (INIS)

    Journeau, Ch.; Steinbruck, M.; Repetto, G.; Duriez, Ch.; Koundy, V.; Ma, W.M.; Burger, M.; Spindler, B.

    2009-01-01

    Within the SARNET Severe Accident Research Network of excellence, the Corium topic covers all the behaviour of corium (mixture formed by the molten materials arising from a postulated nuclear reactor severe accident) from early phase of core degradation to in or ex-vessel corium recovery with the exception of corium interaction with water, direct containment heating and fission product release. The Corium topic regroups in three work packages the critical mass of competence to improve significantly the corium behaviour knowledge. The spirit of the SARNET networking is to share the knowledge, the facilities and the simulation tools for severe accidents, so to reach a better efficiency and to rationalize the R and D effort at European level. Extensive benchmarking has been launched in most of the areas of research. These benchmarks were mainly dedicated to the recalculation of analytical experiments, integral experiments or reactor applications. Eventually, all the knowledge will be accumulated in the ASTEC severe accident simulation code through physical model improvements and extension of validation database. This report summarizes the progress that has been achieved in the frame of the networking activities for the four and half years of the FP6 project. (authors)

  7. Ex-Vessel Steam Explosion Analysis of Central Melt Pour Scenario

    International Nuclear Information System (INIS)

    Ursic, M.; Leskovar, M.

    2008-01-01

    An ex-vessel steam explosion may develop during a severe reactor accident when the reactor vessel fails and the molten core interacts with the coolant in the reactor cavity. At this process part of the corium energy is intensively transferred to water in a very short time scale. The water vaporizes at high pressure and expands, doing work on its surrounding. Although the steam explosion has probably a low probability of occurrence, it is an important nuclear safety issue in case of a severe reactor accident. Namely, the formed very high pressure region induces dynamic loadings on the surrounding structures that may potentially lead to an early release of the radioactive material into the environment. Although the steam explosion events have being studied for several years, the level of the process and consequences understanding is still not adequate. To increase the level of confidence the OECD programme SERENA (Steam Explosion REsolution for Nuclear Applications) was established in 2002. The objectives of the program were to evaluate capabilities of the current generation of the FCI (Fuel-Coolant Interaction) computer codes in predicting the steam explosion induced loads, identifying key FCI phenomena and associated uncertainties impacting the predictability of the steam explosion energetics in the reactor situations and proposing confirmatory research to reduce the uncertainties to acceptable levels for the steam explosion risk assessment. To get a better insight into the most challenging ex-vessel steam explosions, analyses for different locations of the melt release, the cavity water sub-cooling, the primary system pressure overpressure and the triggering time were preformed for a typical pressurized water reactor cavity. The results of some scenarios revealed that significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. Among the performed analyses for the central melt pour scenarios, the maximum pressure loads were

  8. Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.; D'Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

    1993-09-01

    The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The 'Concrete Benchmark' experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the 'Concrete Benchmark' experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs

  9. Application of CAMP code to analysis of debris coolability experiments in ALPHA program

    International Nuclear Information System (INIS)

    Maruyama, Yu; Moriyama, Kiyofumi; Park, Hyun-Sun; Yang, Yanhua; Sugimoto, Jun

    1999-01-01

    An analytical code for thermo-fluid dynamics of a molten debris, CAMP, was applied to the analysis of the ex-vessel and in-vessel debris coolability experiments performed in ALPHA program. The analysis on the ex-vessel debris coolability experiments, where water was added onto a layer of thermite melt, indicated that the upper surface of the melt was remained molten during a period when melt eruptions followed by a mild steam explosion were observed. This might imply that a coarse mixing between the melt and the overlying water could have been formed if a sufficient force was generated at the interface between the two fluids. In the analysis of the in-vessel debris coolability experiments, where an aluminum oxide (Al 2 O 3 ) melt was poured into a water-filled lower head experimental vessel, a temperature increase at the outer surface of the vessel was qualitatively reproduced when a gap was assumed to be at the interface between the solidified Al 2 O 3 and the vessel wall. (author)

  10. Thermal hydraulic phenomena in corium pools: the BALI experiment

    International Nuclear Information System (INIS)

    Bonnet, J.M.

    1999-01-01

    In the framework of severe accident studies, the BALI experiment has been designed to create a data base about heat transfer distribution at corium pool boundaries for in-vessel or ex-vessel configurations. The mechanism investigated is natural convection at high internal Rayleigh number (10 15 to 10 17 ) in cavities with volumetric heating. After a description of the facility and a synthesis of results obtained for in-vessel configurations, the purpose of this paper is to present or extend local or average heat transfer correlations in the prototypic range of dimensionless parameters. (author)

  11. Thermodynamic study on the in-vessel corium - Application to the corium/concrete interaction

    International Nuclear Information System (INIS)

    Quaini, Andrea

    2015-01-01

    During a severe accident in a pressurised water reactor, the nuclear fuel can interact with the Zircaloy cladding, the neutronic absorber and the surrounding metallic structure forming a partially or completely molten mixture. The molten core can then interact with the reactor steel vessel forming a mixture called in-vessel corium. In the worst case, this mixture can pierce the vessel and pour onto the concrete underneath the reactor, leading the formation of the ex-vessel corium. Furthermore, depending on the considered scenario, the corium can be formed by a liquid phase or by two liquids, one metallic the other oxide. The objective of this thesis is the investigation of the thermodynamics of the prototypic in-vessel corium U-Pu-Zr- Fe-O. The approach used during the thesis is based on the CALPHAD method, which allows to obtain a thermodynamic model for this complex system starting from phase diagram and thermodynamic data. Heat treatments performed on the O-U-Zr system allowed to measure two tie-lines in the miscibility gap in the liquid phase at 2567 K. Furthermore, the liquidus temperatures of three Zr-enriched samples have been obtained by laser heating in collaboration with ITU. With the same laser heating technique, solidus temperatures have been obtained on the UO 2 -PuO 2 -ZrO 2 system. The influence of the reducing or oxidising on the melting behaviour of this system has been studied for the first time. The results show that the oxygen stoichiometry of these oxides strongly depends on the oxygen potential and on the metal composition of the samples. The miscibility gap in the liquid phase of the U-Zr-Fe-O system has been also observed. The whole set of experimental results with the literature data allowed to develop the thermodynamic model of the U-Pu-Zr-Fe-O system. Solidification path calculations have been performed for all the investigated samples to interpret the microstructures of the solidified samples. A good accordance has been obtained between

  12. Corium melt researches at VESTA test facility

    Directory of Open Access Journals (Sweden)

    Hwan Yeol Kim

    2017-10-01

    Full Text Available VESTA (Verification of Ex-vessel corium STAbilization and VESTA-S (-small test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging ZrO2 melt jet on a sacrificial material were performed to investigate the ablation characteristics. ZrO2 melt in an amount of 65–70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40, and the other is a stainless steel (SUS304 melt. Metallic melt in an amount of 1.5–2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. ZrO2 melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is UO2 60%, Zr 10%, ZrO2 15%, SUS304 14%, and B4C 1%, was melted in a

  13. Experimental results on the coolability of a debris bed with multidimensional cooling effects

    International Nuclear Information System (INIS)

    Rashid, M.; Kulenovic, R.; Laurien, E.; Nayak, A.K.

    2011-01-01

    Research highlights: ► Performing of dryout experiments with a polydispersed bed for top- and bottom-flooding. ► Study of influence of different down comer configurations on the coolability of debris bed. ► Measurement of temperature profiles, pressure drops and determination of dryout heat flux. ► Observation of noticeable increase in coolability of debris bed with the use of down comer is observed. - Abstract: Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long term coolability of debris beds, the scaled test facility “DEBRIS” (Fig. 1) has been built at IKE. A large number of experiments had been carried out to investigate the coolability limits for different bed configurations (). Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favoured. Following the experiments with top- and bottom-flooding flow conditions this paper presents experimental results of boiling and dryout tests at different system pressures based on top- and bottom-flooding via a down comer configuration. A down comer with an internal diameter of 10 mm has been installed at the centre of the debris bed. The debris bed is built up in a cylindrical crucible with an inner diameter of 125 mm. The bed of height 640 mm is composed of polydispersed particles with particle diameters 2, 3 and 6 mm. Since the long term coolability of such particle bed is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the bottom inflow of water improves the coolability of the debris bed and an increase of the dryout heat flux can be observed. With increasing system pressure, the coolability limits are enhanced (increased dryout heat flux).

  14. Remote maintenance of Compact Ignition Tokamak ex-vessel systems

    International Nuclear Information System (INIS)

    DePew, R.E.; Macdonald, D.

    1989-01-01

    The use of deuterium-tritium (D-T) fuel in the Compact Ignition Tokamak (CIT) will require applying remote handling technology for ex-vessel maintenance and replacement of machine components. Highly activated and contaminated components of the fusion device's auxiliary systems, such as diagnostics and RF heating, must be replaced using remotely operated maintenance equipment in the test cell. Throughout the CIT remote maintenance (RM) studies conducted to date, computer modeling has been used extensively to investigate manipulator access in these complex, tightly packed, and cluttered surroundings. A recent refinement of computer modeling involves the use of an intelligent engineering work station for realtime interactive display of task simulations. This paper discusses the use of three-dimensional (3-D) kinematic computer models of the CIT machines that are proving to be powerful tools in our efforts to evaluate RM requirements. This presentation includes a video-taped simulation of remote replacement of a plasma viewing assembly. The simulation illustrates some of the constraints associated with typical RM activities and the ways in which computer modeling enhances the design process. 1 ref., 3 figs

  15. Design and development of the CRBRP ex-vessel transfer machine

    International Nuclear Information System (INIS)

    Jones, C.E. Jr.

    1977-01-01

    The Reactor Refueling System (RRS) for the Clinch River Breeder Reactor Project (CRBRP) uses the Ex-Vessel Transfer Machine (EVTM) for transferring core assemblies outside the reactor vessel. The design of the Ex-Vessel Transfer Machine (EVTM) and its gantry-trolly for the CRBRP is discussed. The development tests required for the design are presented, in conjunction with the impact of the test results on the design. The impact of the increased seismic requirements on the design are also presented

  16. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon, E-mail: rjpark@kaeri.re.kr; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-03-15

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m{sup 2} s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  17. Detailed evaluation of two phase natural circulation flow in the cooling channel of the ex-vessel core catcher for EU-APR1400

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Ha, Kwang-Soon; Rhee, Bo-Wook; Kim, Hwan Yeol

    2016-01-01

    Highlights: • Ex-vessel core catcher of PECS is installed in EU-APR1400. • CE-PECS has been conducted to test a cooling capability of the PECS. • Two phase flow in CE-PECS and PECS was analyzed using RELAP5/MOD3. • RELAP5 results are very similar to the CE-PECS data. • The super-step design is suitable for steam injection into the downcomer in PECS. - Abstract: The ex-vessel core catcher of the PECS (Passive Ex-vessel corium retaining and Cooling System) is installed to retain and cool down the corium in the reactor cavity of the EU (European Union)-APR (Advanced Power Reactor) 1400. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. Simulations of a two-phase natural circulation flow using the RELAP5/MOD3 computer code in the CE-PECS and PECS have been conducted to predict the two-phase flow characteristics, to determine the natural circulation mass flow rate in the cooling channel, and to evaluate the scaling in the experimental design of the CE-PECS. Particularly from a comparative study of the prototype PECS and the scaled test facility of the CE-PECS, the orifice loss coefficient in the CE-PECS was found to be 6 to maintain the coolant circulation mass flux, which is approximately 273.1 kg/m"2 s. The RELAP5 results on the coolant circulation mass flow rate are very similar to the CE-PECS experimental results. An increase in the coolant injection temperature and the heat flux lead to an increase in the coolant circulation mass flow rate. In the base case simulation, a lot of vapor was injected into the downcomer, which leads to an instability of the two-phase natural circulation flow. A super-step design at a downcomer inlet is suitable to prevent vapor injection into the downcomer piping.

  18. Corium crust strength measurements

    Energy Technology Data Exchange (ETDEWEB)

    Lomperski, S. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4840 (United States)], E-mail: lomperski@anl.gov; Farmer, M.T. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4840 (United States)], E-mail: farmer@anl.gov

    2009-11-15

    Corium strength is of interest in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the containment basemat. Some accident management strategies involve pouring water over the melt to solidify it and halt corium/concrete interactions. The effectiveness of this method could be influenced by the strength of the corium crust at the interface between the melt and coolant. A strong, coherent crust anchored to the containment walls could allow the yet-molten corium to fall away from the crust as it erodes the basemat, thereby thermally decoupling the melt from the coolant and sharply reducing the cooling rate. This paper presents a diverse collection of measurements of the mechanical strength of corium. The data is based on load tests of corium samples in three different contexts: (1) small blocks cut from the debris of the large-scale MACE experiments, (2) 30 cm-diameter, 75 kg ingots produced by SSWICS quench tests, and (3) high temperature crusts loaded during large-scale corium/concrete interaction (CCI) tests. In every case the corium consisted of varying proportions of UO{sub 2}, ZrO{sub 2}, and the constituents of concrete to represent a LWR melt at different stages of a molten core/concrete interaction. The collection of data was used to assess the strength and stability of an anchored, plant-scale crust. The results indicate that such a crust is likely to be too weak to support itself above the melt. It is therefore improbable that an anchored crust configuration could persist and the melt become thermally decoupled from the water layer to restrict cooling and prolong an attack of the reactor cavity concrete.

  19. In- and ex-vessel flooding as part of the severe accident strategy in the KERENA reactor

    International Nuclear Information System (INIS)

    Levi, P.; Fischer, M.

    2011-01-01

    Currently, AREVA NP is finalizing the basic design of the KERENA reactor, an advanced boiling water reactor with a net electric output of about 1250 MWe. The safety concept in the KERENA reactor is founded on reliable active and passive systems for water supply and heat removal. The passive systems are based on simple physics and do not require operator action. Therefore, a severe accident (SA) with core damage, caused by the subsequent and multiple failures of the safety systems, has an extremely low probability. Despite this, the KERENA design is intended to involve measures that can limit and stop the progression of the severe accident which further reduces the frequency and extent of radioactive releases into the environment. These additional measures include in-vessel and ex-vessel flooding. Flooding is intended to remove the heat from the core or from the reactor pressure vessel (RPV) and transfer it into the containment. There the heat is removed by the active RHR (residual heat removal) system or by the passive CCCs (containment cooling condensers). Both flooding measures are passive and actuated independent of each other by different signals. The study shows that the in-vessel flooding is capable of arresting the core melt progression before a large molten pool can develop. In the unlikely event that the passive in-vessel flooding cannot be actuated or fails, the core will melt and relocate into the lower head of the RPV. In this case, as a further line of defense, decay heat removal can be achieved through the RPV wall into the water in the cavity. In order to assess whether the ex-vessel cooling can ensure RPV wall integrity a dedicated thermodynamics code has been developed which considers heat transfer from the molten corium pool into the RPV wall and the resulting wall ablation. As an input for the code the stratification behavior of the oxidic and metallic phase of the molten pool is examined. In the case of a light metallic phase on top, high heat

  20. Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

  1. Analysis of corium recovery concepts by the EUROCORE group

    International Nuclear Information System (INIS)

    Seiler, J.-M.; Latrobe, A.; Sehgal, B.R.; Alsmeyer, H.; Kymaelaeinen, O.; Turland, B.; Grange, J.-L.; Fischer, M.; Azarian, G.; Buerger, M.; Cirauqui, C.J.; Zurita, A.

    2003-01-01

    The objective of the EUROCORE (European Group for Analysis of Corium Recovery Concepts) Concerted Action is to obtain a clear view of the state-of-the-art for melt stabilisation as considered in accident management schemes and to better identify Research and Development (R and D) needs. Five different melt stabilisation concepts have been discussed: in-vessel retention with external cooling, core-concrete interaction with top cooling, ex-vessel spreading with top flooding, water injection by bottom flooding, and crucible concept with sacrificial material. For each concept, main unresolved problems are discussed in this paper and recommended R and D actions are outlined. The project started on 1 March 2000 and ended on 28 February 2002

  2. Investigation of debris bed formation, spreading and coolability

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, P.; Konovalenko, A.; Grishchenko, D.; Yakush, S.; Basso, S.; Lubchenko, N.; Karbojian, A. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    The work is motivated by the severe accident management strategy adopted in Nordic type BWRs. It is assumed that core melt ejected from the vessel will fragment, quench and form a coolable debris bed in a deep water pool below the vessel. In this work we consider phenomena relevant to the debris bed formation and coolability. Several DEFOR-A (Debris Bed Formation - Agglomeration) tests have been carried out with new corium melt material and a melt releasing nozzle mockup. The influence of the melt material, melt superheat, jet free fall height on the (i) faction of agglomerated debris, (ii) particle size distribution, (iii) ablation/plugging of the nozzle mockup has been addressed. Results of the DECOSIM (Debris Coolability Simulator) code validation against available COOLOCE data are presented in the report. The dependence of DHF on system pressure from COOLOCE experiments can be reproduced quite accurately if either the effective particle diameter or debris bed porosity is increased. For a cylindrical debris bed, good agreement is achieved in DECOSIM simulations for the particle diameter 0.89 mm and porosity 0.4. The results obtained are consistent with MEWA simulation where larger particle diameters and porosities were found to be necessary to reproduce the experimental data on DHF. It is instructive to note that results of DHF prediction are in better agreement with POMECO-HT data obtained for the same particles. It is concluded that further clarification of the discrepancies between different experiments and model predictions. In total 13 exploratory tests were carried out in PDS (particulate debris spreading) facility to clarify potential influence of the COOLOCE (VTT) facility heaters and TCs on particle self-leveling process. Results of the preliminary analysis suggest that there is no significant influence of the pins on self-leveling, at least for the air superficial velocities ranging from 0.17 up to 0.52 m/s. Further confirmatory tests might be needed

  3. Investigation of debris bed formation, spreading and coolability

    International Nuclear Information System (INIS)

    Kudinov, P.; Konovalenko, A.; Grishchenko, D.; Yakush, S.; Basso, S.; Lubchenko, N.; Karbojian, A.

    2013-08-01

    The work is motivated by the severe accident management strategy adopted in Nordic type BWRs. It is assumed that core melt ejected from the vessel will fragment, quench and form a coolable debris bed in a deep water pool below the vessel. In this work we consider phenomena relevant to the debris bed formation and coolability. Several DEFOR-A (Debris Bed Formation - Agglomeration) tests have been carried out with new corium melt material and a melt releasing nozzle mockup. The influence of the melt material, melt superheat, jet free fall height on the (i) faction of agglomerated debris, (ii) particle size distribution, (iii) ablation/plugging of the nozzle mockup has been addressed. Results of the DECOSIM (Debris Coolability Simulator) code validation against available COOLOCE data are presented in the report. The dependence of DHF on system pressure from COOLOCE experiments can be reproduced quite accurately if either the effective particle diameter or debris bed porosity is increased. For a cylindrical debris bed, good agreement is achieved in DECOSIM simulations for the particle diameter 0.89 mm and porosity 0.4. The results obtained are consistent with MEWA simulation where larger particle diameters and porosities were found to be necessary to reproduce the experimental data on DHF. It is instructive to note that results of DHF prediction are in better agreement with POMECO-HT data obtained for the same particles. It is concluded that further clarification of the discrepancies between different experiments and model predictions. In total 13 exploratory tests were carried out in PDS (particulate debris spreading) facility to clarify potential influence of the COOLOCE (VTT) facility heaters and TCs on particle self-leveling process. Results of the preliminary analysis suggest that there is no significant influence of the pins on self-leveling, at least for the air superficial velocities ranging from 0.17 up to 0.52 m/s. Further confirmatory tests might be needed

  4. Experimental analysis of ex-vessel core catcher cooling system performance for EU-APR1400 during severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Song, K. W.; Park, H. S.; Revankar, S. T. [POSTECH, Pohang (Korea, Republic of); Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In the coolant channel which has a unique design and large scale flow paths, natural circulation is passively activated by buoyancy driven force. Since two-phase flow behavior in a large scale channel is different from that in a small scale channel, the two-phase flow affecting the cooling capability is difficult to be predicted in the large channel. Therefore, cooling experiment in the core catcher coolant path is necessary. Cooling Experiment - Passive Ex-vessel corium retaining and Cooling System(CE-PECS) is constructed in full scale(in height and width) slice of half prototype. It actually simulates steam-water flow in the coolant channel for different decay heat condition of the corium. In this study, thermal power considering of total amount of decay heat 190 kW which corresponds to 40MW of thermal power in the prototype is loaded on the top wall of the CE-PECS coolant channel. Natural circulation flow rate and pressure drops at the two-phase region are measured in various power level. Temperatures of heater block and working fluid in various position along the flow path enable to calculate heat fluxes and heat transfer coefficients distribution. These results are used for evaluating heat removal capability of core catcher facility. Two-phase natural circulation experiment is carried out in CE-PECS facility. Based on the prototypic condition, 190 kW of total power is supplied to the top of the coolant path. Uniform distribution of heat load on the downward facing heater bock produces -300 kW/m2 at 100 % power ratio. Although the experiment should consider the heat loss and heat flux uniformity, several noticeable conclusions have been made as followings; 1. Mass flow rate and two-phase pressure drop are measured in various power conditions. 2. Slightly inclined top wall at the downstream of the channel shows better heat exchange performance than horizontal top wall because enhanced convection due to the increase of void fraction improves local cooling. This

  5. Experimental study on in-vessel debris coolability during severe accident

    International Nuclear Information System (INIS)

    Kim, S. B.; Park, R. J.; Kim, H. D.

    2002-05-01

    A research program, called SONATA-IV(Simulation of Naturally Arrested Thermal Attack In-Vessel), has been performed to verify the gap cooling mechanism of corium in the lower plenum, and to develop management and mitigation strategies under severe accident conditions. For the proof-of-principles experiment, the LAVA(Lower-plenum Arrested Vessel Attack) experiments have been performed to gather proof of gap formation and to evaluate the gap effect on in-vessel cooling, using Al 2 O 3 /Fe (or Al 2 O 3 only) thermite melt as corium simulant. And also the CHFG(Critical Heat Flux in Gap) experiments have been performed to measure the critical power and to investigate the inherent cooling mechanism in the hemispherical narrow gap. In addition to the experiments, LILAC code was developed to analyze and predict the thermo-hydraulic phenomena of the corium relocated in the reactor lower plenum. It could be found from the LAVA and CHFG experimental results that continuous gap ranged from 1 to 5 mm was formed and that maximum heat removal capacity through a gap is a key factor in determining the potentials of the integrity of the vessel. After all the possibility of IVR(In-Vessel corium Retention) through gap cooling highly depends on the melt relocated into the lower plenum and the gap size. So, feasibility experiments have been performed for the assessment of improved IVR concepts using an internal engineered gap device and a dual strategy of In/Ex-vessel cooling using the LAVA facility. It is preliminarily concluded that these cooling measures lead to an enhanced cooling of the corium in the lower plenum of the reactor vessel. The additional studies will be performed to verify the quantitative heat removal capacity for these cooling measures in the 2nd phase of mid- and long term project period

  6. Analysis of ex-vessel steam explosion with MC3D

    International Nuclear Information System (INIS)

    Leskovar, M.; Mavko, B.

    2007-01-01

    An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which was developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central, left and right side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. The performed analysis shows that for some ex-vessel steam explosion scenarios significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. (author)

  7. Study of the rheological behaviour of corium/concrete mixtures

    International Nuclear Information System (INIS)

    Ramacciotti, M.

    1999-01-01

    In the hypothetical event of a severe accident in a Light Water Reactor, scenarios in which the reactor pressure vessel (RPV) fails and the core melt mixture (called corium) relocates into the reactor cavity, cannot be excluded. The viscosity (in fact, corium rheological behaviour) plays a major role in many phenomena such as core melt down, discharge from reactor pressure vessel, interaction with structural materials (concrete,...) and spreading in a core-catcher. For these reasons, it is important to be able to predict the rheological behaviour of corium melts of different compositions (essentially based on UO 2 , ZrO 2 , Fe x O y and Fe for in-vessel scenarios, plus SiO 2 and CaO for ex-vessel scenarios) at temperatures above solidus temperature. In the case of corium-concrete mixtures, the increase of viscosity depends not only on the increase of particles in the melts but also on the increase of the residual liquid phase viscosity (due to the increase in silica contents). The Urban correlation is used to calculate the viscosity of the carrying liquid with silica. This model was tested and gave good agreements between measured and estimated viscosities of various basalts among which one contained 18 wt% of UO 2 . Then, in the solidification range, the analysis of published data showed that the viscosity cannot be described by a suspension viscosity model of non-interactive spherical particles; consequently we proposed an Arrhenius type law with a multiplying factor such as η r = exp(2.5 Cφ) and the C factor value varies between 4 and 8. This factor is more important in the case of low shear rates and low cooling rates. The analysis of the samples structure after quenching shows a dependence of this factor on the particle morphology. Finally, for a value of 6.1 of the C factor, we obtained the best agreement with experimental data for a corium spreading test at 2100 K on a horizontal surface. (author)

  8. An experimental study on feasibility of ex-vessel cooling through the external guide vessel

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Kim, Jong-Hwan; Park, Rae-Jun; Kim, Sang-Baik

    2000-01-01

    This paper presents the results of a series of experiments for assessing the efficacy of ex-vessel cooling through the external guide vessel during a severe accident. Four tests were performed in the LAVA test facility at KAERI, varying the boundary conditions at the outer surface of the vessel. The first test was a dry condition test conducted without cooling the outside of the vessel. On the other hand, in the second test, the cooling of the vessel surface was produced by gravity-driven forced injection of water along the annular gap of 25 mm between the vessel and the external guide vessel. Water flow rate was about 0.85 kg/s and total mass of available water was 300 kg. For the evaluation of the water flow rate effect, the third test was performed with a pool type cooling in the annulus without any circulation of water. These two external cooling tests were performed under elevated pressure of about 1.6 MPa. Finally, the fourth test was conducted under atmospheric pressure to evaluate the effect of system pressure on boiling heat transfer characteristics. In the dry test and the pool type ex-vessel cooling test performed under atmospheric pressure, the vessel was failed by a melt penetration at about 40 degree upper position from the vessel bottom, which is coincident with the boundary of the Al 2 O 3 /Fe melt separated layers. On the other hand, in both of the ex-vessel cooling tests conducted under elevated pressure of about 1.6 MPa, the vessel didn't fail. Compared with the pool boiling test, the vessel experienced effective cooling due to the inlet flow in the forced flow test. Synthesized the results of the tests, it was shown that the heat removal with ex-vessel cooling through the guide vessel is feasible, but the additional evaluations should be performed to guarantee enough thermal margin. (author)

  9. Parametric model to estimate containment loads following an ex-vessel steam spike

    International Nuclear Information System (INIS)

    Lopez, R.; Hernandez, J.; Huerta, A.

    1998-01-01

    This paper describes the use of a relatively simple parametric model to estimate containment loads following an ex-vessel steam spike. The study was motivated because several PSAs have identified containment loads accompanying reactor vessel failures as a major contributor to early containment failure. The paper includes a detailed description of the simple but physically sound parametric model which was adopted to estimate containment loads following a steam spike into the reactor cavity. (author)

  10. Ex-vessel Fish Price Database: Disaggregating Prices for Low-Priced Species from Reduction Fisheries

    Directory of Open Access Journals (Sweden)

    Travis C. Tai

    2017-11-01

    Full Text Available Ex-vessel fish prices are essential for comprehensive fisheries management and socioeconomic analyses for fisheries science. In this paper, we reconstructed a global ex-vessel price database with the following areas of improvement: (1 compiling reported prices explicitly listed as “for reduction to fishmeal and fish oil” to estimate prices separately for catches destined for fishmeal and fish oil production, and other non-direct human consumption purposes; (2 including 95% confidence limit estimates for each price estimation; and (3 increasing the number of input data and the number of price estimates to match the reconstructed Sea Around Us catch database. Our primary focus was to address this first area of improvement as ex-vessel prices for catches destined for non-direct human consumption purposes were substantially overestimated, notably in countries with large reduction fisheries. For example in Peru, 2010 landed values were estimated as 3.8 billion real 2010 USD when using separate prices for reduction fisheries, compared with 5.8 billion using previous methods with only one price for all end-products. This update of the price database has significant global and country-specific impacts on fisheries price and landed value trends over time.

  11. In-vessel coolability and steam explosion in Nordic BWRs

    International Nuclear Information System (INIS)

    Ma, W.; Hansson, R.; Li, L.; Kudinov, P.; Cadinu, F.; Tran, C-.T.

    2010-05-01

    The INCOSE project is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in Nordic BWR plants with the cavity flooding as a severe accident management (SAM) measure. During 2009 substantial advances and new insights into physical mechanisms were gained for studies of: (i) in-vessel corium coolability - development of the methodologies to assess the efficiency of the control rod guide tube (CRGT) cooling as a potential SAM measure; (ii) debris bed coolability - characterization of the effective particle diameter of multi-size particles and qualification of friction law for two-phase flow in the beds packed with multi-size particles; and (iii) steam explosion - investigation of the effect of binary oxides mixtures properties on steam explosion. An approach for coupling of ECM/PECM models with RELAP5 was developed to enhance predictive fidelity for melt pool heat transfer. MELCOR was employed to examine the CRGT cooling efficiency by considering an entire accident scenario, and the simulation results show that the nominal flowrate (∼10kg/s) of CRGT cooling is sufficient to maintain the integrity of the vessel in a BWR of 3900 MWth, if the water injection is activated no later than 1 hour after scram. The POMECO-FL experimental data suggest that for a particulate bed packed with multi-size particles, the effective particle diameter can be represented by the area mean diameter of the particles, while at high velocity (Re>7) the effective particle diameter is closer to the length mean diameter. The pressure drop of two-phase flow through the particulate bed can be predicted by Reed's model. The steam explosion experiments performed at high melt superheat (>200oC) using oxidic mixture of WO3-CaO didn't detect an apparent difference in steam explosion energetics and preconditioning between the eutectic and noneutectic melts. This points out that the next step of MISTEE experiment will be conducted at lower superheat. (author)

  12. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Huh, Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)] (and others)

    1999-03-15

    Cooling methodologies for the molten corium resulted from the severe accident of the Nuclear Power Plant is suggested as one of most important items for the safety of the NPP. In this regard, considerable experimental and analytical works have been devoted. In the second phase of this project, current status of research about corium-concrete interaction and corium coolability which can occur on the reactor cavity has been surveyed, and the researches about lower head failure mechanism have also been surveyed. And, severe accident analysis for Ulchin 3 and 4 has been conducted, and collapse load of lower head has been analyzed through structural analysis considering various heat transfer conditions. The results of accident analysis can be used as a basic input for structural analysis which will be conducted in 3rd phase of this study.

  13. Experimental Study on the Molten Corium Interaction with Structure by Induction Heating Technique

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang Mo; Ha, Kwang Soon; Min, Beong Tae; Hong, Seong Ho; Kim, Hwan Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The corium compositions strongly depend on the accident scenarios, and thus the melt generation technique for various melt compositions is essential to investigate the corium-structural material interaction characteristics according to the accident scenarios. Since 1997, KAERI has several years of experiences with melt generation to investigate the material ablation characteristics and steam explosion phenomena. Based on the experiences of the TROI (Test for Real cOrium Interaction with water) facility for the steam explosion experiments, the VESTA (Verification of Ex-vessel corium STAbilization) test facility was designed and constructed in 2010 for the development of a core catcher under the APR+ project. At the same time, the VESTA-S (VESTA-Small) was established for small scale material ablation experiments. Some experimental results were reported for the interactions of metallic or oxidic melt with the structural materials such as special concrete or penetration weld. The objective of this paper is to provide the specific features of the VESTA and VESTA-S facilities including information on the melt generation technique adopted for the facilities. Some issues are also addressed in this paper for further facility improvement. In the present paper, the principles of induction heating adopted for the VESTA and VESTA-S facilities were summarized briefly and the system features for the melt-structural material interaction experiments were explained. As a major characteristic of the VESTA facility, up to 400 kg of corium melt is expected to be generated using the currently installed system. The jet impingement effect on the material ablation characteristics was demonstrated successfully in the VESTA facility. In the VESTA-S facility, the small scale material ablation experiments by long term melt interaction were performed properly by adopting the melt delivery method. However, for a more realistic severe accident simulation, we need to improve the melt temperature

  14. Experimental investigation of multidimensional cooling effects on the coolability of a debris bed

    International Nuclear Information System (INIS)

    Rashidi, M.; Kulenovici, R.; Laurieni, E.

    2011-01-01

    During a severe accident in a light water reactor, the core can melt and be relocated to the lower plenum of the reactor pressure vessel. There it can form a particulate debris bed due to the possible presence of water. Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long-term coolability of debris beds, the down-scaled non nuclear test facility DEBRIS has been established at IKE. The major objectives of the experimental investigations at this test facility are the determination of local pressure drops for steady state boiling to check friction laws, the determination of dryout heat fluxes under various conditions for validation of numerical models, and the analysis of quenching processes of dry hot debris beds. A large number of 1D-experiments were carried out to investigate the coolability limits for different bed configurations at various thermohydraulic conditions, and to validate numerical models which can be used in reactor safety studies. Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favored. This paper presents 2D experimental results, based on various kinds of water inflow conditions into the bed, boiling and dryout tests with different bed configurations and different system pressures. Preliminary results show that the system pressure has no significant effect on the fundamental shape of the pressure gradient inside the bed, whereas with increasing system pressure the coolability limits are increased

  15. Optimized design of an ex-vessel cooling thermosyphon for decay heat removal in SFR

    International Nuclear Information System (INIS)

    Choi, Jae Young; Jeong, Yong Hoon; Song, Sub Lee; Chang, Soon Heung

    2017-01-01

    Passive decay heat removal and sodium fire are two major key issues of nuclear safety in sodium-cooled fast reactor (SFR). Several decay heat removal systems (DHR) were suggested for SFR around the world so far. Those DHRS mainly classified into two concepts: Direct reactor cooling system and ex-vessel cooling system. Direct reactor cooling method represented by PDHRS from PGSFR has disadvantages on its additional in-vessel structure and potential sodium fire risk due to the sodium-filled heat exchanger exposed to air. Contrastively, ex-vessel cooling method represented by RVACS from PRISM has low decay heat removal performance, which cannot be applicable to large scale reactors, generally over 1000 MWth. No passive DHRSs which can solve both side of disadvantages has been suggested yet. The goal of this study was to propose ex-vessel cooling system using two-phase closed thermosyphon to compensate the disadvantages of the past DHRSs. Reference reactor was Innovative SFR (iSFR), a pool-type SFR designed by KAIST and featured by extended core lifetime and increased thermal efficiency. Proposed ex-vessel cooling system consisted of 4 trains of thermosyphons and designed to remove 1% of thermal power with 10% of margin. The scopes of this study were design of proposed passive DHRS, validation of system analysis and optimization of system design. Mercury was selected as working fluid to design ex-vessel thermosyphon in consideration of system geometry, operating temperature and required heat flux. SUS 316 with chrome coated liner was selected as case material to resist against high corrosivity of mercury. Thermosyphon evaporator was covered on the surface of reactor vessel as the geometry of hollow shell filled with mercury. Condenser was consisted of finned tube bundles and was located in isolated water pool, the ultimate heat sink. Operation limits and thermal resistance was estimated to guarantee whether the design was adequate. System analysis was conducted by in

  16. Design, fabrication and operating experience of Monju ex-vessel fuel storage tank

    International Nuclear Information System (INIS)

    Yokota, Yoshio; Yamagishi, Yoshiaki; Kuroha, Mitsuo; Inoue, Tatsuya

    1995-01-01

    In FBRs there are two methods of storing and cooling the spent fuel - the in-vessel storage and the ex-vessel storage. Because of the sodium leaks through the tank at the beginning of pre-operation, the utilization of the ex-vessel fuel storage tank (EVST) of some FBR plant has been changed from the ex-vessel fuel storage to the interim fuel transfer tank. This led to reactor designers focusing on the material, structure and fabrication of the carbon steel sodium storage tanks worldwide. The Monju EVST was at the final stage of the design, when the leaks occurred. The lesson learned from that experience and the domestic fabrication technology are reflected to the design and fabrication of the Monju EVST. This paper describes the design, fabrication and R and D results for the tank, and operating experience in functional test. The items to be examined are as follows: (1) Overall structure of the tank and design philosophy on the function, (2) Structure of the cover shielding plug and its design philosophy, (3) Structures of the rotating rack and its bearings, and their design philosophy, (4) Cooling method and its design philosophy, (5) Structure and fabrication of the cooling coil support inside EVST with comparison of leaked case, (6) R and D effort for items above. The fabrication of the Monju EVST started in August 1986 and it was shipped to the site in March 1990. Installation was completed in November 1990, and sodium fill after pre-heating started in 1991. The operation has been continued since September 1992. In 1996 when the first spent fuel is stored, its total functions will be examined. (author)

  17. Preliminary Analysis of Ex-Vessel Steam Explosion using TEXAS-V code for APR1400

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Chu; Lee, Jung Jae; Cho, Yong Jin; Hwang, Taesuk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    The purpose of this study is to explore input development and the audit calculation using TEXAS-V code for ex-vessel steam explosion for a flooded reactor cavity of APR1400. TEXAS computational models are one of the simplified tools for simulations of fuel-coolant interaction during mixing, triggering and explosion phase. The models of TEXAS code were validated by performing the fundamental experimental investigation in the KROTOS facility at JRC, Ispra. The experiments such as KROTOS and FARO experiment are aimed at providing benchmark data to examine the effect of fuel-coolant initial conditions and mixing on explosion energetics with alumina and prototypical core material. TEXAS-V code used in this study was to analyze and predict the ex-vessel steam explosion for a reactor scale. The input deck to simulate the flooded reactor cavity of APR1400 is developed and base case calculation is performed. This study will provide a base for further study. The code will be of use for the evaluation and sensitivity study of ex-vessel steam explosion for ERVC strategy in the future studies. Analysis result of this study is similar to the result of other study. Through this study, it is found that TEXAS-V could be the used as a tool for predicting the steam explosion load on a reactor scale, as fast running computer code. In addition, TEXAS-V code could be to evaluate the impact of various uncertainties, which are not clearly understood yet, to provide a conservative envelope for the steam explosion.

  18. Preliminary Analysis of Ex-Vessel Steam Explosion using TEXAS-V code for APR1400

    International Nuclear Information System (INIS)

    Song, Sung Chu; Lee, Jung Jae; Cho, Yong Jin; Hwang, Taesuk

    2013-01-01

    The purpose of this study is to explore input development and the audit calculation using TEXAS-V code for ex-vessel steam explosion for a flooded reactor cavity of APR1400. TEXAS computational models are one of the simplified tools for simulations of fuel-coolant interaction during mixing, triggering and explosion phase. The models of TEXAS code were validated by performing the fundamental experimental investigation in the KROTOS facility at JRC, Ispra. The experiments such as KROTOS and FARO experiment are aimed at providing benchmark data to examine the effect of fuel-coolant initial conditions and mixing on explosion energetics with alumina and prototypical core material. TEXAS-V code used in this study was to analyze and predict the ex-vessel steam explosion for a reactor scale. The input deck to simulate the flooded reactor cavity of APR1400 is developed and base case calculation is performed. This study will provide a base for further study. The code will be of use for the evaluation and sensitivity study of ex-vessel steam explosion for ERVC strategy in the future studies. Analysis result of this study is similar to the result of other study. Through this study, it is found that TEXAS-V could be the used as a tool for predicting the steam explosion load on a reactor scale, as fast running computer code. In addition, TEXAS-V code could be to evaluate the impact of various uncertainties, which are not clearly understood yet, to provide a conservative envelope for the steam explosion

  19. Evaluation of upward heat flux in ex-vessel molten core heat transfer using MELCOR

    International Nuclear Information System (INIS)

    Park, S.Y.; Park, J.H.; Kim, S.D.; Kim, D.H.; Kim, H.D.

    2000-01-01

    The purpose of this study is to share experiences of MELCOR application to resolve the molten corium-concrete interaction (MCCI) issue in the Korea Next Generation Reactor (KNGR). In the evaluation of concrete erosion, the heat transfer modeling from the molten corium internal to the corium pool surface is very important and uncertain. MELCOR employs Kutateladze or Greene's bubble-enhanced heat transfer model for the internal heat transfer. The phenomenological uncertainty is so large that the model provides several model parameters in addition to the phenomenological model for user flexibility. However, the model parameters do not work on Kutateladze correlation at the top of the molten layer. From our experience, a code modification is suggested to match the upward heat flux with the experimental results. In this analysis, minor modification was carried out to calculate heat flux from the top molten layer to corium surface, and efforts were made to find out the best value of the model parameter based on upward heat flux of MACE test M1B. Discussion also includes its application to KNGR. (author)

  20. A simple evaluation of containment integrity against ex-vessel steam explosion

    International Nuclear Information System (INIS)

    Nishiura, Hiroshi

    2000-01-01

    The guideline for consideration to severe accidents on containment design for next-generation LWR was published in 1999. In order to verify the validity of future containment designs, we have developed a method of assessing for the containment integrity against ex-vessel steam explosion. First, we conducted a simple evaluation on an Advanced PWR. The strength of the reactor cavity wall was assumed to be equivalent to the total strain energy which would accumulate by the time one reinforcing bar element would first reach the failure strain in FEM analyses. As a result, the strength was evaluated to be about 72 MJ. The explosion energy was assumed to be a function of the mass of the dropping melted core and the conversion ratio. Assuming the conversion ratio of 1%, it was estimated that the explosion energy would amount to about 1 MJ if the melt mass corresponds to the break of one instrumentation guide tube penetration, and about 40 MJ if the mass corresponds to the simultaneous break of all penetrations. Therefore, it is expected that the explosion energy would be less than the wall strength; thus, the containment integrity would be maintained even if an ex-vessel steam explosion were to occur. (author)

  1. Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

    1995-01-01

    Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

  2. Ex-vessel remote maintenance design for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) imposes a requirement for remote handling technology for ex-vessel maintenance operations on auxiliary machine components. These operations consist of repairing and replacing components such as diagnostic, radio-frequency (rf) heating, and fueling systems using remotely operated maintenance equipment in the test cell. In addition, ex-vessel maintenance design also includes developing hot cell facilities for equipment decontamination, repair, and solid radioactive waste handling. The test cell maintenance philosophy is markedly influenced by the neutron/gamma shield surrounding the machine that allows personal access into the test cell one day after shutdown. Hence, maintenance operations can be performed hands-on in the test cell with the shield intact and must be remotely performed when the shield is disassembled for machine access. The constricted access to the auxiliary components of the machine affect the design requirements for the maintenance equipment and impose major spatial constraints. Several major areas of the maintenance system design are being addressed in fiscal year 1987. These include conceptual design of the manipulator system, preliminary remote equipment research and development, and definition of the hot cell, decontamination, and equipment repair facility requirements. The manipulator work includes investigating transporters and viewing/lighting subsystems. 2 figs

  3. Experimental Validation of Ex-Vessel Neutron Spectrum by Means of Dosimeter Materials Activation Method

    Directory of Open Access Journals (Sweden)

    S.A. Santa

    2017-06-01

    Full Text Available Neutron spectrum information in reactor core and around of ex-vessel reactor needs to be known with a certain degree of accuracy to support the development of fuels, materials, and other components. The most common method to determine neutron spectra is by utilizing the radioactivation of dosimeter materials. This report presents the evaluation of neutron flux incident on M3dosimeter sets which were irradiated outside the reactor vessel,as well as the validation of  neutron spectrum calculation. Al capsules containing both dosimeter set covered withCd and dosimeter set without Cd cover have been irradiated during the 35th operational cycle in the M3 ex-vessel irradiation hole position207 cmfrom core centerline at the space between the reactor vessel and the safety vessel. The capsules were positioned at Z=0.0 cm of core midplane. Each dosimeter set consists of Co-Al, Sc, Fe, Np, Nb, Ni, B, and Ta. The gamma-ray spectra of irradiated dosimeter materials were measured by 63 cc HPGe solid-state detector and photo-peak spectra were analyzed using BOB75 code. The reaction rates of each dosimeter materials and its uncertainty were analyzed based on 59Co (n,g 60Co, 237Np (n,f 95Zr-103Ru,  45Sc (n,g 46Sc, 58Fe (n,g 59Fe, 181Ta (n,g 182Ta, and 58Ni (n,p58Co reactions. The measured Cd ratios indicate that neutron spectrum at the irradiated dosimeter sets was dominated by low energy neutron. The experimental result shows that the calculated neutron spectra by DORT code at the ex-vessel positions need correction, especially in the fast neutron energy region, so as to obtain reasonable unfolding result consistent with the reaction rate measurement without any exception. Using biased DORT initial spectrum, the neutron spectrum and its integral quantity were unfolded by NEUPAC code. The result shows that total neutron flux, flux above 1.0 MeV, flux above 0.1 MeV, and the displacement rate of the dosimeter set not covered with Cd were 1.75× 1012 n cm2 s-1, 1

  4. PWR neutron ex-vessel detection calculations using three-dimensional codes

    International Nuclear Information System (INIS)

    Dekens, O.; Lefebvre, J.C.; Rohart, M.; Chiron, M.

    1997-01-01

    During the accident of TM12, the signal delivered by source detectors was exceptionally high. This phenomenon was found out to be due to the water inventory in the primary system. Thus, in their research activity, Electricite de France (EdF) and Commissariat a l'Energie Atomique (CEA) have jointly launched a programme, whose aim was to determine to what extent the response of ex-vessel neutron detectors are representative of reactor water level (or sources positions) in a French 900 MWe PWR. In this framework, both partners developed the methods needed for each step of the calculation chain. Finally, a simulation of a LOCA indicates that the loss of coolant can be detected by existing monitoring system, and could be more efficiently found by changing the position of the source range detectors. (authors)

  5. Ex-vessel break in ITER divertor cooling loop analysis with the ECART code

    CERN Document Server

    Cambi, G; Parozzi, F; Porfiri, MT

    2003-01-01

    A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal-hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal-hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed.

  6. Proposal of Ex-Vessel dosimetry for pressure vessel Atucha II

    International Nuclear Information System (INIS)

    Chiaraviglio, N.; Bazzana, S.

    2013-01-01

    Nuclear reactor dosimetry has the purpose of guarantee that changes in material mechanical properties of critical materials do not compromise the reactor safety. In PWR in which the top of the reactor vessel is open once a year, is possible to use Charpy specimens to measure the change in mechanical properties. Atucha II nuclear power plant is a reactor with on-line refueling so there is no access to the inside of the pressure vessel. Because of this, ex-vessel dosimetry must be performed and mechanical properties changes must be inferred from radiation damage estimations. This damage can be calculated using displacement per atom cross sections and a transport code such as MCNP. To increase results reliability it is proposed to make a neutron spectrum unfolding using activation dosimeters irradiated during one operation cycle of the power plant. In this work we present a dosimetry proposal for such end, made in base of unfolding procedures and experimental background. (author) [es

  7. Ex-vessel water-level and fission-product monitoring for LWR

    International Nuclear Information System (INIS)

    DeVolpi, A.; Markoff, D.

    1988-01-01

    Given that the need for direct measurement of reactor coolant inventory under operational or abnormal conditions remains unsatisfied, a high-energy gamma-ray detection system is described for ex-vessel monitoring. The system has been modeled to predict response in a PWR, and the model has been validated with a LOFT LOCA sequence. The apparatus, situated outside the pressure vessel, would give relative water level and density over the entire vessel height and distinguish differing levels in the downcomer and core. It would also have significant sensitivity after power shutdown because of high-energy gamma rays from photoneutron capture, the photoneutrons being the result of fission-product decay in the core. Fission-products released to the coolant and accumulated in the top of a PWR vessel would also be theoretically detectable

  8. Characterization of photo-multiplier tube as ex-vessel radiation detector in tokamak

    Science.gov (United States)

    Jo, Jungmin; Cheon, MunSeong; Kim, Junghee; An, YoungHwa; Park, Seungil; Chung, Kyoung-Jae; Hwang, Y. S.

    2017-09-01

    Feasibility of using conventional photo-multiplier tubes (PMTs) without a scintillator as an ex-vessel radiation detector in a tokamak environment is studied. Basic irradiation tests using standard gamma ray sources and a d-d neutron generator showed that the PMT is responding both to gamma photons and neutrons, possibly due to the direct generation of secondary electrons inside the PMT by the impingement of high energy photons. Because of the selective sensitivity of the PMT to hard x-ray and neutrons in ohmic and neutral beam injected plasmas, respectively, it is shown that the PMT with certain configuration can be utilized either to monitor the fluctuation in the fusion neutron generation rate or to study the behavior of runaway electrons in tokamaks.

  9. Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

    Energy Technology Data Exchange (ETDEWEB)

    Almstroem, H.; Sundel, T. [National Defence Research Establishment, Stockholm (Sweden); Frid, W.; Engelbrektson, A.

    1998-01-01

    When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion at the rigid wall is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied, and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to about 40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (author)

  10. Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

    Energy Technology Data Exchange (ETDEWEB)

    Almstroem, H.; Sundel, T. (Nat. Defence Res. Establ., Tumba (Sweden)); Frid, W. (Swedish Nuclear Power Inspectorate, SE-10658, Stockholm (Sweden)); Engelbrektson, A. (VBB/SWECO, Box 34044, SE-10026, Stockholm (Sweden))

    1999-05-01

    When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion, at the rigid wall, is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to [approx]40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (orig.) 5 refs.

  11. Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

    International Nuclear Information System (INIS)

    Almstroem, H.; Sundel, T.; Frid, W.; Engelbrektson, A.

    1999-01-01

    When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion, at the rigid wall, is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to ∼40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work. (orig.)

  12. The results of the CCI-3 reactor material experiment investigating 2-D core-concrete interaction and debris coolability with a siliceous concrete crucible

    International Nuclear Information System (INIS)

    Farmer, M.T.; Basu, S.

    2006-01-01

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs for future plants. Despite years of international research, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion predicted by codes such as MELCOR, WECHSL, and COSACO. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test involved the interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined ablation depth was reached, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a summary description of the test facility and an overview of test results

  13. LACOMERA - large scale experiments on core degradation, melt retention and coolability at the Forschungszentrum Karslruhe

    International Nuclear Information System (INIS)

    Miassoedov, A.; Alsmeyer, H.; Meyer, L.

    2003-01-01

    The LACOMERA project at the Forschungszentrum Karlsruhe is a 3 year shared-cost action within the Fifth Framework Programme which started in September 2002. The overall objectives of the LACOMERA project are to provide research institutions from the EU member countries and associated states access to large scale experimental facilities at the Forschungszentrum Karlsruhe which shall be used to increase the knowledge of the quenching of a degraded core and regaining melt coolability in the reactor pressure vessel, of possible melt dispersion to the cavity, of molten core concrete interaction and of ex-vessel melt coolability. One major aspect is to understand how these events affect the safety of European reactors so as to lead to soundly-based accident management procedures. The project will bring together interested partners of different European member states in the area of severe accident analysis and control, with the goal to increase the public confidence in the use of nuclear energy. Moreover, partners from the newly associated states should be included as far as possible, and therefore the needs of Eastern, as well as Western, reactors will be considered in LACOMERA project. The project offers a unique opportunity to get involved in the networks and activities supporting VVER safety, and for Eastern experts to get an access to large scale experimental facilities in a Western research organisation to improve understanding of material properties and core behaviour under severe accident conditions. As a result of the first call for proposals a project on air ingress test in the QUENCH facility has been selected. A second call for proposals is opened with a deadline of 31 December 2003. (author)

  14. Study of the rheological behaviour of corium/concrete mixtures; Etude du comportement rheologique de melanges issus de l'interaction corium/beton

    Energy Technology Data Exchange (ETDEWEB)

    Ramacciotti, M

    1999-09-24

    In the hypothetical event of a severe accident in a Light Water Reactor, scenarios in which the reactor pressure vessel (RPV) fails and the core melt mixture (called corium) relocates into the reactor cavity, cannot be excluded. The viscosity (in fact, corium rheological behaviour) plays a major role in many phenomena such as core melt down, discharge from reactor pressure vessel, interaction with structural materials (concrete,...) and spreading in a core-catcher. For these reasons, it is important to be able to predict the rheological behaviour of corium melts of different compositions (essentially based on UO{sub 2}, ZrO{sub 2}, Fe{sub x}O{sub y} and Fe for in-vessel scenarios, plus SiO{sub 2} and CaO for ex-vessel scenarios) at temperatures above solidus temperature. In the case of corium-concrete mixtures, the increase of viscosity depends not only on the increase of particles in the melts but also on the increase of the residual liquid phase viscosity (due to the increase in silica contents). The Urban correlation is used to calculate the viscosity of the carrying liquid with silica. This model was tested and gave good agreements between measured and estimated viscosities of various basalts among which one contained 18 wt% of UO{sub 2}. Then, in the solidification range, the analysis of published data showed that the viscosity cannot be described by a suspension viscosity model of non-interactive spherical particles; consequently we proposed an Arrhenius type law with a multiplying factor such as {eta}{sub r} = exp(2.5 C{phi}) and the C factor value varies between 4 and 8. This factor is more important in the case of low shear rates and low cooling rates. The analysis of the samples structure after quenching shows a dependence of this factor on the particle morphology. Finally, for a value of 6.1 of the C factor, we obtained the best agreement with experimental data for a corium spreading test at 2100 K on a horizontal surface. (author)

  15. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  16. OECD/NEA SERENA Project for a Resolution of Ex-vessel Steam Explosion Risks

    International Nuclear Information System (INIS)

    Hong, S. W.; Kim, J. H.; Min, B. T.; Park, I. K.; Ha, K. S.; Hong, S. H.; Song, J. H.; Lee, J. Y.; Kim, H. D.

    2008-01-01

    Korea Atomic Energy Research Institute (KAERI) has conducted the TROI (Test for Real cOrium Interaction with water) program for a study on a fuel coolant interaction (FCI) since 2001. More than 50 experiments using several prototypic materials have been carried out so far. SERENA phase 2 project which has been conducting since 1st Oct. 2007 is aimed a the resolution of the uncertainties on the void fraction and the melt composition effect by performing a limited number of well-designed tests with advanced instrumentations to clarify the nature of a prototypic material with mild steam explosion characteristics and to provide innovative experimental data for a computer code validation

  17. Development and validation of corium oxidation model for the VAPEX code

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, V.I.; Davydov, M.V.; Melikhov, O.I.; Borovkova, E.M.

    2011-01-01

    In light water reactor core melt accidents, the molten fuel (corium) can be brought into contact with coolant water in the course of the melt relocation in-vessel and ex-vessel as well as in an accident mitigation action of water addition. Mechanical energy release from such an interaction is of interest in evaluating the structural integrity of the reactor vessel as well as of the containment. Usually, the source for the energy release is considered to be the rapid transfer of heat from the molten fuel to the water ('vapor explosion'). When the fuel contains a chemically reactive metal component, there could be an additional source for the energy release, which is the heat release and hydrogen production due to the metal-water chemical reaction. In Electrogorsk Research and Engineering Center the computer code VAPEX (VAPor EXplosion) has been developed for analysis of the molten fuel coolant interaction. Multifield approach is used for modeling of dynamics of following phases: water, steam, melt jet, melt droplets, debris. The VAPEX code was successfully validated on FARO experimental data. Hydrogen generation was observed in FARO tests even though corium didn't contain metal component. The reason for hydrogen generation was not clear, so, simplified empirical model of hydrogen generation was implemented in the VAPEX code to take into account input of hydrogen into pressure increase. This paper describes new more detailed model of hydrogen generation due to the metal-water chemical reaction and results of its validation on ZREX experiments. (orig.)

  18. Modeling of melt retention in EU-APR1400 ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V. S.; Sulatsky, A. A.; Khabensky, V. B.; Sulatskaya, M. B. [Alexandrov Research Inst. of Technology NITI, Sosnovy Bor (Russian Federation); Gusarov, V. V.; Almyashev, V. I.; Komlev, A. A. [Saint Petersburg State Technological Univ. SPbSTU, St.Petersburg (Russian Federation); Bechta, S. [KTH, Stockholm (Sweden); Kim, Y. S. [KHNP, 1312 Gil 70, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Park, R. J.; Kim, H. Y.; Song, J. H. [KAERI, 989 Gil 111, Daedeokdaero, Yuseong-gu, Daejeon (Korea, Republic of)

    2012-07-01

    A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)

  19. Atucha I nuclear power plant azimuthal ex-vessel flux profile evaluation

    International Nuclear Information System (INIS)

    Ferraro, Diego

    2008-01-01

    Irradiation damage in RPV (Reactor Pressure Vessel) in nuclear power plants is a key parameter to be analyzed in order to assess the plant integrity up to end of life and planning for a possible plant life extension. In this work a neutronic model in MCNP that represents a sector of 30 degrees of the Atucha I power plant nucleus has been consolidated with the results of an ex-vessel dosimetry made in the outer surface of the RPV s power plant in order to analyse the irradiation damage through the dpa rate. A strong dependents of the maximum point of damage with the loading of a peripheral channel was found, so a mitigation strategy was proposed, which is basically to empty this channel and its analogs in the rest of the nucleus. Analysing this second case a notable decrease of the damage is found in the zone considerated on the model (shown through the drop of de dpa rate in the zone). [es

  20. Feasibility studies on plasma vertical position control by ex-vessel coils in ITER-like tokamak fusion reactors

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Sugihara, Masayoshi; Shimomura, Yasuo

    1993-01-01

    Feasibility of the plasma vertical position control by control coils installed outside the vacuum vessel (ex-vessel) in a tokamak fusion reactor is examined for an ITER-like device. When a pair of ex-vessel control coils is made of normal conductor material and located near the outmost superconducting (SC) poloidal field (PF) coils, the applied voltage of several hundred volts on the control coils is the maximum allowable value which is limited by the maximum allowable induced voltage and eddy current heating on the SC PF coils, under the conditions that the SC PF coils are connected in series and a partitioning connection is employed for each of these PF coils. A proportional and derivative (PD) controller with and without voltage limitation has been employed to examine the feasibility. Indices of settling time and overshoot are introduced to measure the controllability of the control system. Based on these control schemes and indices, higher elongation (κ=2) and moderate elongation (κ=1.6) plasmas are examined for normal and deteriorated (low beta value and peaked current profile) plasma conditions within the restriction of applied voltage and current of control coils. The effect of the time constant of the passive stabilizer is also examined. The major results are: (1) A plasma with an elongation of 2.0 inevitably requires a passive stabilizer close to the plasma surface, (2) in case of a higher elongation than κ=2, even the ex-vessel control coil system is marginally controllable under normal plasma conditions, while it is difficult to control the deteriorated plasma conditions, (3) the time constant of the passive stabilizer is not an essential parameter for the controllability, (4) when the elongation is reduced down to 1.6, the ex-vessel control coil system can control the plasma even under deteriorated plasma conditions. (orig.)

  1. Oxidation kinetics of corium pool

    Energy Technology Data Exchange (ETDEWEB)

    Sulatsky, A.A., E-mail: andrei314@mail.ru [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Smirnov, S.A. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA), St. Petersburg (Russian Federation); Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Kotova, S.Yu. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Fischer, M.; Hellmann, S. [AREVA NP GmbH, Erlangen (Germany); Tromm, W.; Miassoedov, A. [Forschungzentrum Karlsruhe (FZK), Karlsruhe (Germany); Bottomley, D. [EUROPÄISCHE KOMMISSION, Joint Research Centre Institut für Transurane (ITU), Karlsruhe (Germany); Piluso, P. [CEA Cadarache-DEN/DTN/STRI, St.Paul-lez-Durance (France); Barrachin, M. [Institut de Radioprotection et Sûreté Nucléaire, St.Paul-lez-Durance (France)

    2013-09-15

    Highlights: • The analysis of experimental data on molten corium oxidation was been carried out. • The analysis has revealed the main factors influencing the oxidation kinetics. • The analysis was used for developing a qualitative analytical model. • The numerical modeling has confirmed the results of experimental data analysis. -- Abstract: Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations.

  2. Oxidation kinetics of corium pool

    International Nuclear Information System (INIS)

    Sulatsky, A.A.; Smirnov, S.A.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Kotova, S.Yu.; Fischer, M.; Hellmann, S.; Tromm, W.; Miassoedov, A.; Bottomley, D.; Piluso, P.; Barrachin, M.

    2013-01-01

    Highlights: • The analysis of experimental data on molten corium oxidation was been carried out. • The analysis has revealed the main factors influencing the oxidation kinetics. • The analysis was used for developing a qualitative analytical model. • The numerical modeling has confirmed the results of experimental data analysis. -- Abstract: Experimental, theoretical and numerical studies of oxidation kinetics of an open surface corium pool have been reported. The experiments have been carried out within OECD MASCA program and ISTC METCOR, METCOR-P and EVAN projects. It has been shown that the melt oxidation is controlled by an oxidant supply to the melt free surface from the atmosphere, not by the reducer supply from the melt. The project experiments have not detected any input of the zirconium oxidation kinetics into the process chemistry. The completed analysis puts forward a simple analytical model, which gives an explanation of the main features of melt oxidation process. The numerical modeling results are in good agreement with experimental data and theoretical considerations

  3. Numerical Analysis on Heat Flux Distribution through the Steel Liner of the Ex-vessel Core Catcher

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se Hong; Choi, Choeng Ryul [ELSOLTEC, Yongin (Korea, Republic of); Kim, Byung Jo; Lee, Kyu Bok [KEPCO, Gimcheon (Korea, Republic of); Hwang, Do Hyun [KHNP-CRI, Daejeon (Korea, Republic of)

    2016-05-15

    In order to prevent material failure of steel container of the core catcher system due to high temperatures, heat flux through the steel liner wall must be kept below the critical heat flux (CHF), and vapor dry-out of the cooling channel must be avoided. In this study, CFD methodology has been developed to simulate the heat flux distribution in the core catcher system, involving following physical phenomena: natural convection in the corium pool, boiling heat transfer and solidification/melting of the corium. A CFD methodology has been developed to simulate the thermal/hydraulic phenomena in the core catcher system, and a numerical analysis has been carried out to estimate the heat flux through the steel liner of the core catcher. High heat flux values are formed at the free surface of the corium pool. However, the heat flux through the steel liner is maintained below the critical heat flux.

  4. Revisiting the analysis of passive plasma shutdown during an ex-vessel loss of coolant accident in ITER blanket

    International Nuclear Information System (INIS)

    Rivas, J.C.; Dies, J.; Fajarnés, X.

    2015-01-01

    Highlights: • We have repeated the safety analysis for the hypothesis of passive plasma shutdown for beryllium evaporation during an ex-vessel LOCA of ITER first wall, with AINA code. • We have performed a sensitivity analysis over some key parameters that represents uncertainties in physics and engineering, to identify cliff edge effects. • The obtained results for the 500 MW inductive scenario, with an ex-vessel LOCA affecting a third of first wall surface are similar to those of previous studies and point to the possibility of a passive plasma shutdown during this safety case, before a serious damage is inflicted to the ITER wall. • The sensitivity analysis revealed a new scenario potentially damaging for the first wall if we increase fusion power and time delay for impurity transport, and decrease fraction of affected first wall area and initial beryllium fraction in plasma. • After studying the 700 MW inductive scenario, with an ex-vessel LOCA affecting 10% of first wall surface, with 0.5% of Be in plasma and a time delay twice the energy confinement time, it was found that affected area of first wall would melt before a passive plasma shutdown occurs. - Abstract: In this contribution, the analysis of passive safety during an ex-vessel loss of coolant accident (LOCA) in the first wall/shield blanket of ITER has been studied with AINA safety code. In the past, this case has been studied using robust safety arguments, based on simple 0D models for plasma balance equations and 1D models for wall heat transfer. The conclusion was that, after first wall heating up due to the loss of all coolant, the beryllium evaporation in the wall surface would induce a growing impurity flux into core plasma that finally would end in a passive shut down of the discharge. The analysis of plasma-wall transients in this work is based in results from AINA code simulations. AINA (Analyses of IN vessel Accidents) code is a safety code developed at Fusion Energy Engineering

  5. The Results of the CCI-3 Reactor Material Experiment Investigating 2-D Core-Concrete Interaction and Debris Coolability with a Siliceous Concrete Crucible

    International Nuclear Information System (INIS)

    Farmer, M.T.; Lomperski, S.; Basu, S.

    2006-01-01

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program conducted reactor materials experiments and associated analysis to achieve the following two objectives: 1) resolve the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs of future plants. With respect to the second objective, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion profiles predicted by codes such as WECHSL, COSACO, TOLBIAC, MEDICIS, and MELCOR. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test investigated the long-term interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined time interval, the cavity was flooded with water to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the facility and an overview of results from this test. (authors)

  6. Impact of Zr metal and coking reactions on the ex-vessel source term predictions of CORCON/VANESA

    International Nuclear Information System (INIS)

    Lee, M.; Davis, R.E.; Khatib-Rahbar, M.

    1987-01-01

    During a core meltdown accident in a LWR, molten core materials (corium) could leave the reactor vessel and interact with concrete. In this paper, the impact of the zirconium content of the corium pool and the coking reaction on the release of fission products are quantified using CORCON/Mod2 and VANESA computer codes. Detailed calculations show that the total aerosol generation is proportional to the zirconium content of the corium pool. Among the twelve fission product groups treated by the VANESA code, CsI, Cs 2 O and Nb 2 O 5 are completely released over the course of the core/concrete interaction, while an insignificant quantity of Mo, Ru and ZrO 2 are predicted to be released. The release of BaO, SrO and Ce 2 O increase, while the releases of Te and La 2 O 3 are relatively unaffected by the Zr content of the corium pool. The impact of the coking reaction on the radionuclide release and aerosol production was found to be insignificant

  7. Coolability of volumetrically heated particle beds

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Muhammad

    2017-03-22

    In case of a severe nuclear reactor accident, with loss of coolant, a particle bed may be formed from the fragmentation of the molten core in the residual water at different stages of the accident. To avoid further propagation of the accident and maintain the integrity of the reactor pressure vessel, the decay heat of the particle bed must be removed. To better understand the various thermo-hydraulic processes within such heat-generating particle beds, the existing DEBRIS test facility at IKE has been modified to be able to perform novel boiling, dryout and quenching experiments. The essential experimental data includes the pressure gradients measured by 8 differential pressure transducers along the bed height as a function of liquid and vapour superficial velocities, the determination of local dryout heat fluxes for different system pressures as well as the local temperature distribution measured by a set of 51 thermocouples installed inside the particle bed. The experiments were carried out for two different particle beds: a polydispersed particle bed which consisted of stainless steel balls (2 mm, 3 mm and 6 mm diameters) and an irregular particle bed which consisted of a mixture of steel balls (3 mm and 6 mm) and irregularly shaped Al{sub 2}O{sub 3} particles. Additionally, all experiments were carried out for different flow conditions, such as the reference case of passive 1D top-flooding, 1D bottom flooding (driven by external pumps and different downcomer configurations) and 2D top-/bottom-/lateral flooding with a perforated downcomer. In this work, it has been observed that for both particle beds with downcomer configurations an open downcomer leads to the best coolability (dryout heat flux = 1560 kW/m{sup 2}, polydispersed particle bed, psys = 1 bar) of the particle bed, mainly due to bottom-flow with enhanced natural convection. It has also been shown that a potential lateral flow via a perforation of the downcomer does not bring any further improvements

  8. Coolability of volumetrically heated particle beds

    International Nuclear Information System (INIS)

    Rashid, Muhammad

    2017-01-01

    In case of a severe nuclear reactor accident, with loss of coolant, a particle bed may be formed from the fragmentation of the molten core in the residual water at different stages of the accident. To avoid further propagation of the accident and maintain the integrity of the reactor pressure vessel, the decay heat of the particle bed must be removed. To better understand the various thermo-hydraulic processes within such heat-generating particle beds, the existing DEBRIS test facility at IKE has been modified to be able to perform novel boiling, dryout and quenching experiments. The essential experimental data includes the pressure gradients measured by 8 differential pressure transducers along the bed height as a function of liquid and vapour superficial velocities, the determination of local dryout heat fluxes for different system pressures as well as the local temperature distribution measured by a set of 51 thermocouples installed inside the particle bed. The experiments were carried out for two different particle beds: a polydispersed particle bed which consisted of stainless steel balls (2 mm, 3 mm and 6 mm diameters) and an irregular particle bed which consisted of a mixture of steel balls (3 mm and 6 mm) and irregularly shaped Al 2 O 3 particles. Additionally, all experiments were carried out for different flow conditions, such as the reference case of passive 1D top-flooding, 1D bottom flooding (driven by external pumps and different downcomer configurations) and 2D top-/bottom-/lateral flooding with a perforated downcomer. In this work, it has been observed that for both particle beds with downcomer configurations an open downcomer leads to the best coolability (dryout heat flux = 1560 kW/m 2 , polydispersed particle bed, psys = 1 bar) of the particle bed, mainly due to bottom-flow with enhanced natural convection. It has also been shown that a potential lateral flow via a perforation of the downcomer does not bring any further improvements in

  9. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  10. Analysis of three ex-vessel loss-of-coolant accidents in the first wall cooling system of NET/ITER

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1993-01-01

    An ex-vessel LOCA may be caused by a rupture of a cooling pipe located outside the vacuum vessel. No plasma shutdown and no other counteractions have been assumed in order to study the worst case conditions of the accidents. The next three ex-vessel LOCAs in the primary cooling system of the first wall have been analysed: 1. a large break ex-vessel LOCA caused by a rupture of the cold leg (inner diameter 0.314 m) of the main circuit; 2. an intermediate break ex-vessel LOCA caused by a rupture of a sector inlet feeder (inner diameter 0.158 m); 3. an intermediate break ex-vessel LOCA caused by a rupture of the surge line (inner diameter 0.180 m) of the pressurizer. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the first two scenarios, melting in the first wall starts about 90 s after break initiation. In the third scenario, melting in the first wall start about 323 s after break initiation. Special emphasis has been paid to the characteristics of the break flows, the transient thermal-hydraulic behaviour of the cooling system, and the temperature development in the first wall. (orig.)

  11. In-calandria retention of corium in Indian PHWR - experimental simulations with decay heat

    International Nuclear Information System (INIS)

    Nayak, A.K.

    2015-01-01

    The severe accident at Fukushima has compelled the nuclear community to relook at the safety of existing nuclear power plants (NPP) against natural origin events of beyond design basis and prolonged station black out (SBO). A major lesson learned is to assess the capability of the safety systems to cool the reactor core and spent fuel storage facilities in the event of a prolonged station black out (SBO). Similar safety review is planned for the Indian Pressurized Heavy Water Reactors (PHWRs) considering a prolonged SBO. The Indian PHWR is a heavy water-moderated and cooled, natural uranium-fuelled reactor in which the horizontal fuel channels are submerged in a pool of heavy water moderator located inside the calandria vessel. The calandria vessel is surrounded by a calandria vault having large volume of light water. Concerns are raised that in the event of an unmitigated SBO, it may result into a low probable severe accident leading to core melt down. The core melt may further fail the calandria vessel in case the melt is not quenched. If the calandria vessel fails, the corium shall interact with the cold calandria vault water and concrete resulting in generation of large amount of non-condensable gases and steam which will lead to over pressurization of containment and may cause its failure. Therefore, in-calandria corium retention via external cooling using vault water can be considered as an important accident management program in PHWR. In this strategy, the core melt retains inside the calandria vessel by continually removing the stored heat and decay heat through outer surface of the vessel by cooling water and maintaining the integrity of the vessel. The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat by using the calandria vault water. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics

  12. Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

  13. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Li, Nam Jin; Hong, Joon Wha

    2007-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 16 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  14. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

  15. Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

  16. Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  17. Corium quench in deep pool mixing experiments

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.; Gregorash, D.; Aeschlimann, R.; Sienicki, J.J.

    1985-01-01

    The results of two recent corium-water thermal interaction (CWTI) tests are described in which a stream of molten corium was poured into a deep pool of water in order to determine the mixing behavior, the corium-to-water heat transfer rates, and the characteristic sizes of the quenched debris. The corium composition was 60% UO 2 , 16% ZrO 2 , and 24% stainless steel by weight; its initial temperature was 3080 K, approx.160 K above the oxide phase liquidus temperature. The corium pour stream was a single-phase 2.2 cm dia liquid column which entered the water pool in film boiling at approx.4 m/s. The water subcooling was 6 and 75C in the two tests. Test results showed that with low subcooling, rapid steam generation caused the pool to boil up into a high void fraction regime. In contrast, with large subcooling no net steam generation occurred, and the pool remained relatively quiescent. Breakup of the jet appeared to occur by surface stripping. In neither test was the breakup complete during transit through the 32 cm deep water pool, and molten corium channeled to the base where it formed a melt layer. The characteristic heat transfer rates measured 3.5 MJ/s and 2.7 MJ/s during the fall stage for small and large subcooling, respectively; during the initial stage of bed quench, the surface heat fluxes measured 2.4 MW/m 2 and 3.7 MW/m 2 , respectively. A small mass of particles was formed in each test, measuring typically 0.1 to 1 mm and 1 to 5 mm dia for the large and small subcooling conditions, respectively. 9 refs., 13 figs., 1 tab

  18. Analysis for the coolability of the reactor cavity in a Korean 1000 MWe PWR using MELCOR 1.8.3 computer code

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Ju Yeul; Chung, Chang Hyun; Park, Soo Yong

    1996-01-01

    The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction (MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass. The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment

  19. Corium Oxidation at Temperatures Above 2000 K

    International Nuclear Information System (INIS)

    Hagrman, Donald L.; Rempe, Joy L.

    2001-01-01

    A mechanistic model, based on a quasi-equilibrium analysis of oxidation reactions, is proposed for predicting high-temperature corium oxidation. The analysis suggests that oxide forming on the surface of corium containing uranium, zirconium, and iron is similar to the oxides formed on zirconium and uranium as long as there is a small percentage of unoxidized zirconium or uranium in the metallic phase. This is because of the higher affinity of zirconium and uranium for oxygen. Hence, oxidation rates and heat production rates are similar to (U,Zr) compounds until nearly all the uranium and zirconium in the corium oxidizes. Oxidation rates after this point are predicted to be similar to those implied by the oxide thickness present when the forming oxide ceases to be protective, and heat generation rates should be similar to those implied by iron oxidation, i.e., ∼4% of the zirconium oxidation heating rate.The maximum atomic ratio of unoxidized iron to unoxidized liquid zirconium plus uranium for the formation of a solid protective oxide below 2800 K is estimated for a temperature, T (in Kelvin), as follows:(unoxidized iron)/(unoxidized zirconium + turanium) = (1/28){5.7/exp[-(147 061 + 12.08T log(T) - 61.03T - 0.000555T 2 /1.986T)]} 1/2 .As long as this limit is not exceeded, either zirconium or uranium metal oxidation rates and heating describe the corium oxidation rate. If this limit is exceeded, diffusion of steam to the corium surface will limit the oxidation rate, and linear time-dependent growth of a nonprotective, mostly FeO, layer will occur below the protective (Zr,U) O 2 scale. When this happens, the oxidation should be at the constant rate given by the thickness of the protective layer. Heat generation should be similar to that of iron oxidation

  20. Corium Oxidation at Temperatures Above 2000 K

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, Donald Lee; Rempe, Joy Lynn

    2001-02-01

    A mechanistic model, based on a quasi-equilibrium analysis of oxidation reactions, is proposed for predicting high-temperature corium oxidation. The analysis suggests that oxide forming on the surface of corium containing uranium, zirconium, and iron is similar to the oxides formed on zirconium and uranium as long as there is a small percentage of unoxidized zirconium or uranium in the metallic phase. This is because of the higher affinity of zirconium and uranium for oxygen. Hence, oxidation rates and heat production rates are similar to (U,Zr) compounds until nearly all the uranium and zirconium in the corium oxidizes. Oxidation rates after this point are predicted to be similar to those implied by the oxide thickness present when the forming oxide ceases to be protective, and heat generation rates should be similar to those implied by iron oxidation, i.e., ~4% of the zirconium oxidation heating rate. The maximum atomic ratio of unoxidized iron to unoxidized liquid zirconium plus uranium for the formation of a solid protective oxide below 2800 K is estimated for a temperature, T (in Kelvin), as follows: (unoxidized iron)/(unoxidized zirconium + turanium) = (1/28){5.7/exp[-(147 061 + 12.08T log(T) - 61.03T - 0.000555T2/1.986T)]}1/2. As long as this limit is not exceeded, either zirconium or uranium metal oxidation rates and heating describe the corium oxidation rate. If this limit is exceeded, diffusion of steam to the corium surface will limit the oxidation rate, and linear time-dependent growth of a nonprotective, mostly FeO, layer will occur below the protective (Zr,U) O2 scale. When this happens, the oxidation should be at the constant rate given by the thickness of the protective layer. Heat generation should be similar to that of iron oxidation.

  1. Internal corium catcher of a nuclear reactor

    International Nuclear Information System (INIS)

    Anatolii S Vlasov; Vladimir N Mineev; Aleksandr S Sidorov; Yuri A Zeigarnik

    2005-01-01

    Full text of publication follows: A corium catcher is one of the main devices of a nuclear reactor that provides corium melt and fission products retention within a containment during severe accidents. Several studies and design developments have shown that corium retention within a reactor vessel can be attained with a moderate capacity of the latter (up to 600 - 650 MW el.). With a higher reactor capacity external corium catchers are applied both at Russian (VVER-1000) and European (EPR) reactors. In the external catcher of a VVER-1000 reactor, most technological problems are solved due to using sacrificial material. They are as follows: (a) endo-thermal interaction of corium and sacrificial material reduces a level of the temperatures in the final melt pool; (b) solution in the melt of a great amount of the sacrificial material reduces the specific heat release density and the heat flux density at the boundaries of a melt; (c) due to changing of the oxide-component density an inverse stratification of the metallic and oxide components of the corium takes place, thus excluding heat-flux focusing in the zone of the metallic layer and making it possible to supply water on the free surface of the corium without a danger of incipience of the vapor explosion; (d) final oxidation of zirconium occurs without hydrogen generation. The above principles have been realized in the external catcher of the VVER- 1000 reactor at Tyanvan NPS that is presently under construction in China. Successfully solving of the problems concerning to the external catcher makes it possible to return on the new conceptual and technological basis to the idea of retention of the corium melt inside the vessel of a nuclear reactor of large capacity, that is, to provide the reactor vessel to play a role of an internal catcher. For this purpose, a reactor vessel is elongated by approximately two meters. In the lower part of the vessel, on elliptical bottom, pieces of sacrificial material are arranged

  2. Development of a surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Basso, Simone, E-mail: simoneb@kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2016-12-15

    Highlights: • Severe accident. • Steam explosion. • Surrogate model. • Sensitivity study. • Artificial neural networks. - Abstract: Severe accident mitigation strategy adopted in Nordic type Boiling Water Reactors (BWRs) employs ex-vessel core melt cooling in a deep pool of water below reactor vessel. Energetic fuel–coolant interaction (steam explosion) can occur during molten core release into water. Dynamic loads can threaten containment integrity increasing the risk of fission products release to the environment. Comprehensive uncertainty analysis is necessary in order to assess the risks. Computational costs of the existing fuel–coolant interaction (FCI) codes is often prohibitive for addressing the uncertainties, including the effect of stochastic triggering time. This paper discusses development of a computationally efficient surrogate model (SM) for prediction of statistical characteristics of steam explosion impulses in Nordic BWRs. The TEXAS-V code was used as the Full Model (FM) for the calculation of explosion impulses. The surrogate model was developed using artificial neural networks (ANNs) and the database of FM solutions. Statistical analysis was employed in order to treat chaotic response of steam explosion impulse to variations in the triggering time. Details of the FM and SM implementation and their verification are discussed in the paper.

  3. COOLOCE debris bed experiments and simulations investigating the coolability of cylindrical beds with different materials and flow modes

    Energy Technology Data Exchange (ETDEWEB)

    Takasuo, E.; Kinnunen, T.; Holmstroem, S.; Lehtikuusi, T. [VTT Technical Research Centre of Finland (Finland)

    2013-07-15

    The COOLOCE experiments aim at investigating the coolability of debris beds of different geometries, flow modes and materials. A debris bed may be formed of solidified corium as a result of a severe accident in a nuclear power reactor. The COOLOCE-8 test series consisted of experiments with a top-flooded test bed with irregular gravel as the simulant material. The objective was to produce comparison data useful in estimating the effects of different particle materials and the possible effect of the test arrangement on the results. It was found that the dryout heat flux (DHF) measured for the gravel was lower compared to previous experiments with spherical beads, and somewhat lower compared to the early STYX experiments. The difference between the beads and gravel is at least partially explained by the smaller average size of the gravel particles. The COOLOCE-9 test series included scoping experiments examining the effect of subcooling of the water pool in which the debris bed is immersed. The experiments with initially subcooled pool suggest that the subcooling may increase DHF and increase coolability. The aim of the COOLOCE-10 experiments was to investigate the effect of lateral flooding on the DHF a cylindrical test bed. The top of the test cylinder and its sidewall were open to water infiltration. It was found that the DHF is increased compared to a top-flooded cylinder by more than 50%. This suggests that coolability is notably improved. 2D simulations of the top-flooded test beds have been run with the MEWA code. Prior to the simulations, the effective particle diameter for the spherical beads and the irregular gravel was estimated by single-phase pressure loss measurements performed at KTH in Sweden. Parameter variations were done for particle size and porosity used as input in the models. It was found that with the measured effective particle diameter and porosity, the simulation models predict DHF with a relatively good accuracy in the case of spherical

  4. The MELTSPREAD Code for Modeling of Ex-Vessel Core Debris Spreading Behavior, Code Manual – Version3-beta

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-09-01

    MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input that describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of

  5. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

    International Nuclear Information System (INIS)

    Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

    2009-01-01

    The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

  6. Analysis of the corium phases by X-ray diffraction; Analyses des phases du corium par diffraction des rayons X

    Energy Technology Data Exchange (ETDEWEB)

    Trillon, G

    2004-07-01

    In the framework of the severe accidents R and D studies led by CEA, the better knowledge of the corium behaviour, corium coming from the melting of a nuclear reactor, are fundamental stakes in order to master this kind of accident. Among the available physical properties of the corium, the nature of the final crystalline compounds which have been made during the, cooling gives information about its solidification and its stabilisation. X-Rays Diffraction is the reference method used in order to characterize the corium coming from the different facilities of the European platform PLINIUS of CEA-Cadarache. This work presents the scientific approach that has been followed in order to obtain information both qualitative and quantitative on corium, using X-Rays Diffraction. For instance, a specific method for identifying U{sub 1-x}Zr{sub x}O{sub 2} solid solutions has been developed, and the validity of quantitative analysis of corium crystalline phases using the Rietveld method (with an internal standard), has been tested. This last method has also permitted semi-quantitative measurements of amorphous phases within corium. For these studies, analysis of prototypical corium has been conducted on samples coming from the experiences led on the different facilities of the PLINIUS platform. These analysis allowed for the first time to obtain quantitative data of the corium crystalline phases in order to validate thermodynamic databases and has been used to estimate the thereto-physical properties of the corium. New information on crystalline phases of corium has also been found, especially for the UO{sub 2}-ZrO{sub 2} pseudo binary system. (author)

  7. Success for the Vulcano's team. First real corium flow

    International Nuclear Information System (INIS)

    Carnoy, M.

    1998-01-01

    The aim of the joint CEA-DRN/EDF-DER project 'Vulcano' is the mastery of the corium spreading and cooling on a recovery device. The first real corium spreading test has been successfully performed at the CEA/Cadarache centre (France). This short paper describes the experimental setup and the first results of the experiment. (J.S.)

  8. Post-accident core coolability of light water reactors

    International Nuclear Information System (INIS)

    Michio, I.; Teruo, I.; Tomio, Y.; Tsutao, H.

    1983-01-01

    A study on post-accident core coolability of LWR is discussed based on the practical fuel failure behavior experienced in NSRR, PBF, PNS and others. The fuel failure behavior at LOCA, RIA and PCM conditions are reviewed, and seven types of fuel failure modes are extracted as the basic failure mechanism at accident conditions. These are: cladding melt or brittle failure, molten UO 2 failure, high temperature cladding burst, low temperature cladding burst, failure due to swelling of molten UO 2 , failure due to cracks of embrittled cladding for irradiated fuel rods, and TMI-2 core failure. The post-accident core coolability at each failure mode is discussed. The fuel failures caused actual flow blockage problems. A characteristic which is common among these types is that the fuel rods are in the conditions violating the present safety criteria for accidents, and UO 2 pellets are in melting or near melting hot conditions when the fuel rods failed

  9. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  10. Experimental investigation of the coolability of blocked hexagonal bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hózer, Zoltán, E-mail: zoltan.hozer@energia.mta.hu; Nagy, Imre; Kunstár, Mihály; Szabó, Péter; Vér, Nóra; Farkas, Róbert; Trosztel, István; Vimi, András

    2017-06-15

    Highlights: • Experiments were performed with electrically heated hexagonal fuel bundles. • Coolability of ballooned VVER-440 type bundle was confirmed up to high blockage rate. • Pellet relocation effect causes delay in the cool-down of the bundle. • The bypass line does not prevent the reflood of ballooned fuel rods. - Abstract: The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

  11. Modeling of corium dispersion in DCH accidents

    International Nuclear Information System (INIS)

    Wu, Q.

    1996-01-01

    A model that governs the dispersion process in the direct containment heating (DCH) reactor accident scenario is developed by a stepwise approach. In this model, the whole transient is subdivided into four phases with an isothermal assumption. These are the liquid and gas discharge, the liquid film flow in the cavity before gas blowdown, the liquid and gas flow in the cavity with droplet entrainment, and the liquid transport and re-entrainment in the subcompartment. In each step, the dominant driving mechanisms are identified to construct the governing equations. By combining all the steps together, the corium dispersion information is obtained in detail. The key parameters are predicted quantitatively. These include the fraction of liquid that flows out of the cavity before gas blowdown, the dispersion fraction and the mean droplet diameter in the cavity, the cavity pressure rise due to the liquid friction force, and the dispersion fractions in the containment via different paths. Compared with the data of the 1:10 scale experiments carried out at Purdue University, fairly good agreement is obtained. A stand-alone prediction of the corium dispersion under prototypic Zion reactor conditions is carried out by assuming an isothermal process without chemical reactions. (orig.)

  12. Recent progress in the LACOMERA Project (Large-Scale Experiments on Core Degradation, Melt Retention and Coolability) at the Forschungszentrum Karslruhe

    International Nuclear Information System (INIS)

    Miassoedov, A.; Alsmeyer, H.; Eppinger, B.; Meyer, L.; Steinbrueck, M.

    2004-01-01

    The LACOMERA Project at the Forschungszentrum Karlsruhe (FZK) is a 3 year action within the 5 th Framework Programme of the EU. The overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO-H, and COMET which can be used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. As a result of two calls for proposals, seven organisations from four countries are expected to profit from the LACOMERA Project participating in preparation, conduct and analysis of the following experiments: QUENCH-L1: Air ingression impact on core degradation. The test has provided unique data for the investigation of air ingress phenomenology in conditions as representative as possible of the reactor case regarding the source term. QUENCH-L2: Boil-off of a flooded bundle. The test will be of a generic interest for all reactor types, providing a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. LIVE-L1: Simulation of melt relocation into the Reactor Pressure Vessel (RPV) lower head for VVER conditions. The experiment will provide important information on the melt pool behaviour during the stages of air circulation at the outer RPV surface with a subsequent flooding of the lower head. LIVE-L2: Transient corium spreading and its impact on the heat fluxes to the RPV wall and on the final shape of the melt in the RPV lower head. The test will address the questions of melt stabilisation and the effects of crust formation near the RPV wall for a nonsymmetrical melt pool shape. COMET-L1: Long-term 2D concrete ablation in siliceous concrete cavity at intermediate decay heat power level with

  13. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    International Nuclear Information System (INIS)

    Thakre, S.; Ma, W.

    2013-08-01

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  14. Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

  15. Debris bed coolability using a 3-D two phase model in a porous medium

    Energy Technology Data Exchange (ETDEWEB)

    Bechaud, C.; Duval, F.; Fichot, F. [CEA Cadarache, Inst. de Protection et de Surete Nucleaire13 - Saint-Paul-lez-Durance (France); Quintard, M. [Institut de Mecanique des Fluides de Toulouse, 31 (France); Parent, M. [CEA Grenoble, Dept. de Thermohydraulique et de Physique, 38 (France)

    2001-07-01

    During a severe nuclear accident, a part of the molten corium resulting from the core degradation may relocate in the lower plenum of the reactor vessel. In order to predict the safety margin of the reactor under such conditions, the coolability of this porous heat-generating medium is evaluated in this study and compared with other investigations. In this work, conservation equations derived for debris beds are implemented in the three dimensional thermal-hydraulic module of the CATHARE code. The coolant flow is a two phase flow with phase change. The momentum balance equation for each fluid phase is an extension of Darcy's law. This extension takes into account the capillary effects between the two phases, the relative permeabilities and passabilities of each phase, the interfacial drag force between liquid and gas, and the porous bed configuration (porosity, particle diameter,... ). The model developed is three-dimensional which is important to better predict the flow in configuration such as counter-current flow or to emphasize preferential ways induced by porous geometry. The energy balance equations of the three phases (liquid, gas and solid phase) are obtained by a volume averaging process of the local conservation equations. In this method, the local thermal non-equilibrium between the three phases is considered and the heat exchanges, the phase change rate as well as the thermal dispersion coefficients are calculated as a function of the local geometry of the porous medium. Such a method allows the numerical estimation of these thermal properties which are very difficult to determine experimentally. This feature is a great advantage of this approach. After a brief description of the thermal-hydraulic model, one-dimensional predictions of critical dryout fluxes are presented and compared with results from the literature. Reasonable agreement is obtained. Then a two-dimensional calculation is presented and shows the influence of the porous medium

  16. Concrete benchmark experiment: ex-vessel LWR surveillance dosimetry; Experience ``Benchmark beton`` pour la dosimetrie hors cuve dans les reacteurs a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, H.; D`Hondt, P.; Oeyen, J.; Risch, P.; Bioux, P.

    1993-09-01

    The analysis of DOEL-1 in-vessel and ex-vessel neutron dosimetry, using the DOT 3.5 Sn code coupled with the VITAMIN-C cross-section library, showed the same C/E values for different detectors at the surveillance capsule and the ex-vessel cavity positions. These results seem to be in contradiction with those obtained in several Benchmark experiments (PCA, PSF, VENUS...) when using the same computational tools. Indeed a strong decreasing radial trend of the C/E was observed, partly explained by the overestimation of the iron inelastic scattering. The flat trend seen in DOEL-1 could be explained by compensating errors in the calculation such as the backscattering due to the concrete walls outside the cavity. The `Concrete Benchmark` experiment has been designed to judge the ability of this calculation methods to treat the backscattering. This paper describes the `Concrete Benchmark` experiment, the measured and computed neutron dosimetry results and their comparison. This preliminary analysis seems to indicate an overestimation of the backscattering effect in the calculations. (authors). 5 figs., 1 tab., 7 refs.

  17. Viscosities of corium-concrete mixtures

    International Nuclear Information System (INIS)

    Seiler, J.M.; Ganzhorn, J.

    1997-01-01

    Severe accidents on nuclear reactors involve many situations such as pools of molten core material, melt spreading, melt/concrete interactions, etc. The word 'corium' designates mixtures of materials issued from the molten core at high temperature; these mixtures involve mainly: UO2, ZrO2, Zr and, in small amounts, Ni, Cr, Ag, In, Cd. These materials, when flowing out of the reactor vessel, may interact with the concrete of the reactor building thus introducing decomposition products of concrete into the original mixture. These decomposition products are mainly: SiO 2 , FeO, MgO, CaO and Al 2 O 3 in different amounts depending on the nature of the concrete being considered. Siliceous concrete is rich in SiO 2 , limestone concrete contains both SiO 2 and CaO. Liquidus temperatures of such mixtures are generally obove 2300 K whereas solidus temperatures are ∝1400 K. (orig.)

  18. Technical evaluation of corium cooling at the reactor cavity

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Chan, Eun Sun; Lee, Jae Hun; Lee, Jong In

    1998-01-01

    To terminate the progression of the severe accident and mitigate the accident consequences, corium cooling has been suggested as one of most important design features considered in the severe accident mitigation. Till now, some kinds of cooling methodologies have been identified and, specially, the corium cooling at the reactor cavity has been considered as one of the most promising cooling methodologies. Moreover, several design requirements related to the corium cooling at the reactor cavity have been also suggested and applied to the design of the next generation reactor. In this study, technical descriptions are briefly described for the important issues related to the corium cooling at the reactor cavity, i.e. cavity area, cavity flooding system, etc., and simple evaluations for those items have been performed considering present technical levels including the experiment and analytical works

  19. Oxidation effects during corium melt in-vessel retention

    Energy Technology Data Exchange (ETDEWEB)

    Almyashev, V.I.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Sulatsky, A.A.; Vitol, S.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V. [Ioffe Institute, St. Petersburg (Russian Federation); Bechta, S. [Royal Institute of Technology (KHT), Stockholm (Sweden); Barrachin, M.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [Joint Research Centre, Institut für Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [CEA Cadarache-DEN/DTN/STRI (France)

    2016-08-15

    Highlights: • Corium–steel interaction tests were re-examined particularly for transient processes. • Oxidation of corium melt was sensitive to oxidant supply and surface characteristics. • Consequences for vessel steel corrosion rates in severe accidents were discussed. - Abstract: In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.

  20. Melt cooling by bottom flooding: The experiment CometPC-H3. Ex-vessel core melt stabilization research

    International Nuclear Information System (INIS)

    Alsmeyer, H.; Cron, T.; Merkel, G.; Schmidt-Stiefel, S.; Tromm, W.; Wenz, T.

    2003-03-01

    and bottom crusts had formed, and included the bulk of the melt, which at this stage was still in the liquid state. Further cool down of the melt continued slowly over the next 40 min. (orig.) concrete underneath are completely intact as always cooled by the presence of the flooding water. In summary, the experiment has demonstrated that the CometPC cooling concept is able to stop and to cool the melt, although the expected porosity formation of the majority of the melt did not prevail in the present experiment. Under this aspect, the porous, water filled concrete layer has proven its reliability to stop the melt. On the other side, volcanic melt eruptions which occurred during a limited period of the test, did not improve the coolability significantly, as no major new porosity was created. (orig.)

  1. Steam explosion triggering phenomena: stainless steel and corium-E simulants studied with a floodable arc melting apparatus

    International Nuclear Information System (INIS)

    Nelson, L.S.; Buxton, L.D.

    1978-05-01

    Laboratory-scale experiments on the thermal interaction of light water reactor core materials with water have been performed. Samples (10--35 g) of Type 304 stainless steel and Corium-E simulants were each flooded with approximately 1.5 litres of water to determine whether steam explosions would occur naturally. Many of the experiments also employed artificially induced pressure transients in an attempt to initiate steam explosions. Vigorous interactions were not observed when the triggering pulse was not applied, and for stainless steel the triggering pulse initiated only coarse fragmentation. Two-stage, pressure-producing interactions were triggered for an ''oxidic'' Corium-E simulant. An impulse-initiated gas release theory has been simulated to explain the initial sample fragmentation. Although the delayed second stage of the event is not fully understood, it does not appear to be readily explained with classical vapor explosion theory. Rather, some form of metastability of the melt seems to be involved

  2. In-vessel coolability and steam explosion in Nordic BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ma, W.; Li, L.; Hansson, R.; Villanueva, W.; Kudinov, P.; Manickam, L.; Tran, C.-T. (Royal Institute of Technology (KTH) (Sweden))

    2011-05-15

    The objective of this research is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in the Nordic BWR plants which employ cavity flooding as severe accident management (SAM) strategy. To quantify the coolability of debris bed packed with irregular particles, the friction laws of fluid flow in particulate beds packed with non-spherical particles were investigated on the POMECO-FL test facility, and the experimental data suggest that the Ergun equation is applicable if the effective particle diameter of the particles is represented by the equivalent diameter of the particles, which is the product of Sauter mean diameter and shape factor of the particles. One-way coupling analysis between PECM model for melt pool heat transfer and ANSYS thermo-structural mechanics was performed to analyze the vessel creep, and the results revealed two different modes of vessel failure: a 'ballooning' of the vessel bottom and a 'localized creep' concentrated within the vicinity of the top surface of the melt pool. Single-droplet steam explosion experiments were carried out by using oxidic mixture of WO{sub 3}-CaO, and the results show an apparent difference in steam explosion energetics between the eutectic and non-eutectic melts at low melt superheat (100 deg. C). (Author)

  3. Crust formation and its effect on the molten pool coolability

    Energy Technology Data Exchange (ETDEWEB)

    Park, R.J.; Lee, S.J.; Sim, S.K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-09-01

    Experimental and analytical studies of the crust formation and its effect on the molten pool coolability have been performed to examine the crust formation process as a function of boundary temperatures as well as to investigate heat transfer characteristics between molten pool and overlying water in order to evaluate coolability of the molten pool. The experimental test results have shown that the surface temperature of the bottom plate is a dominant parameter in the crust formation process of the molten pool. It is also found that the crust thickness of the case with direct coolant injection into the molten pool is greater than that of the case with a heat exchanger. Increasing mass flow rate of direct coolant injection to the molten pool does not affect the temperature of molten pool after the crust has been formed in the molten pool because the crust behaves as a thermal barrier. The Nusselt number between the molten pool and the coolant of the case with no crust formation is greater than that of the case with crust formation. The results of FLOW-3D analyses have shown that the temperature distribution contributes to the crust formation process due to Rayleigh-Benard natural convection flow.

  4. In-vessel coolability and steam explosion in Nordic BWRs

    International Nuclear Information System (INIS)

    Ma, W.; Li, L.; Hansson, R.; Villanueva, W.; Kudinov, P.; Manickam, L.; Tran, C.-T.

    2011-05-01

    The objective of this research is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in the Nordic BWR plants which employ cavity flooding as severe accident management (SAM) strategy. To quantify the coolability of debris bed packed with irregular particles, the friction laws of fluid flow in particulate beds packed with non-spherical particles were investigated on the POMECO-FL test facility, and the experimental data suggest that the Ergun equation is applicable if the effective particle diameter of the particles is represented by the equivalent diameter of the particles, which is the product of Sauter mean diameter and shape factor of the particles. One-way coupling analysis between PECM model for melt pool heat transfer and ANSYS thermo-structural mechanics was performed to analyze the vessel creep, and the results revealed two different modes of vessel failure: a 'ballooning' of the vessel bottom and a 'localized creep' concentrated within the vicinity of the top surface of the melt pool. Single-droplet steam explosion experiments were carried out by using oxidic mixture of WO 3 -CaO, and the results show an apparent difference in steam explosion energetics between the eutectic and non-eutectic melts at low melt superheat (100 deg. C). (Author)

  5. A study on corium behaviour under external vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Kim, Sang Baik; Kang, Kyung Ho; Koo, Kil Mo; Kim, Hee Dong

    2000-04-01

    This study presents the results of evaluation and analysis on the second phase of the RASPLAV project for three years between July 1, 1997 and June 30, 2000. In the RASPLAV Phase II study, two large-scale experiments of RASPLAV-AW-200-3, 4 were conducted to estimate the heat flux distribution in the corium and thermal interaction between the corium and the reactor vessel. Several small-scale experiments such as TULPAN, TF, and STF were conducted to analyze thermal stratification and additive effect of core materials on corium behavior. The Salt experiments were conducted to estimate the crust and the mushy region formation, as well as natural convection heat transfer in the corium. Material properties of the corium and the salt were measured in the RASPLAV project. During the RASPLAV-AW-200-3 test, approximately 22 kg of the corium leaked from the test furnace, because Fe from the FeO, which was additive to reduce the melting temperature of fuel pellet, interacted with Tungsten protector. It is concluded from the AW-200-3 test results that the oxidized U-Zr-O is not separated. From the RASPLAV-AW-200-4 test results, the C-32 fuel with the miscibility gap and low content of carbon was not separated thermally. The carbon is known as a dominant factor in the thermal stratification of the corium from the small and medium scale test results such as TULPAN, TF, and STF. The fuel composition, test method and condition in the RASPLAV-AW-2003,4 were selected using the small and medium scale test results. It is confirmed from the Salt test that the analytical model of the CONV code predicts heat transfer with crust formation in the molten pool very well.

  6. A study on corium behaviour under external vessel cooling

    International Nuclear Information System (INIS)

    Park, Rae Joon; Kim, Sang Baik; Kang, Kyung Ho; Koo, Kil Mo; Kim, Hee Dong

    2000-04-01

    This study presents the results of evaluation and analysis on the second phase of the RASPLAV project for three years between July 1, 1997 and June 30, 2000. In the RASPLAV Phase II study, two large-scale experiments of RASPLAV-AW-200-3, 4 were conducted to estimate the heat flux distribution in the corium and thermal interaction between the corium and the reactor vessel. Several small-scale experiments such as TULPAN, TF, and STF were conducted to analyze thermal stratification and additive effect of core materials on corium behavior. The Salt experiments were conducted to estimate the crust and the mushy region formation, as well as natural convection heat transfer in the corium. Material properties of the corium and the salt were measured in the RASPLAV project. During the RASPLAV-AW-200-3 test, approximately 22 kg of the corium leaked from the test furnace, because Fe from the FeO, which was additive to reduce the melting temperature of fuel pellet, interacted with Tungsten protector. It is concluded from the AW-200-3 test results that the oxidized U-Zr-O is not separated. From the RASPLAV-AW-200-4 test results, the C-32 fuel with the miscibility gap and low content of carbon was not separated thermally. The carbon is known as a dominant factor in the thermal stratification of the corium from the small and medium scale test results such as TULPAN, TF, and STF. The fuel composition, test method and condition in the RASPLAV-AW-2003,4 were selected using the small and medium scale test results. It is confirmed from the Salt test that the analytical model of the CONV code predicts heat transfer with crust formation in the molten pool very well

  7. The VULCANO VE-U7 Corium spreading benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Christophe; Haquet, Jean-Francois [CEA Cadarache, Severe Accident Mastering experimental Laboratory (DEN/DTN/STRI/LMA), 13108 St Paul lez Durance (France); Spindler, Bertrand [CEA Grenoble, Physicochemistry and Multiphasic Thermalhydraulics Laboratory (DEN/DTN/SE2T/LPTM), 17 rue des Martyrs, F-38054 Grenoble CEDEX 9 (France); Spengler, Claus [Gesellschaft fuer Reaktorsicherheit mbH, Department for Thermohydraulics/Process Engineering, Schwertnergasse 1, D-50667 Koeln (Germany); Foit, Jerzy [Forschungszentrum Karlsruhe GmbH, Institut fuer Kern nd Energietechnik (IKET), P.O. Box 3640, D-76021 Karlsruhe (Germany)

    2006-07-01

    In a hypothetical nuclear reactor severe accident, corium spreading is one possible mitigation measure that has been selected for the EPR design. A post-test benchmark exercise has been organized on the VULCANO VE-U7 corium spreading experiment. In this test, a prototypic corium mixture representative of what could be expected at the opening of EPR reactor-pit gate has been spread on siliceous concrete and on a reference channel in inert refractory ceramic. The spreading progression was not much affected by the presence of concrete and sparging gases. The procedure used to estimate the corium physical properties from its composition and temperature provided a satisfactory data set. The CORFLOW, LAVA and THEMA codes provide satisfactory calculations of the spreading front evolution and of its final length. LAVA and THEMA estimations of the substrate temperatures, which are the initial conditions for longer term Molten Core Concrete Interaction or Corium Ceramic Interaction computations, are also close to the measured data, within the experimental uncertainties. (authors)

  8. The VULCANO VE-U7 Corium spreading benchmark

    International Nuclear Information System (INIS)

    Journeau, Christophe; Haquet, Jean-Francois; Spindler, Bertrand; Spengler, Claus; Foit, Jerzy

    2006-01-01

    In a hypothetical nuclear reactor severe accident, corium spreading is one possible mitigation measure that has been selected for the EPR design. A post-test benchmark exercise has been organized on the VULCANO VE-U7 corium spreading experiment. In this test, a prototypic corium mixture representative of what could be expected at the opening of EPR reactor-pit gate has been spread on siliceous concrete and on a reference channel in inert refractory ceramic. The spreading progression was not much affected by the presence of concrete and sparging gases. The procedure used to estimate the corium physical properties from its composition and temperature provided a satisfactory data set. The CORFLOW, LAVA and THEMA codes provide satisfactory calculations of the spreading front evolution and of its final length. LAVA and THEMA estimations of the substrate temperatures, which are the initial conditions for longer term Molten Core Concrete Interaction or Corium Ceramic Interaction computations, are also close to the measured data, within the experimental uncertainties. (authors)

  9. Study on coolability of melt pool with different strategies

    International Nuclear Information System (INIS)

    Kulkarni, P.P.; Nayak, A.K.

    2014-01-01

    Highlights: • Experiments have been performed to test quenching of molten pool with different schemes. • Top flooding, bottom flooding and indirect cooling schemes were used. • A single simulant material with same mass and initial temperature was used. • Bottom flooding technique is found to be the most effective technique. • A comparison of all the three techniques has been presented. - Abstract: After the Fukushima accident, there have been a lot of concerns regarding long term core melt stabilization following a severe accident in nuclear reactors. Several strategies have been contemplated for quenching and stabilization of core melt like top flooding, bottom flooding, indirect cooling, etc. However, the effectiveness of these schemes is yet to be determined properly, for which, lot of experiments are needed. Several experiments have been performed for coolability of molten pool under top flooding condition. A few experiments have been performed for study of coolability of melt pool under bottom flooding as well as for indirect cooling. Besides, these tests are very scattered because they involve different simulant materials, initial temperatures and masses of melt, which makes it very difficult to judge the effectiveness of a particular technique and advantage over the other. In the present paper we have carried out different experiments wherein a single simulant material with same mass was cooled with different techniques starting from the same initial temperature. The result showed that, while top flooding and indirect cooling took several hours to cool, bottom flooding took a few minutes to cool the melt which makes it the most effective technique

  10. Effects of thermohydraulics on clad ballooning, flow blockage and coolability in a LOCA

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Neitzel, H.J.; Wiehr, K.

    1983-01-01

    Thermohydraulic boundary conditions have a dominating effect on clad ballooning, flow blockage and coolability: Increasing heat transfer to the fluid decreases the total circumferential strain; Countercurrent flow in a combined injection leads to a relatively small flow blockage; Burst claddings exhibit premature quenching. Differences in the test results obtained in several countries are mainly due to different thermohydraulic test conditions; all test data are consistent with the understanding elaborated within the REBEKA program. Core coolability in a LOCA can be maintained. (author)

  11. Corrosion of vessel steel during its interaction with molten corium

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

    2006-01-01

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments

  12. Corrosion of vessel steel during its interaction with molten corium

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

    2006-01-01

    This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheres during an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities and oxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium-specimen ingot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction

  13. Corrosion of vessel steel during its interaction with molten corium

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation)]. E-mail: bechta@sbor.spb.su; Khabensky, V.B. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Vitol, S.A. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Krushinov, E.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Granovsky, V.S. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Lopukh, D.B. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Gusarov, V.V. [Institute of Silicate Chemistry of Russian Academy of Science (ISC of RAS), Odoevsky str., b. 24/2, 199155 St. Petersburg (Russian Federation); Martinov, A.P. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Martinov, V.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Fieg, G. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Tromm, W. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Bottomley, D. [Europaeische Kommission, General Direktion GFS, Institut fuer Transurane (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Tuomisto, H. [Fortum Engineering Ltd. 00048 FORTUM, Rajatorpantie 8, Vantaa (Finland)

    2006-07-15

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments.

  14. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

    Directory of Open Access Journals (Sweden)

    Saas Laurent

    2017-01-01

    Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…. During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.

  15. An experimental study on coolability of a particulate bed with radial stratification or triangular shape

    International Nuclear Information System (INIS)

    Thakre, Sachin; Li, Liangxing; Ma, Weimin

    2014-01-01

    Highlights: • Dryout heat flux of a particulate bed with radial stratification is obtained. • It was found to be dominated by hydrodynamics in the bigger size of particle layer. • Coolability of a particulate bed with triangular shape is investigated. • The coolability is improved in the triangular bed due to lateral ingression of coolant. • Coolability of both beds is enhanced by a downcomer. - Abstract: This paper deals with the results of an experimental study on the coolability of particulate beds with radial stratification and triangular shape, respectively. The study is intended to get an idea on how the coolability is affected by the different features of a debris bed formed in a severe accident of light water reactors. The experiments were performed on the POMECO-HT facility which was constructed to investigate two-phase flow and heat transfer in particulate beds under either top-flooding or bottom-fed condition. A downcomer is designed to enable investigation of the effectiveness of natural circulation driven coolability. Two homogenous beds were also employed in the present study to compare their dryout power densities with those of the radially stratified bed and the triangular bed. The results show that the dryout heat fluxes of the homogeneous beds at top-flooding condition can be predicted by the Reed model. For the radially stratified bed, the dryout heat flux is dominated by two-phase flow in the columns packed with larger particles, and the dryout occurred initially near the boundary between the middle column and a side column. Given the same volume of particles under top-flooding condition, the dryout power density of the triangular bed is about 69% higher than that of the homogenous bed. The coolability of all the beds is enhanced by bottom-fed coolant driven by either forced injection or downcomer-induced natural circulation

  16. An experimental study on coolability of a particulate bed with radial stratification or triangular shape

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, Sachin; Li, Liangxing; Ma, Weimin, E-mail: ma@safety.sci.kth.se

    2014-09-15

    Highlights: • Dryout heat flux of a particulate bed with radial stratification is obtained. • It was found to be dominated by hydrodynamics in the bigger size of particle layer. • Coolability of a particulate bed with triangular shape is investigated. • The coolability is improved in the triangular bed due to lateral ingression of coolant. • Coolability of both beds is enhanced by a downcomer. - Abstract: This paper deals with the results of an experimental study on the coolability of particulate beds with radial stratification and triangular shape, respectively. The study is intended to get an idea on how the coolability is affected by the different features of a debris bed formed in a severe accident of light water reactors. The experiments were performed on the POMECO-HT facility which was constructed to investigate two-phase flow and heat transfer in particulate beds under either top-flooding or bottom-fed condition. A downcomer is designed to enable investigation of the effectiveness of natural circulation driven coolability. Two homogenous beds were also employed in the present study to compare their dryout power densities with those of the radially stratified bed and the triangular bed. The results show that the dryout heat fluxes of the homogeneous beds at top-flooding condition can be predicted by the Reed model. For the radially stratified bed, the dryout heat flux is dominated by two-phase flow in the columns packed with larger particles, and the dryout occurred initially near the boundary between the middle column and a side column. Given the same volume of particles under top-flooding condition, the dryout power density of the triangular bed is about 69% higher than that of the homogenous bed. The coolability of all the beds is enhanced by bottom-fed coolant driven by either forced injection or downcomer-induced natural circulation.

  17. Success for the Vulcano`s team. First real corium flow; Succes pour l`equipe de Vulcano. Premiere coulee de corium reel

    Energy Technology Data Exchange (ETDEWEB)

    Carnoy, M. [CEA Cadarache, 13 - Saint-Paul-Lez-Durance (France). Dept. d`Etude des Reacteurs

    1998-03-01

    The aim of the joint CEA-DRN/EDF-DER project `Vulcano` is the mastery of the corium spreading and cooling on a recovery device. The first real corium spreading test has been successfully performed at the CEA/Cadarache centre (France). This short paper describes the experimental setup and the first results of the experiment. (J.S.)

  18. Overview of the projects sponsored within th EU-R and D framework programme

    International Nuclear Information System (INIS)

    Zurita, A.; Goethem, G. van; Bermejo, J.M.

    1999-01-01

    Assuming that preventive measures to avoid reactor pressure vessel (RPV) failure under core degradation scenarios have failed, the stabilisation of the core melts after its release from the RPV is a key issue. Adequate cooling of the ex-vessel corium and the control of its interactions with the coolant and structures are the main challenges to mitigate and stabilise the situation preserving the containment integrity. In this regard, the on-going Fourth Euratom Framework Programme (4 th EFP) contributes with experimental and theoretical research activities aimed at responding to the main challenges mentioned, by satisfying three objectives: - To improve the understanding of the basic physics related to ex-vessel corium behaviour from the phenomenological and technological viewpoints, as well as to provide a methodology for investigating it and setting up joint multi-partner projects to be co-sponsored and co-ordinated by the EC; - to quantify and reduce the uncertainties associated with the risk issues by conducting experimental and numerical investigations and eventually to achieve a European consensus on the phenomenology and on accident mitigation strategies; - to provide a technological response to the risk issues by developing engineered safety systems (e.g. core-catchers) and severe accident management strategies (e.g. guidelines), and to discuss such a technological response with the end users of these technologies, i.e. designers and licensers. To fulfil these objectives, the 4 th EFP co-sponsors a total of five projects within the cluster 'Ex-Vessel Corium Behaviour and Coolability' of the Nuclear Fission Safety Programme. The research undertaken addresses the main different aspects such as the determination of the composition and thermodynamic data of the melt; experiments on spreading behaviour on various types of surfaces, as well as on corium coolability by flooding or water injection; investigation of corium stratification, crust and heat transfer

  19. Study on ex-vessel cooling of RPV (behavior of coalesced bubbles and trigger condition of critical heat flux on inclined plate)

    International Nuclear Information System (INIS)

    Ohtake, H.; Koizumi, Y.; Takano, K.I.

    2001-01-01

    The Ex-vessel cooling of Reactor-Pressure-Vessel in Light-Water-Reactor at the severe accident have been proposed for future nuclear reactors. The estimation of Critical-Heat-Flux on a downward-facing curvilinear surface, like a hemisphere, is important to the assessment of the cooling. In this study, the CHFs on inclined surfaces were examined experimentally focusing on orientation of the heating surface. In order to discuss detailed mechanism of the CHF, the behaviors of coalesced bubbles near the heating surface were investigated through visual observations. The critical heat flux obtained in the present experiments increased with the inclined angle over the present experimental range. The dependence of the inclined angle on the critical heat flux was q CHF,R-113 [q] = f (q 0.33 ) for the present experimental results. The effect of the surface orientation on the critical heat flux was roughly explained by using the simple analytical model based on the macro-layer model and Kelvin-Helmholtz instability. From visual observations for behavior of bubbles near the heating surface, whereas the coalesced bubble covered over the heating surface for the inclined angle of 0 degree, the coalesced bubble moved upward to avoid packing the bubble on the surface above 5 degree. As the inclined angle increased, the velocity of the coalesced bubble was high, the period covered the heater and the bubble length were small. The results suggested that the CHF was closely related to forming the coalesced bubble and the behavior of the bubble. (author)

  20. On the air coolability of TRIGA reactors following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    El-Genk, Mohamed S.; Kim, Sung-Ho; Zaki, Galal M.; Foushee, Fabian; Philbin, Jeffrey S.; Schulze, James

    1986-01-01

    This paper describes the experiments on the air-coolability of a heated rod in a vertical open annulus at near atmospheric pressure. This data can be applied to the coolability of reactor fuel rods that are totally uncovered in a Loss-of-Coolant Accident (LOCA). As a prelude to measuring air coolability of specific core geometries (bundles), heat transfer data was collected for natural convection of atmospheric air in open vertical annuli with an isoflux inner wall and an insulated outer wall (diameter ratios, annulus ratio, of 1.155, 1.33, 1.63, and 12). Although the inner heated tube had the same overall dimensions as the fuel rod in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (3.81 cm o.d. and 55.5 cm long), the heated length was only 36.0 cm rather than the entire 50.5 cm for the ACRR's rods. The test assembly was operated at heat fluxes up to 1.38 W/cm 2 with a corresponding surface temperature of 852 K. The annulus data was extrapolated to an equilibrium surface temperature of 1200 K (as a coolability limit of TRIGA reactors) to provide a qualitative estimate of the coolability of multirod bundles by free convection of atmospheric air. The results suggest that for a typical pitch-to-diameter ratio of 1.12 in the ACRR the decay heat removal level is about 1.0 kW/m. This corresponds to an initial decay power following sustained operations at about 12.5 kW/m in the ACRR. However, because of the uncertainties in duplicating the actual thermal-hydraulic conditions in a multirod bundle using a single rod annulus, the actual coolability of open pool reactors could be different from those suggested in this paper. (author)

  1. Assessment of in-vessel corium retention in CPR1000

    International Nuclear Information System (INIS)

    Chen Xing; Zhang Shishun; Lin Jiming

    2011-01-01

    The In-Vessel corium Retention (IVR) strategy of Chinese 1000 MW class commercial pressurized water reactor (CPR1000) is assessed by Risk-Oriented Accident Analysis Methodology (ROAAM). Four representative severe accident scenarios are selected for the IVR assessment in this paper. According to four representative severe accident scenarios consequence calculated by the deterministic code combined with engineering judgment, the input probability distribution of the assessment is determined. Success probability of IVR from the viewpoint of thermal failure is then predicted using MOPOL code. MOPOL is a code developed basing on the well known ROAAM frame and heat transfer model of corium. It is demonstrated that the success probability of IVR by Reactor Cavity Flooding in CPR1000 is potentially higher than 99%. Application of IVR strategy in CPR1000 is envisioned probable if a further more comprehensive risk-benefit evaluation conclusion is positive. (authors)

  2. The effect of self-leveling on debris bed coolability under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Basso, S.; Konovalenko, A. [Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, D5, Stockholm 106 91 (Sweden); Yakush, S.E. [Institute for Problems in Mechanics of the Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow 119526 (Russian Federation); Kudinov, P. [Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, D5, Stockholm 106 91 (Sweden)

    2016-08-15

    Highlights: • A model for coolability of a self-leveling, variable-shape debris bed is proposed. • Sensitivity analysis is performed to screen out the less influential input parameters. • A small fraction of scenarios has initially a non-coolable debris bed configuration. • The fraction of non-coolable scenarios decreases with time due to self-leveling. - Abstract: Nordic-type boiling water reactors employ melt fragmentation, quenching, and long term cooling of the debris bed in a deep pool of water under the reactor vessel as a severe accident (SA) mitigation strategy. The height and shape of the bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry depends on its formation process (melt release, fragmentation, sedimentation and settlement on the containment basemat), but it also changes with time afterwards, due to particle redistribution promoted by coolant flow (self-leveling). The ultimate goal of this work is to develop an approach to the assessment of the probability that debris in such a variable-shape bed can reach re-melting (which means failure of SA mitigation strategy), i.e. the time necessary for the slumping debris bed to reach a coolable configuration is larger than the time necessary for the debris to reach the re-melting temperature. For this purpose, previously developed models for particulate debris spreading by self-leveling and debris bed dryout are combined to assess the time necessary to reach a coolable state and evaluate its uncertainty. Sensitivity analysis was performed to screen out less important input parameters, after which Monte Carlo simulation was carried out in order to collect statistical characteristics of the coolability time. The obtained results suggest that, given the parameters ranges typical of Nordic BWRs, only a small fraction of debris beds configurations exhibits the occurrence of dryout. Of the

  3. Results of recent KROTOS FCI tests. Alumina vs. corium melts

    Energy Technology Data Exchange (ETDEWEB)

    Huhtiniemi, I.; Magallon, D.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

    1998-01-01

    Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150 K - 750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1 - 0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed. (author)

  4. Experimental simulation of corium dispersion phenomena in direct containment heating

    International Nuclear Information System (INIS)

    Wu, Q.

    1996-01-01

    In a direct containment heating (DCH) accident scenario, the degree of corium dispersion is one of the most significant factors responsible for the reactor containment heating and pressurization. To study the mechanisms of the corium dispersion phenomenon, a DCH separate effect test facility of 1:10 linear scale for Zion PWR geometry is constructed. Experiments are carried out with air-water and air-woods metal simulating steam and molten core materials. The physical process of corium dispersion is studied in detail through various instruments, as well as with flow visualization at several locations. The accident transient begins with the liquid jet discharge at the bottom of the reactor pressure vessel. Once the jet impinges on the cavity bottom floor, it immediately spreads out and moves rapidly to the cavity exit as a film flow. Part of the discharged liquid flows out of the cavity before gas blowdown, and the rest is subjected to the entrainment process due to the high speed gas stream. The liquid film and droplet flows from the reactor cavity will then experience subcompartment trapping and re-entrainment. Consequently, the dispersed liquid droplets that follow the gas stream are transported into the containment atmosphere, resulting in containment heating and pressurization in the prototypic condition. Comprehensive measurements are obtained in this study, including the liquid jet velocity, liquid film thickness and velocity transients in the test cavity, gas velocity and velocity profile in the cavity, droplet size distribution and entrainment rate, and the fraction of dispersed liquid in the containment building. These data are of great importance for better understanding of the corium dispersion mechanisms. (orig.)

  5. An experimental study of steam explosions involving CORIUM melts

    International Nuclear Information System (INIS)

    Millington, R.A.

    1984-05-01

    An experimental programme to investigate molten fuel coolant interactions involving 0.5 kg thermite-generated CORIUM melts and water has been carried out. System pressures and initial coolant subcoolings were chosen to enhance the probability of steam explosions. Yields and efficiencies of the interactions were found to be very close to those obtained from similar experiments using molten UO 2 generated from a Uranium/Molybdenum Trioxide thermite. (author)

  6. Water boiling on the corium melt surface under VVER severe accident conditions

    International Nuclear Information System (INIS)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Sulatsky, A.A.; Khabensky, V.B.; Lopukh, D.B.; Petrov, Y.B.; Pechenkov, A.Y.

    2000-01-01

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the 'Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO 2+x -16% ZrO 2 -15% Fe 2 O 3 -6% Cr 2 O 3 -3% Ni 2 O 3 . The melt surface temperature ranged within 1920-1970 K. (orig.)

  7. Comparison of different surface quantitative analysis methods. Application to corium

    International Nuclear Information System (INIS)

    Guilbaud, N.; Blin, D.; Perodeaud, Ph.; Dugne, O.; Gueneau, Ch.

    2000-01-01

    In case of a severe hypothetical accident in a pressurized water reactor, the reactor assembly melts partially or completely. The material formed, called corium, flows out and spreads at the bottom of the reactor. To limit and control the consequences of such an accident, the specifications of the O-U-Zr basic system must be known accurately. To achieve this goal, the corium mix was melted by electron bombardment at very high temperature (3000 K) followed by quenching of the ingot in the Isabel 1 evaporator. Metallographic analyses were then required to validate the thermodynamic databases set by the Thermo-Calc software. The study consists in defining an overall surface quantitative analysis method that is fast and reliable, in order to determine the overall corium composition. The analyzed ingot originated in a [U+Fe+Y+UO 2 +ZrO 2 ) mix, with a total mass of 2253.7 grams. Several successive heating with average power were performed before a very brief plateau at very high temperature, so that the ingot was formed progressively and without any evaporation liable to modify its initial composition. The central zone of the ingot was then analyzed by qualitative and quantitative global surface methods, to yield the volume composition of the analyzed zone. Corium sample analysis happens to be very complex because of the variety and number of elements present, and also because of the presence of oxygen in a heavy element like the uranium based matrix. Three different global quantitative surface analysis methods were used: global EDS analysis (Energy Dispersive Spectrometry), with SEM, global WDS analysis (Wavelength Dispersive Spectrometry) with EPMA, and coupling of image analysis with EDS or WDS point spectroscopic analyses. The difficulties encountered during the study arose from sample preparation (corium is very sensitive to oxidation), and the choice of acquisition parameters of the images and analyses. The corium sample studied consisted of two zones displaying

  8. Cold crucible technique for interaction test of molten corium with structure

    International Nuclear Information System (INIS)

    Ha, Kwang Soon; An, Sang Mo; Min, Beong Tae; Kim, Hwan Yeol

    2012-01-01

    During a severe accident, the molten corium might interact with several structures in a nuclear power plant such as core peripheral structures, lower plenum, lower head vessel, and external structures of a reactor vessel. The interaction of the molten corium with the structure depends on the molten corium composition, temperature, structural materials, and environmental conditions such as pressure and humidity. For example, the interaction of a metallic molten corium containing metal uranium (U) and zirconium (Zr) with the oxidized steel structure (Fe 2O3 ) is affected by not only thermal ablation but oxidation reduction reaction because the oxidation quotients of the U and Zr are higher than that of Fe. KAERI set up an experimental facility and technique using a cold crucible melting method to verify the interaction mechanism between the metallic molten corium and structural materials. This technique includes the generation of the metallic melt, melt delivery, measurement of the interaction process, and post analyses after the test

  9. ASTEC application to in-vessel corium retention

    International Nuclear Information System (INIS)

    Tarabelli, D.; Ratel, G.; Pelisson, R.; Guillard, G.; Barnak, M.; Matejovic, P.

    2009-01-01

    This paper summarizes the work done in the SARNET European Network of Excellence on Severe Accidents (6th Framework Programme of the European Commission) on the capability of the ASTEC code to simulate in-vessel corium retention (IVR). This code, jointly developed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) and the German Gesellschaft fuer Anlagen und Reaktorsicherheit mbH (GRS) for simulation of severe accidents, is now considered as the European reference simulation tool. First, the DIVA module of ASTEC code is briefly introduced. This module treats the core degradation and corium thermal behaviour, when relocated in the reactor lower head. Former ASTEC V1.2 version assumed a predefined stratified molten pool configuration with a metallic layer on the top of the volumetrically heated oxide pool. In order to reflect the results of the MASCA project, improved models that enable modelling of more general corium pool configurations were implemented by the CEA (France) into the DIVA module of the ASTEC V1.3 code. In parallel, the CEA was working on ASTEC modelling of the external reactor vessel cooling (ERVC). The capability of the ASTEC CESAR circuit thermal-hydraulics to simulate the ERVC was tested. The conclusions were that the CESAR module is capable of simulating this system although some numerical and physical instabilities can occur. Developments were then made on the coupling between both DIVA and CESAR modules in close collaboration with IRSN. In specific conditions, code oscillations remain and an analysis was made to reduce the numerical part of these oscillations. A comparison of CESAR results of the SULTAN experiments (CEA) showed an agreement on the pressure differences. The ASTEC V1.2 code version was applied to IVR simulation for VVER-440/V213 reactors assuming defined corium mass, composition and decay heat. The external cooling of reactor wall was simulated by applying imposed coolant temperature and heat transfer

  10. Coolability of oxidized particulate debris bed accumulated in horizontal narrow gaps

    International Nuclear Information System (INIS)

    Arai, Y.; Sugiyama, K.; Narabayashi, T.

    2007-01-01

    When LOCA occurs in a nuclear reactor system, the coolability of the core would be kept as reported at a series of presentations in ICONE14. Therefore the probability of the core meltdown is negligible small. However, from the view point of defense in depth, it is necessary to be sure that the coolability of the bottom of reactor pressure vessel (RPV) is maintained even if a part of the core should melt and a substantial amount of debris should be deposited on the lower plenum. We carried out an experimental study in order to observe the coolability of particulate core-metal debris bed with 12 mm thickness accompanied with rapid heat generation because of oxidization, which was reported at ICONE14. The coolability was assured by a small amount of coolant supply because of high capillary force of oxidized fine particulate debris produced. In the present study, we examined the coolability of particulate debris bed deposited in narrower gap of 1 mm or 5 mm that coolant supply is hard. The particulate debris beds were piled up on the stainless steel sheet with 0.1 mm thickness, which was used to measure the bottom temperatures of particulate debris bed by using a thermo-video camera. We set up a heat supply section with heat input of 2.1 kW, which simulates the hard debris bed deposited on the particulate debris bed as reported for the TMI-2 accident. We measured the temperatures of the bottom surface of the heat supply section and the heat fluxes released into debris bed as well as the temperatures at the bottom of debris bed on the stainless steel sheet. It is found that when only the upper surface of particulate debris bed is in the film boiling, capillary force causes coolant supply to the particulate debris bed. Therefore, in the condition of thicker gap with small particulate debris, coolability of debris bed is improved. We find out that smaller particulate debris is moved by vapor movement. As a result, the area that high capillary force is caused because of

  11. Corium spreading: hydrodynamics, rheology and solidification of a high-temperature oxide melt

    International Nuclear Information System (INIS)

    Journeau, Ch.

    2006-06-01

    In the hypothesis of a nuclear reactor severe accident, the core could melt and form a high- temperature (2000-3000 K) mixture called corium. In the hypothesis of vessel rupture, this corium would spread in the reactor pit and adjacent rooms as occurred in Chernobyl or in a dedicated core-catcher s in the new European Pressurized reactor, EPR. This thesis is dedicated to the experimental study of corium spreading, especially with the prototypic corium material experiments performed in the VULCANO facility at CEA Cadarache. The first step in analyzing these tests consists in interpreting the material analyses, with the help of thermodynamic modelling of corium solidification. Knowing for each temperature the phase repartition and composition, physical properties can be estimated. Spreading termination is controlled by corium rheological properties in the solidification range, which leads to studying them in detail. The hydrodynamical, rheological and solidification aspects of corium spreading are taken into account in models and computer codes which have been validated against these tests and enable the assessment of the EPR spreading core-catcher concept. (author)

  12. Oxide-metal corium-concrete interaction test in the Vulcano facility

    International Nuclear Information System (INIS)

    Journeau, Ch.; Piluso, P.; Haquet, J.F.; Saretta, S.; Boccaccio, E.; Bonnet, J.M.

    2007-01-01

    Corium is likely to melt through the vessel and interact with the reactor pit concrete. Corium is made of a UO 2 -rich oxidic part, in which most of the decay heat is dissipated, and of a metallic part, mainly molten steel. An experiment has been set up in the Vulcano facility in which oxidic and metallic mixtures are molten in separate furnaces and poured in a concrete cavity. Induction heating is provided to the pool upper part thanks to shielding coils, so that, in case of stratification, the lighter oxidic corium-concrete mixture receives most of the power. Pre-calculations with the TOLBIAC-ICB corium-concrete interaction code based on the phase segregation model have provided valuable information for the dimensioning of this test: a thick metallic layer (>10 kg or 4 cm) has been chosen in order to obtain significant cavity ablation profiles depending on the selected heat transfer and stratification models. Stratification of the two liquid phases is predicted to occur in less than 10 minutes. In September 2006, the experiment was performed in the Vulcano facility. The corium was made of about 15 kg of steel at 1700 C and 30 kg of oxides (70% UO 2 , 16 % ZrO 2 and 14% concrete load) above 2000 C. It was poured in a limestone-rich concrete. This concrete type was selected for the first test, since the ablation is isotropic except for the initial transient, during oxidic corium-concrete interaction tests. 32 kW of induction power have been provided to the pool during the 4-hour test. The destruction of in-concrete thermocouples indicates that ablation was first mainly radial then became isotropic. This is quite similar to the ablation progression observed during previous tests with oxidic corium interacting with this type of concrete. Important 'volcanic activity' has been observed at the corium pool surface, compared to the previous oxidic corium experiments at Vulcano. (authors)

  13. An effect of corium composition variations on occurrence of a steam explosion in the TROI experiments

    International Nuclear Information System (INIS)

    Kim, J. W.; Park, I. K.; Hong, S. W.; Min, B. T.; Shin, Y. S.; Song, J. H.; Kim, H. D.

    2003-01-01

    Recently series of steam explosion experiments have been performed in the TROI facility using corium melts of various compositions. The compositions (UO 2 : ZrO 2 ) of the corium were 0 : 100, 50 : 50, 70 : 30, 80 : 20 and 87 : 13 in weight percent and the mass of the corium was about 10kg. An experiment using 0 : 100 corium (pure zirconia) caused a steam explosion. An experiment using 50 : 50 corium did not cause a steam explosion while a steam spike occurred in an experiment using 70 : 30 corium which was the eutectic point of corium. A steam spike is considered to be the fact that a triggering of a steam explosion occurred but a propagation process does not occur so as to cause a weak interaction. However, the possibility of a steam explosion with this composition can not be ruled out since many steam explosions occurred in the previous experiments. In the two experiments using 80 : 20 corium, a steam spike occurred in one experiment but no steam explosion occurred in the other experiment. However, the triggerability of a steam explosion with this composition is not clear since few steam explosions occurred in the previous experiments. And no steam explosion occurred in an experiment using 87 : 13 corium of which urania content was the greatest among the experiments performed in the TROI facility. From this, the possibility of a steam explosion or a steam spike is appeared to be high in the non-mush zone. It is considered that an explosive interaction could easily occur with the eutectic composition. Since the solidification temperature around the eutectic point is low, the melt is likely to maintain its liquid state at the time of triggering so as to cause an explosive phenomenon

  14. Considerations concerning the strategy of corium retention in the reactor vessel

    International Nuclear Information System (INIS)

    2015-01-01

    Third-generation nuclear reactors are characterised by consideration during design of core meltdown accidents. More specifically, dedicated measures or devices must be implemented to avoid basemat melt-through in the reactor building. These devices must have a high level of confidence. The strategy of corium retention in the reactor vessel, if supported by appropriate research and development, makes it possible to achieve this objective. IRSN works alone or in partnerships to address all the issues associated with in-vessel corium retention. This document describes the in-vessel corium retention strategy and its limitations, along with the research programs conducted by IRSN in this area

  15. Water boiling on the corium melt surface under VVER severe accident conditions

    International Nuclear Information System (INIS)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.

    1999-01-01

    Experimental results are presented on the interaction between corium melt and water supplied onto its surface. The tests were conducted on the Rasplav-2' experimental facility. Induction melting in a cold crucible was used to produce the melt. The following data have been obtained: heat transfer at water boiling on the melt surface, aerosol release, structure of the post-interaction solidified corium. The corium melt had the following composition, mass %: 60%UO 2 - 16%ZrO 2 - 15%Fe 2 O 3 - 6%Cr 2 O 3 -3%Ni 2 O 3 . The melt surface temperature was 1650-1700degC. (author)

  16. FCI experiments in the corium/water system

    Energy Technology Data Exchange (ETDEWEB)

    Huhtiniemi, I.; Hohmann, H.; Magallon, D.

    1995-09-01

    The KROTOS fuel coolant interaction (FCI) tests aim at providing benchmark data to examine the effect of fuel/coolant initial conditions and mixing on explosion energetics. Experiments, fundamental in nature, are performed in well-controlled geometries and are complementary to the FARO large scale tests. Recently, a new test series was started using 3 kg of prototypical core material (80 w/o UO{sub 2}, 20 w/o ZrO{sub 2}) which was poured into a water column of {le} 1.25 m in height (95 mm and 200 mm in diameter) under 0.1 MPa ambient pressure. Four tests have been performed in the test section of 95 mm in diameter (ID) with different subcooling levels (10-80K) and with and without an external trigger. Additionally, one test has been performed with a test section of 200 mm in diameter (ID) and with an external trigger. No spontaneous or triggered energetic FCIs (steam explosions) have been observed in these corium tests. This is in sharp contrast with the steam explosions observed in the previously reported Al{sub 2}O{sub 3} test series which had the same initial conditions of ambient pressure and subcooling. The post-test analysis of the corium experiments indicated that strong vaporisation at the melt/water contact led to a partial expulsion of the melt from the test section into the pressure vessel. In order to avoid this and to obtain a good penetration and premixing os the corium melt, an additional test has been performed with a larger diameter test section. In all the UO{sub 2}-ZrO{sub 2} tests an efficient quenching process (0.7-1.2 MW/kg-melt) with total fuel fragmentation (mass mean diameter 1.4-2.5 mm) was observed. Results from Al{sub 2}O{sub 3} tests under the same initial conditions are also presented for further confirmation of the observed differences in behaviour between Al{sub 2}O{sub 3} and UO{sub 2}-ZrO{sub 2} melts.

  17. Corium phase equilibria based on MASCA, METCOR and CORPHAD results

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V.; Granovsky, V.S.; Khabensky, V.B. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almiashev, V.I.; Mezentseva, L.P. [Grebenshikov Institute of Silicate Chemistry, Russian Academy of Sciences (ISCh RAS), St. Petersburg (Russian Federation); Krushinov, E.V.; Kotova, S.Yu.; Kosarevsky, R.A. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Barrachin, M. [Institut de Radioprotection et Surete Nucleaire IRSN/DPAM, St Paul lez Durance (France); Bottomley, D. [EUROPAISCHE KOMMISSION, Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Fichot, F. [Institut de Radioprotection et Surete Nucleaire IRSN/DPAM, St Paul lez Durance (France); Fischer, M. [AREVA NP GmbH, Erlangen (Germany)], E-mail: Manfred.Fischer@areva.com

    2008-10-15

    Experimental data on component partitioning between suboxidized corium melt and steel in the in-vessel melt retention (IVR) conditions are compared. The data are produced within the OECD MASCA program and the ISTC CORPHAD project under close-to-isothermal conditions and in the ISTC METCOR project under thermal gradient conditions. Chemical equilibrium in the U-Zr-Fe(Cr,Ni,...)-O system is reached in all experiments. In MASCA tests the molten pool formed under inert atmosphere has two immiscible liquids, oxygen-enriched (oxidic) and oxygen-depleted (metallic), resulting of the miscibility gap of the mentioned system. Sub-system data of the U-Zr-Fe(Cr,Ni,...)-O phase diagram investigated within the ISTC CORPHAD project are interpreted in relation with the MASCA results. In METCOR tests the equilibrium is established between oxidic liquid and mushy metallic part of the system. Results of comparison are discussed and the implications for IVR noted.

  18. Elements of thought on corium containment strategy in reactor vessel

    International Nuclear Information System (INIS)

    2015-01-01

    As accidents with core fusion are taken into account for the design of third-generation nuclear reactors, this brief document presents the corium containment strategy for a reactor vessel, its limitations, as well as research programs undertaken by the IRSN in this field. The report describes the controlled management of a severe accident, the major objective being to minimise releases in the environment, that which requires to maintain the reactor containment enclosure tightness. Practical actions are briefly indicated. Key points indicating the feasibility of a strategy of containment in vessel are discussed. The impact of reactor power on the robustness of an approach with containment in vessel is also discussed. An overview of technological evolutions and contributions of researches made by the IRSN is finally proposed

  19. Control Carbon to Prevent corium Stratification In-Vessel Retention

    Energy Technology Data Exchange (ETDEWEB)

    Go, A Ra; Hong, Seung Hyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-10-15

    As a result, the thermal margin decreases, and the nuclear reactor vessel may be destroyed. To control Carbons, which is the major cause of stratification, Ruthenium and Hafnium are inserted inside the lower reactor head which initiates a chemical reaction with Carbon. SPARTAN program is used to confirm a reaction probability which is measured in bond energy and strength etc. To analyze the possibility of bonding with Carbon, the initial property of Ruthenium and Carbon are measured during the calculated absorbing process. After following that theory, the Spartan program is able to determine if it can insert the metal. After verifying the combination of Ruthenium and Carbon, the Spartan program analyzes the impact of the Carbon to prevent the corium stratification. It determines the possibility of the success with the introduction of the IVR concept. Ruthenium is suitable to Carbon bonding process to decrease affect to corium behavior which do not form stratification. The metal which can combine with Carbon should be satisfied with temperature as high as 2800 .deg. C. Therefore, the further research works are determined by using the Spartan program to calculate the Carbon and Ruthenium bonding energy, and to check other bonding results as follows. After check the results, review this theory to insert the Ruthenium in reactor vessel. APR1400 and OPR1000, Korea Hydro and Nuclear power plant core meltdown accident has been evaluated a high level in severe accident. When the reactor core is melted down, it is stratified into the metal layer and the ceramic layer. As the heat conductivity of metal layer is higher than that of the ceramic layer, heat concentration occurs in the upper part of the bottom hemisphere which comes into contact with the metal layer.

  20. Thermal behavior of molten corium during TMI-2 core relocation event

    International Nuclear Information System (INIS)

    Anderson, J.L.; Sienicki, J.J.

    1988-01-01

    During the TMI-2 accident, a pool of molten corium formed in the central region of the core and was contained by solidified crusts. Failure of the crust surrounding the molten material, at approximately 224 min, resulted in a relocation of an estimated 20-25 tons of molten corium through peripheral fuel assemblies in the east side of the vessel, as well as through the core barrel assembly (CBA) at the periphery of the core. This paper presents the results of an analyses carried out to investigate the thermal interactions of molten corium with the CBA structures during the relocation event. The principal objectives of the analyses are: (a) to assess the potential for relocation to take place through the CBA versus the flow of molten core material directly downward through the core via the fuel assemblies; and (b) to understand the distribution of prior molten corium observed during vessel defueling examinations. 5 refs., 1 fig

  1. Water boiling on the corium melt surface under VVER severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Sulatsky, A.A.; Khabensky, V.B. [Sci. Res. Technol. Inst., Leningrad (Russian Federation); Lopukh, D.B.; Petrov, Y.B.; Pechenkov, A.Y. [St. Petersburg Electrotechnical University (SPbEU), Prof. Popov st 5/3, St. Petersburg (Russian Federation)

    2000-01-01

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the 'Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO{sub 2+x}-16% ZrO{sub 2}-15% Fe{sub 2}O{sub 3}-6% Cr{sub 2}O{sub 3}-3% Ni{sub 2}O{sub 3}. The melt surface temperature ranged within 1920-1970 K. (orig.)

  2. Quality improvements of thermodynamic data applied to corium interactions for severe accident modelling in SARNET2

    Czech Academy of Sciences Publication Activity Database

    Bakardjieva, Snejana; Barrachin, M.; Bechta, S.; Bezdička, Petr; Bottomley, D.; Brissoneau, L.; Cheynet, B.; Dugne, O.; Fischer, E.; Fischer, M.; Gusarov, V.; Journeau, C.; Khabensky, V.; Kiselová, M.; Manara, D.; Piluso, P.; Sheindlin, M.; Tyrpekl, V.; Wiss, T.

    2014-01-01

    Roč. 74, SI (2014), s. 110-124 ISSN 0306-4549 Institutional support: RVO:61388980 Keywords : Corium * Severe accidents * Thermodynamic database Subject RIV: CA - Inorganic Chemistry Impact factor: 0.960, year: 2014

  3. Two-dimensional interaction of oxidic corium with concretes: The VULCANO VB test series

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Christophe [CEA, DEN, STRI/LMA, Cadarache, F-13108 St Paul lez Durance (France)], E-mail: christophe.journeau@cea.fr; Piluso, Pascal; Haquet, Jean-Francois; Boccaccio, Eric; Saldo, Valerie; Bonnet, Jean-Michel; Malaval, Sophie; Carenini, Laure [CEA, DEN, STRI/LMA, Cadarache, F-13108 St Paul lez Durance (France); Brissonneau, Laurent [CEA, DEN, STPA/LPC, Cadarache, F-13108 St Paul lez Durance (France)

    2009-10-15

    Three two-dimensional Molten Core-Concrete Interaction tests have been conducted in the VULCANO facility with prototypic oxidic corium. The major finding is that for the two tests with silica-rich concrete, the ablation was anisotropic while it was isotropic for limestone-rich concrete. The cause of this behaviour is not yet well understood. Post Test Examinations have indicated that for the silica-rich concrete, the corium melt mixed specifically with mortar, while, for limestone-rich concretes, the analysed samples were in accordance with a corium-concrete mixing. The experimental results are described and compared to numerical codes. Separate Effect Tests with Artificial Concretes and prototypic corium are proposed to understand the phenomena governing the ablation geometry.

  4. Two-dimensional interaction of oxidic corium with concretes: The VULCANO VB test series

    International Nuclear Information System (INIS)

    Journeau, Christophe; Piluso, Pascal; Haquet, Jean-Francois; Boccaccio, Eric; Saldo, Valerie; Bonnet, Jean-Michel; Malaval, Sophie; Carenini, Laure; Brissonneau, Laurent

    2009-01-01

    Three two-dimensional Molten Core-Concrete Interaction tests have been conducted in the VULCANO facility with prototypic oxidic corium. The major finding is that for the two tests with silica-rich concrete, the ablation was anisotropic while it was isotropic for limestone-rich concrete. The cause of this behaviour is not yet well understood. Post Test Examinations have indicated that for the silica-rich concrete, the corium melt mixed specifically with mortar, while, for limestone-rich concretes, the analysed samples were in accordance with a corium-concrete mixing. The experimental results are described and compared to numerical codes. Separate Effect Tests with Artificial Concretes and prototypic corium are proposed to understand the phenomena governing the ablation geometry.

  5. Thermalhydraulic Phenomena in Corium Pools: Numerical Simulation with TOLBIAC and Experimental Validation with BALI

    International Nuclear Information System (INIS)

    Bernaz, L.; Bonnet, J.M.; Spindler, B.; Villermaux, C.

    1999-01-01

    In the frame of severe accidents studies, the behavior of corium pools is simulated by the TOLBIAC code. After a short description of the model and peculiarities of the code, its capacities are illustrated with results of the simulation of the behavior of a corium pool in a core catcher made of concrete. The BALI experiments and first results are then presented, and finally BALI tests simulation with TOLBIAC. (authors)

  6. Interaction of concretes with oxide + metal corium. The VULCANO VBS series

    International Nuclear Information System (INIS)

    Journeau, Christophe; Bonnet, Jean-Michel; Ferry, Lionel; Haquet, Jean-Francois; Piluso, Pascal

    2009-01-01

    In the hypothetical case of a severe accident, the reactor core could melt and the formed mixture, called corium, could melt through the vessel and interact with the reactor pit concrete. Corium is made from a UO 2 -rich oxidic part, in which most of the decay heat is dissipated, and a metallic part, mainly molten steel. Up to now, due to experimental constraints, most of the experiments have been performed with solely oxidic prototypic corium, or where designed so that most of the simulated decay heat was dissipated in the metallic layer. An experimental program has been set up in the VULCANO facility in which oxidic and metallic mixtures are melted in separate furnaces and poured in a concrete cavity. Induction heating is provided to the pool upper part thanks to shielding coils, so that, in case of stratification, the lighter oxidic corium-concrete mixture receives most of the power. Three experiments have been conducted: one with a limestone-rich concrete and two with a silica-rich concrete. Metal stratification has been determined from modifications of the corium electrical properties in front of the inductor and is in good accordance with calculations. Concrete ablation has been monitored. A significant vertical ablation has been observed, even in case of silica-rich concretes, for which largely radial ablation has been observed in the case of pure oxidic corium melts. Post Test Examinations have shown unexpected repartitions of metal in the pool. (author)

  7. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector - 15271

    International Nuclear Information System (INIS)

    Saas, L.; Le Tellier, R.; Bajard, S.

    2015-01-01

    In this document, we present a simplified geometrical model (0D model) for both the in-core corium propagation transient and the characterization of the mode of corium transfer from the core to the vessel. A degraded core with a formed corium pool is used as an initial state. This initial state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate...). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4

  8. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Plevacova, K. [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Journeau, C., E-mail: christophe.journeau@cea.fr [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Piluso, P. [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Zhdanov, V.; Baklanov, V. [IAE, National Nuclear Centre, Material Structure Investigation Dept., Krasnoarmeiskaya, 10, Kurchatov City (Kazakhstan); Poirier, J. [CEMHTI, 1D, av. de la Recherche Scientifique, 45071 Orleans Cedex 2 (France)

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U{sub x}, Zr{sub y})O{sub 2-z} water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO{sub 2}, the zirconium carbide coating keeps its role of protective barrier with UO{sub 2}-Al{sub 2}O{sub 3} below 2000 deg. C but does not resist to a UO{sub 2}-Eu{sub 2}O{sub 3} mixture.

  9. Developing of two-dimensional model of the corium cooling and behavior with non-condensible gas injection

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung [Seoul National University, Seoul (Korea, Republic of)

    1997-07-01

    The purpose of this study is to understand the effect of the non-condensible gas injection into the molten corium on the heat transfer and dynamic behavior within the melt when molten core-concrete interaction occurs during the hypothetical severe accident. Corium behavior with gas injection effect is two phase fluid pattern in which droplet has dispersed gas phase in continuous liquid phase of corium. To analyze this behavior, two dimensional governing equation using the governing equation, the computer program is accomplished using the finite difference method and SIMPLER algorithm. And benchmarking calculation is performed for the KfK experiment, which consider the gas injection effect. After this pre-calculation, an analyses is performed with typical corium under severe accidents. It is concluded that the heat transfer within corium increases as the metal components of the corium and gas injection velocity increase. 88 refs., 23 tabs., 35 figs. (author)

  10. A study on the modeling of molten corium-concrete interaction

    International Nuclear Information System (INIS)

    Park, Soo Yong

    1994-02-01

    The phenomenon known as molten corium concrete interaction (MCCI) has been recognized as important aspects of severe reactor accidents. The potential hazard of a MCCI is the threat to the integrity of the containment building due to the possibility of a basemat melt through, containment overpressurization by noncondensible gases, or oxidation of combustible gases. Over the past several years, a large experimental and analytical effort has been under taken in corium-concrete interaction phenomena by several organization. The purpose of this paper is to investigate the previous analytical results and computer programs, and finally to establish a new stand alone model which can predict the corium-concrete interaction. A model to predict the behavior of molten corium-concrete interaction in the reactor cavity during vessel ruptured accidents is established. Gas film model, gas bubble model, slag model and periodic contact model are employed as a major heat transfer model between corium and concrete. Solidified debris crust is considered at the boundary of molten corium. Upon the experimental observations, no layer stratification is assumed due to the strong dispersion of the metallic melt in the oxidic phase. With the assumption of temperature profile within the corium pool and crust, the temperature distribution of concrete is found by explicit solution of heat conduction equation. The sideward heat transfer rate can be obtained by considering multiplication factor to the downward heat transfer rate. The multiplication factor is treated as a user input because of its large uncertainty. Comparisons are made with two large scale experiments, SURC-2 and BETA V3.3. There is a reasonable agreement in the corium temperature, erosion depth and gas generation between the experimental data and the predicted results with periodic contact model given the uncertainties in the input data or the measurement. The gas bubble model has the highest heat transfer coefficient, and the

  11. Fuel-coolant interaction visualization test for in-vessel corium retention external reactor vessel cooling (IVR-ERVC) condition

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong Ho; Song, Jin Ho; Hong, Seong Wan [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

  12. Investigation of the coolability of a continuous mass of relocated debris to a water-filled lower plenum. Technical report

    International Nuclear Information System (INIS)

    Rempe, J.L.; Wolf, J.R.; Chavez, S.A.; Condie, K.G.; Hagrman, D.L.; Carmack, W.J.

    1994-09-01

    This report documents work performed to support the development of an analytical and experimental program to investigate the coolability of a continuous mass of debris that relocates to a water-filled lower plenum. The objective of this program is to provide an adequate data base for developing and validating a model to predict the coolability of a continuous mass of debris relocating to a water-filled lower plenum. The model must address higher pressure scenarios, such as the TMI-2 accident, and lower pressure scenarios, which recent calculations indicate are more likely for most operating LWR plants. The model must also address a range of possible debris compositions

  13. Lessons learnt from FARO/TERMOS corium melt quenching experiments

    Energy Technology Data Exchange (ETDEWEB)

    Magallon, D.; Huhtiniemi, I.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

    1998-01-01

    The influence of melt quantity, melt composition, water depth and initial pressure on quenching is assessed on the basis of seven tests performed in various conditions in the TERMOS vessel of the FARO facility at JRC-Ispra. Tests involved UO{sub 2}-based melt quantities in the range 18-176 kg at a temperature of approximately 3000 K poured into saturated water. The results suggest that erosion of the melt jet column is an efficient contributor to the amount of break-up, and thus quenching, for large pours of corium melt. The presence of Zr metal in the melt induced a much more efficient quenching than in a similar test with no Zr metal, attributed to the oxidation of the Zr. Significant amounts of H{sub 2} were produced also in tests with pure oxidic melts (e.g. about 300 g for 157 kg melt). In the tests at 5.0 and 2.0 MPa good mixing with significant melt break-up and quenching was obtained during the penetration in the water. At 0.5 MPa, good penetration of the melt into the water could still be achieved, but a jump in the vessel pressurisation occurred when the melt contacted the bottom and part (5 kg) of the debris was re-ejected from the water. (author)

  14. Fragmentation and quench behavior of corium melt streams in water

    International Nuclear Information System (INIS)

    Spencer, B.W.; Wang, K.; Blomquist, C.A.; McUmber, L.M.; Schneider, J.P.

    1994-02-01

    The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a crucial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues which arise during the molten core relocation include: (i) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (ii) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (iii) the quench rate of the molten fuel through the water in the lower plenum, (iv) the steam generation and hydrogen generation during the interaction, (v) the transient pressurization of the primary system, and (vi) the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through -6) was performed in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam/hydrogen generation, and the posttest debris characteristics

  15. Thermal hydraulic study of a corium molten pool

    International Nuclear Information System (INIS)

    Pigny, S.; Grand, D.; Seiler, J.M.; Durin, M.

    1993-01-01

    The thermohydraulic behaviour of a mass of molten core is investigated, in the frame of PWR severe accidents studies. The corium may be located in the vessel lower head or in an external core-catcher. It is assumed to be present in the container instantaneously. Its motion is described by one velocity field. It may be homogeneous or made of two stratified fluids. The residual power is assumed to be constant and uniform in the UO 2 phase. The radiative losses and the external water-cooling are taken into account. The thermal resistance of a peripheral crust is considered. The influence of the crust on the pool geometry may be studied. The wall behaviour is analysed by a conduction calculation. The interest of a sacrificial layer is underlined, so as the necessity of a multicomponent multiphase model to study the behaviour of a core catcher. It is also concluded that some experiments are needed for code validation about volume heated natural convection and multiphase flows. (author). 14 figs., 3 refs

  16. Overview of LWR severe accident research activities at the Karlsruhe Institute of Technology

    International Nuclear Information System (INIS)

    Miassoedov, Alexei; Albrecht, Giancarlo; Foit, Jerzy-Jan; Jordan, Thomas; Steinbrück, Martin; Stuckert, Juri; Tromm, Walter

    2012-01-01

    The research activities in the light water reactor (LWR) severe accidents domain at Karlsruhe Institute of Technology (KIT) are concentrated on the in- and ex-vessel core melt behavior. The overall objective is to investigate the core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity and to the containment, corium concrete interaction and corium coolability in the reactor cavity, and hydrogen behaviour in reactor systems. The results of the experiments contribute to a better understanding of the core melt sequences and thus improve safety of existing and, in the long-term, of future reactors by severe accident mitigation measures and by safety installations where required. This overview paper describes the experimental facilities used at KIT for severe accident research and gives an overview of the main directions and objectives of the R&D work. (author)

  17. Oxide-metal corium-concrete interaction test in the Vulcano facility

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch.; Piluso, P.; Haquet, J.F.; Saretta, S.; Boccaccio, E.; Bonnet, J.M. [CEA Cadarache, Severe Accident Mastery experimental Lab. (DEN/DTN/STRI/LMA), 13 - Saint Paul lez Durance (France)

    2007-07-01

    Corium is likely to melt through the vessel and interact with the reactor pit concrete. Corium is made of a UO{sub 2}-rich oxidic part, in which most of the decay heat is dissipated, and of a metallic part, mainly molten steel. An experiment has been set up in the Vulcano facility in which oxidic and metallic mixtures are molten in separate furnaces and poured in a concrete cavity. Induction heating is provided to the pool upper part thanks to shielding coils, so that, in case of stratification, the lighter oxidic corium-concrete mixture receives most of the power. Pre-calculations with the TOLBIAC-ICB corium-concrete interaction code based on the phase segregation model have provided valuable information for the dimensioning of this test: a thick metallic layer (>10 kg or 4 cm) has been chosen in order to obtain significant cavity ablation profiles depending on the selected heat transfer and stratification models. Stratification of the two liquid phases is predicted to occur in less than 10 minutes. In September 2006, the experiment was performed in the Vulcano facility. The corium was made of about 15 kg of steel at 1700 C and 30 kg of oxides (70% UO{sub 2}, 16 % ZrO{sub 2} and 14% concrete load) above 2000 C. It was poured in a limestone-rich concrete. This concrete type was selected for the first test, since the ablation is isotropic except for the initial transient, during oxidic corium-concrete interaction tests. 32 kW of induction power have been provided to the pool during the 4-hour test. The destruction of in-concrete thermocouples indicates that ablation was first mainly radial then became isotropic. This is quite similar to the ablation progression observed during previous tests with oxidic corium interacting with this type of concrete. Important 'volcanic activity' has been observed at the corium pool surface, compared to the previous oxidic corium experiments at Vulcano. (authors)

  18. SULTAN test facility for large-scale vessel coolability in natural convection at low pressure

    International Nuclear Information System (INIS)

    Rouge, S.

    1997-01-01

    The SULTAN facility (France/CEA/CENG) was designed to study large-scale structure coolability by water in boiling natural convection. The objectives are to measure the main characteristics of two-dimensional, two-phase flow, in order to evaluate the recirculation mass flow in large systems, and the limits of the critical heat flux (CHF) for a wide range of thermo-hydraulic (pressure, 0.1-0.5 MPa; inlet temperature, 50-150 C; mass flow velocity, 5-4400 kg s -1 m -2 ; flux, 100-1000 kW m -2 ) and geometric (gap, 3-15 cm; inclination, 0-90 ) parameters. This paper makes available the experimental data obtained during the first two campaigns (90 , 3 cm; 10 , 15 cm): pressure drop differential pressure (DP) = f(G), CHF limits, local profiles of temperature and void fraction in the gap, visualizations. Other campaigns should confirm these first results, indicating a favourable possibility of the coolability of large surfaces under natural convection. (orig.)

  19. Modeling for evaluation of debris coolability in lower plenum of reactor pressure vessel

    International Nuclear Information System (INIS)

    Maruyama, Yu; Moriyama, Kiyofumi; Nakamura, Hideo; Hirano, Masashi

    2003-01-01

    Effectiveness of debris cooling by water that fills a gap between the debris and the lower head wall was estimated through steady calculations in reactor scale. In those calculations, the maximum coolable debris depth was assessed as a function of gap width with combination of correlations for critical heat flux and turbulent natural convection of a volumetrically heated pool. The results indicated that the gap with a width of 1 to 2 mm was capable of cooling the debris under the conditions of the TMI-2 accident, and that a significantly larger gap width was needed to retain a larger amount of debris within the lower plenum. Transient models on gap growth and water penetration into the gap were developed and incorporated into CAMP code along with turbulent natural convection model developed by Yin, Nagano and Tsuji, categorized in low Reynolds number type two-equation model. The validation of the turbulent model was made with the UCLA experiment on natural convection of a volumetrically heated pool. It was confirmed that CAMP code predicted well the distribution of local heat transfer coefficients along the vessel inner surface. The gap cooling model was validated by analyzing the in-vessel debris coolability experiments at JAERI, where molten Al 2 O 3 was poured into a water-filled hemispherical vessel. The temperature history measured on the vessel outer surface was satisfactorily reproduced by CAMP code. (author)

  20. Presentation of the Vulcano installation which uses a plasma transferred arc rotary furnace for corium melting

    International Nuclear Information System (INIS)

    Cognet, G.; Laffont, G.; Jegou, C.; Pierre, J.; Journeau, C.; Sudreau, F.; Roubaud, A.

    1998-01-01

    In the case of loss coolant accident, the reactor core could melt and turn into a mixture of uranium oxides, zirconium, iron and steel called corium. A large experimental program has been launched to study corium behaviour, to qualify solutions to stabilize it and to confine it in the reactor containment. The Vulcano installation has been designed to that purpose. It is made up of: i) a plasma transferred arc rotary furnace, ii) a testing surface covered with refractory materials, iii) an induction heating system in order to simulate the residual power of corium, iv) instrumentation devices such as video cameras, thermocouples, infra-red pyrometers and flowmeters, and v) a laboratory to perform chemical analysis of corium samples. The first experimental results show that a mixture of corium and concrete spreads better than expected. It seems that a low initial height of matter can produce a great distance flowing while having a chaotic behaviour. This characteristic suggests that the mixture acts as a Bingham type threshold fluid. (A.C.)

  1. Study of corium radial spreading between fuel rods in a PWR core

    International Nuclear Information System (INIS)

    Roche, S.; Gatt, J.M.

    1996-01-01

    In the framework of severe accident studies for PWR like Three Mile Island Unit 2 (TMI-2), the reactor core essentially constituted of fuel rods begins to heat and then to melt. During the early degradation phase, a melt (essentially UO2 and ZrO2) that constitutes the corium flows first along the rods, and after a blockage formation, may radially propagate towards the core periphery. A simplified model has been elaborated to study the corium freezing phenomena during its crossflow between the fuel rods. The corium spreads on an horizontal support made, of either a corium crust, or a grid assembly. The model solves numerically the interface energy balance equation at the solid-liquid corium interface and the monodimensional heat balance equation in transient process with convective terms and heat source (residual power). ''Zukauskas'' correlations are used to calculate heat transfer coefficients. The model can be integrated in severe accident codes like ICARE II (IPSN) describing the in-vessel degradation scenarios. (author). 5 refs, 10 figs

  2. Effect of a blockage length on the coolability during reflood in a 2 × 2 rod bundle with a 90% partially blocked region

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kihwan, E-mail: kihwankim@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Kim, Byung-Jae, E-mail: byoungjae@kaeri.re.kr [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseoung-Gu, Daejeon 34134 (Korea, Republic of); Choi, Hae-Seob, E-mail: hschoi@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Highlights: • This test was conducted to understand the effect of blockage length on the coolability. • Reflood tests were conducted with blockage simulators for various reflood rates. • The coolability in the downstream of the blockage region is significantly enhanced. - Abstract: If fuel rods are ballooned or rearranged during the reflood phase of a large break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the transient heat transfer behavior is entirely different with those of the intact fuel rods owing to the deformed blockage region. The coolability in the blocked region depends on a complex two-phase heat transfer with various thermal hydraulic conditions. In addition, the blockage characteristics, such as the blockage ratio, length, shape, and configurations, are also significant factors affecting the coolability. In the present study, reflood experiments were carried out to understand the effect of the blockage length upon the coolability by varying the reflooding rates. The experiments were performed in electrically heated 2 × 2 rod bundles with blockage simulators having the same blockage ratio but different blockage lengths. The characteristics of quenching and heat transfer were evaluated to investigate the influence of the blockage region on the coolability. The droplet behaviors were also observed by measuring the droplets velocity and size near the blockage region. The coolability in the downstream region of the blockage was significantly enhanced, owing to the reduced flow area of the sub-channel, intensification of turbulence, and the entrained droplets in the blockage region.

  3. Water boiling on the corium melt surface under VVER severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V. [Research Institute of Technology, Sosnovy Bor (NITI) (RU)] [and others

    1999-07-01

    Experimental results are presented on the interaction between corium melt and water supplied onto its surface. The tests were conducted on the Rasplav-2' experimental facility. Induction melting in a cold crucible was used to produce the melt. The following data have been obtained: heat transfer at water boiling on the melt surface, aerosol release, structure of the post-interaction solidified corium. The corium melt had the following composition, mass %: 60%UO{sub 2}- 16%ZrO{sub 2}- 15%Fe{sub 2}O{sub 3} - 6%Cr{sub 2}O{sub 3}-3%Ni{sub 2}O{sub 3}. The melt surface temperature was 1650-1700degC. (author)

  4. Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

    Energy Technology Data Exchange (ETDEWEB)

    Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

  5. Physico-Chemistry and Corium Properties for In-Vessel Retention

    International Nuclear Information System (INIS)

    Froment, K.; Seiler, J.M.; Gueneau, C.; Dauvois, V.; Barbier, F.; Bellon, M.; Tourasse, M.; Ducros, G.; Cognet, G.; Sudreau, F.

    1999-01-01

    This paper focuses on some important aspects of consequences of material behaviour and interactions on in-vessel retention capabilities. It discusses the behaviour of corium oxide mixtures at elevated temperatures (miscibility gap and density effects, separation due to density effects in the solid-liquid mixture according to the analysis of the Rasplav experiment results), and then the interaction between metallic layer and vessel wall (physical-chemical interaction of corium with the carbon steel vessel wall, migration of low melting point metallic elements in the solid vessel wall). It proposes a mode for the calculation of melt viscosity (liquid phase viscosity and viscosity in the solidification range), addresses the issue of barium release and residual power and of distribution of the residual power in an oxidic corium

  6. Transient stratification modelling of a corium pool in a LWR vessel lower head

    International Nuclear Information System (INIS)

    Le Tellier, R.; Saas, L.; Bajard, S.

    2015-01-01

    Highlights: • A kinetic stratification model is proposed for the simulation of the in-vessel corium behaviour during a LWR severe accident. • The different associated “modes” of vessel failure by thermal focusing effect are highlighted and discussed. • A sensitivity study for a 1650 MWe GenIII PWR is presented with this model in order to illustrate the associated R&D issues. - Abstract: In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650 MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters

  7. VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation)], E-mail: bechta@sbor.spb.su; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almiashev, V.I. [Institute of Silicate Chemistry, Russian Academy of Sciences (ISCh RAS), St. Petersburg (Russian Federation); Lopukh, D.B. [SPb State Electrotechnical University (SPbGETU), St. Petersburg (Russian Federation); Bottomley, D. [EUROPAISCHE KOMMISSION, Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA/DEN/DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Rossendorf (FZR), Dresden (Germany); Fichot, F. [IRSN/DPAM/SEMCA, St. Paul lez Durance (France); Kymalainen, O. [FORTUM Nuclear Services Ltd., Espoo (Finland)

    2009-06-15

    The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

  8. VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

    International Nuclear Information System (INIS)

    Bechta, S.V.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

    2009-01-01

    The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

  9. Modeling of the corium cooling and loading factor analysis for containment during severe accidents

    International Nuclear Information System (INIS)

    Konoval, A.V.; Kalvand, Ali.; Kazachkov, I.V.

    2013-01-01

    The paper is devoted to the development and study of the mathematical model for corium melt interaction with low-temperature melting blocks in the passive protection systems (PPS) against severe accidents at the NPP, and learning the peculiarities of construction and operation of the PPS. The configurations of cooling blocks' distributions considered and the results of their work in the corium cooling pool are compared to the data of other PPS's conceptions. The conclusion is made that the models developed and the results obtained may be useful for constructing the PPS against severe accidents

  10. Modeling of the corium cooling and loading factor analysis for containment during severe accidents

    Directory of Open Access Journals (Sweden)

    O. V. Konoval

    2013-09-01

    Full Text Available The paper is devoted to the development and study of the mathematical model for corium melt interaction with low-temperature melting blocks in the passive protection systems (PPS against severe accidents at the NPP, and learning the peculiarities of construction and operation of the PPS. The configurations of cooling blocks’ distributions considered and the results of their work in the corium cooling pool are compared to the data of oth-er PPS’s conceptions. The conclusion is made that the models developed and the results obtained may be useful for constructing the PPS against severe accidents.

  11. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.E.

    1986-01-01

    A model has been developed to calculate the expansion and fragmentation of a corium jet, due to the evolution of dissolved gas, during a postulated core meltdown accident. Parametric calculations have been performed for a PWR high pressure accident scenario. Jet breakup occurs within a few jet diameters from the RPV. The diameter of the fragmented jet at the level of the reactor cavity floor is predicted to be 40-130 times the discharge diameter. Particles generated by fragmentation of corium melt are predicted to be in the 30-150 μm size range

  12. Coolability of a 3D homogeneous debris bed, experimental and numerical investigations

    International Nuclear Information System (INIS)

    Atkhen, K.; Berthoud, G.

    2001-01-01

    Within the framework of nuclear safety analysis, we present here experimental and numerical results in the field of debris bed coolability. Experimental data are provided by the SILFIDE 3D experimental facility in which the debris bed is heated by induction, at Electricite de France (EDF). Numerical computations are obtained with MC3D-REPO which is a 3-phase and 3D code developed by the Commissariat a l'Energie Atomique (CEA). The uniform debris bed consists of 2 and 3,17 mm diameter steel beads contained in a 50 cm x 60 cm x 10 cm vessel. Water is used as a coolant and can be introduced either by the top or the bottom of the bed at a determined temperature. Due to heterogeneous power distribution within the bed, two definitions for the critical heat flux are proposed: the classical mean value and the local flux (much higher). Even in the first case, the measured dryout heat flux is higher than the Lipinsky 1-D flux. Temperature curve analyses show that the dryout phenomenon is very local, therefore one should be careful about the right flux definition to use. As the injected power is being increased stepwise, steady temperature stages above saturation temperature before dryout can be observed. A discussion is proposed. For some very high values of the induction power, some spheres melted together, leading to a bigger non-porous region. Even if the local temperature went over 1300 C, the bed was still coolable and the critical heat flux value was not impacted. Some parametric studies led to the following conclusions: bottom coolant injection leads to a twice time higher critical flux than by top injection, the influence of the height of the water pool above debris bed is negligible, a sub-cooled liquid injection has no influence on the coolability. Fluidization of surface particles is also discussed. The MC3D-REPO model assumes a thermal equilibrium between the three phases, which gives results in agreement with experiments until dryout occurs. (author)

  13. Experimental investigation of coolability behaviour of irregularly shaped particulate debris bed

    International Nuclear Information System (INIS)

    Kulkarni, P.P.; Rashid, M.; Kulenovic, R.; Nayak, A.K.

    2010-01-01

    In case of a severe nuclear reactor accident, the core can melt and form a particulate debris bed in the lower plenum of the reactor pressure vessel (RPV). Due to the decay heat, the particle bed, if not cooled properly, can cause failure of the RPV. In order to avoid further propagation of the accident, complete coolability of the debris bed is necessary. For that, understanding of various phenomena taking place during the quenching is important. In the frame of the reactor safety research, fundamental experiments on the coolability of debris beds are carried out at IKE with the test facility 'DEBRIS'. In the present paper, the boiling and dry-out experimental results on a particle bed with irregularly shaped particles mixed with stainless steel balls have been reported. The pressure drops and dry-out heat fluxes of the irregular-particle bed are very similar to those for the single-sized 3 mm spheres bed, despite the fact that the irregular-particle bed is composed of particles with equivalent diameters ranging from 2 to 10 mm. Under top-flooding conditions, the pressure gradients are all smaller than the hydrostatic pressure gradient of water, indicating an important role of the counter-current interfacial drag force. For bottom-flooding with a liquid inflow velocity higher than about 2.7 mm/s, the pressure gradient generally increases consistently with the vapour velocity and the fluid-particle drag becomes important. The system pressures (1 and 3 bar) have negligible effects on qualitative behaviour of the pressure gradients. The coolability of debris beds is mainly limited by the counter-current flooding limit (CCFL) even under bottom-flooding conditions with low flow rates. The system pressure and the flow rate are found to have a distinct effect on the dry-out heat flux. Different classical models have been used to predict the pressure drop characteristics and the dry-out heat flux (DHF). Comparisons are made among the models and experimental results for

  14. A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

    International Nuclear Information System (INIS)

    Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

    2012-01-01

    A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

  15. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

    1997-01-01

    For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

  16. Corium spreading: hydrodynamics, rheology and solidification of a high-temperature oxide melt; L'etalement du corium: hydrodynamique, rheologie et solidification d'unbain d'oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch

    2006-06-15

    In the hypothesis of a nuclear reactor severe accident, the core could melt and form a high- temperature (2000-3000 K) mixture called corium. In the hypothesis of vessel rupture, this corium would spread in the reactor pit and adjacent rooms as occurred in Chernobyl or in a dedicated core-catcher s in the new European Pressurized reactor, EPR. This thesis is dedicated to the experimental study of corium spreading, especially with the prototypic corium material experiments performed in the VULCANO facility at CEA Cadarache. The first step in analyzing these tests consists in interpreting the material analyses, with the help of thermodynamic modelling of corium solidification. Knowing for each temperature the phase repartition and composition, physical properties can be estimated. Spreading termination is controlled by corium rheological properties in the solidification range, which leads to studying them in detail. The hydrodynamical, rheological and solidification aspects of corium spreading are taken into account in models and computer codes which have been validated against these tests and enable the assessment of the EPR spreading core-catcher concept. (author)

  17. Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

    1995-09-01

    Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

  18. Prototypic corium oxidation and hydrogen release during the Fuel-Coolant Interaction

    Czech Academy of Sciences Publication Activity Database

    Tyrpekl, J.; Piluso, P.; Bakardjieva, Snejana; Nižňanský, D.; Rehspringer, J.L.; Bezdička, Petr; Dugne, O.

    2015-01-01

    Roč. 75, JAN (2015), s. 210-218 ISSN 0306-4549 Institutional support: RVO:61388980 Keywords : Corium * Fuel -Coolant Interaction * Hydrogen release * Material effect * Nuclear reactor severe accident Subject RIV: CA - Inorganic Chemistry Impact factor: 1.174, year: 2015

  19. Validation of ASTEC V2 models for the behaviour of corium in the vessel lower head

    International Nuclear Information System (INIS)

    Carénini, L.; Fleurot, J.; Fichot, F.

    2014-01-01

    The paper is devoted to the presentation of validation cases carried out for the models describing the corium behaviour in the “lower plenum” of the reactor vessel implemented in the V2.0 version of the ASTEC integral code, jointly developed by IRSN (France) and GRS (Germany). In the ASTEC architecture, these models are grouped within the single ICARE module and they are all activated in typical accident scenarios. Therefore, it is important to check the validity of each individual model, as long as experiments are available for which a single physical process is involved. The results of ASTEC applications against the following experiments are presented: FARO (corium jet fragmentation), LIVE (heat transfer between a molten pool and the vessel), MASCA (separation and stratification of corium non miscible phases) and OLHF (mechanical failure of the vessel). Compared to the previous ASTEC V1.3 version, the validation matrix is extended. This work allows determining recommended values for some model parameters (e.g. debris particle size in the fragmentation model and criterion for debris bed liquefaction). Almost all the processes governing the corium behaviour, its thermal interaction with the vessel wall and the vessel failure are modelled in ASTEC and these models have been assessed individually with satisfactory results. The main uncertainties appear to be related to the calculation of transient evolutions

  20. A review of dryout heat fluxes and coolability of particle beds. APRI 4, Stage 2 Report

    International Nuclear Information System (INIS)

    Lindholm, Ilona

    2002-04-01

    were studied. Significant amount of data with prototypic material tests exists. All of the tests show significant fragmentation in case of deep subcooled pool. An additional observation is that no energetic melt coolant interaction (steam explosion) has been reported for prototypic materials. A set of most relevant data for reactor applications have been chosen. Based on this, a general particle size distribution has been constructed. The average particle size obtained by this way was about 3.5 mm. Information from fragmentation and dryout tests and the Lipinski 0-D correlation have been utilised to assess the debris bed coolability for the Olkiluoto severe accident scenario. The calculation shows that for well-mixed beds with 3.5 mm particles the dryout heat flux would be close to 1 MW/m 2 , well above the estimated heat flux due to decay heat. Stratification of finer particles on top of the bed due to e.g. a steam explosion would reduce the dryout heat flux to 50-200 kW/m 2 . This would be below heat fluxes produced by decay heat in Nordic BWRs. The key uncertainty considering particle bed coolability is due to the particle size distribution and stratification. If the possibility of a thick fine particle layer on top of the bed can be ruled out, the particulate debris bed in Nordic BWRs will be coolable. A rough estimate of melt pool coolability in Nordic BWRs has also been conducted. The MACE and COTELS experimental data have been summarised. Based on the data, the melt pools in the pedestal are slowly coolable. The concrete erosion does not threaten the containment failure margins, except maybe at Forsmark 1 and 2 units. Release of non-condensable gases during MCCI may cause an earlier start of filtered venting in Olkiluoto, Forsmark and Oskarshamn 3 plants

  1. Improvement and evaluation of debris coolability analysis module in severe accident analysis code SAMPSON using LIVE experiment

    International Nuclear Information System (INIS)

    Wei, Hongyang; Erkan, Nejdet; Okamoto, Koji; Gaus-Liu, Xiaoyang; Miassoedov, Alexei

    2017-01-01

    Highlights: • Debris coolability analysis module in SAMPSON is validated. • Model for heat transfer between melt pool and pressure vessel wall is modified. • Modified debris coolability analysis module is found to give reasonable results. - Abstract: The purpose of this work is to validate the debris coolability analysis (DCA) module in the severe accident analysis code SAMPSON by simulating the first steady stage of the LIVE-L4 test. The DCA module is used for debris cooling in the lower plenum and for predicting the safety margin of present reactor vessels during a severe accident. In the DCA module, the spreading and cooling of molten debris, gap cooling, heating of a three-dimensional reactor vessel, and natural convection heat transfer are all considered. The LIVE experiment is designed to investigate the formation and stability of melt pools in a reactor pressure vessel (RPV). By comparing the simulation results and experimental data in terms of the average melt pool temperature and the heat flux along the vessel wall, a bug is found in the code and the model for the heat transfer between the melt pool and RPV wall is modified. Based on the Asfia–Dhir and Jahn–Reineke correlations, the modified version of the DCA module is found to give reasonable results for the average melt pool temperature, crust thickness in the steady state, and crust growth rate.

  2. Ex-vessel nuclear fuel transfer system

    International Nuclear Information System (INIS)

    1977-01-01

    A system is described for transferring reactor fuel assemblies between a fuel storage area and a fuel transfer area while the fuel assemblies remain completely submerged in a continuous body of coolant. The invention relates particularly to sodium cooled fast breeder reactors. (UK)

  3. Ex-vessel nuclear fuel transfer system

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    A system for transferring fuel assemblies between a fuel transfer area and a fuel storage area while the fuel assemblies remain completely submerged in a continuous body of coolant is described. A fuel transfer area filled with reactor coolant communicating with the reactor vessel below the reactor coolant level provides a transfer area for fuel assemblies in transit to and from the reactor vessel. A positioning mechanism comprising at least one rotatable plug disposed on a fuel transfer tank located outside the reactor vessel cooperates with either the fuel transfer area or the fuel storage area to position a fuel assembly in transit. When in position, a transporting mechanism cooperating with the positioning mechanism lifts or lowers a chosen fuel assembly. The transporting mechanism together with the positioning mechanism are capable of transferring a fuel assembly between the fuel transfer area and the fuel storage area

  4. Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

    International Nuclear Information System (INIS)

    Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong; Ha, Kwang Soon; Kim, Hwan-Yeol

    2014-01-01

    Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

  5. Investigation on influence of crust formation on VULCANO VE-U7 corium spreading with MPS method

    International Nuclear Information System (INIS)

    Yasumura, Yusan; Yamaji, Akifumi; Furuya, Masahiro; Ohishi, Yuji; Duan, Guangtao

    2017-01-01

    Highlights: • The new crust formation model was developed for the MPS spreading analysis code. • The VULCANO VE-U7 corium spreading experiment was analyzed by the developed code. • The termination of the spreading was governed by the crust formation at the leading edge. - Abstract: In a severe accident of a light water reactor, the corium spreading behavior on a containment floor is important as it may threaten the containment vessel integrity. The Moving Particle Semi-implicit (MPS) method is one of the Lagrangian particle methods for simulation of incompressible flow. In this study, the MPS method is further developed to simulate corium spreading involving not only flow, but also heat transfer, phase change and thermo-physical property change of corium. A new crust formation model was developed, in which, immobilization of crust was modeled by stopping the particle movement when its solid fraction is above the threshold and is in contact with the substrate or any other immobilized particles. The VULCANO VE-U7 corium spreading experiment was analyzed by the developed MPS spreading analysis code to investigate influences of different particle sizes, the corium viscosity changes, and the “immobilization solid fraction” of the crust formation model on the spreading and its termination. Viscosity change of the corium was influential to the overall progression of the spreading leading edge, whereas termination of the spreading was primarily determined by the immobilization of the leading edge (i.e., crust formation). The progression of the leading edge and termination of the spreading were well predicted, but the simulation overestimated the substrate temperature. Further investigations may be necessary for the future study to see if thermal resistance at the corium-substrate boundary has significant influence on the overall spreading behavior and its termination.

  6. Study on effective particle diameters and coolability of particulate beds packed with irregular multi-size particles

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W.; Kudinov, P.; Bechta, S. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    One of the key questions in severe accident research is the coolability of the debris bed, i.e., whether decay heat can be completely removed by the coolant flow into the debris bed. Extensive experimental and analytical work has been done to substantiate the coolability research. Most of the available experimental data is related to the beds packed with single size (mostly spherical) particles, and less data is available for multi-size/irregular-shape particles. There are several analytical models available, which rely on the mean particle diameter and porosity of the bed in their predictions. Two different types of particles were used to investigate coolability of particulate beds at VTT, Finland. The first type is irregular-shape Aluminum Oxide gravel particles whose sizes vary from 0.25 mm to 10 mm, which were employed in the STYX experiment programme (2001-2008). The second type is spherical beads of Zirconium silicate whose sizes vary between 0.8 mm to 1 mm, which were used in the COOLOCE tests (Takasuo et al., 2012) to study the effect of multi-dimensional flooding on coolability. In the present work, the two types of particles are used in the POMECO-FL and POMECO-HT test facility to obtain their effective particle diameters and dryout heat flux of the beds, respectively. The main idea is to check how the heaters' orientations (vertical in COOLOCE vs. horizontal in POMECO-HT) and diameters (6 mm in COOLOCE vs. 3 mm in POMECO-HT) affect the coolability (dryout heat flux) of the test beds. The tests carried out on the POMECO-FL facility using a bed packed with aluminum oxide gravel particles show the effective particle diameter of the gravel particles is 0.65 mm, by which the frictional pressure gradient can be predicted by the Ergun equation. After the water superficial velocity is higher than 0.0025 m/s, the pressure gradient is underestimated. The effective particle diameter of the zirconium particles is found as 0.8 mm. The dryout heat flux is measured on

  7. Study on effective particle diameters and coolability of particulate beds packed with irregular multi-size particles

    International Nuclear Information System (INIS)

    Thakre, S.; Ma, W.; Kudinov, P.; Bechta, S.

    2013-08-01

    One of the key questions in severe accident research is the coolability of the debris bed, i.e., whether decay heat can be completely removed by the coolant flow into the debris bed. Extensive experimental and analytical work has been done to substantiate the coolability research. Most of the available experimental data is related to the beds packed with single size (mostly spherical) particles, and less data is available for multi-size/irregular-shape particles. There are several analytical models available, which rely on the mean particle diameter and porosity of the bed in their predictions. Two different types of particles were used to investigate coolability of particulate beds at VTT, Finland. The first type is irregular-shape Aluminum Oxide gravel particles whose sizes vary from 0.25 mm to 10 mm, which were employed in the STYX experiment programme (2001-2008). The second type is spherical beads of Zirconium silicate whose sizes vary between 0.8 mm to 1 mm, which were used in the COOLOCE tests (Takasuo et al., 2012) to study the effect of multi-dimensional flooding on coolability. In the present work, the two types of particles are used in the POMECO-FL and POMECO-HT test facility to obtain their effective particle diameters and dryout heat flux of the beds, respectively. The main idea is to check how the heaters' orientations (vertical in COOLOCE vs. horizontal in POMECO-HT) and diameters (6 mm in COOLOCE vs. 3 mm in POMECO-HT) affect the coolability (dryout heat flux) of the test beds. The tests carried out on the POMECO-FL facility using a bed packed with aluminum oxide gravel particles show the effective particle diameter of the gravel particles is 0.65 mm, by which the frictional pressure gradient can be predicted by the Ergun equation. After the water superficial velocity is higher than 0.0025 m/s, the pressure gradient is underestimated. The effective particle diameter of the zirconium particles is found as 0.8 mm. The dryout heat flux is measured on

  8. In-Vessel Coolability. Workshop Proceedings, in collaboration with EC-SARNET

    International Nuclear Information System (INIS)

    2011-01-01

    Severe Accident Management Guidelines increase focus on containment integrity after some progression in the course of a severe accident. This change in priorities is made according to criteria that vary depending on reactor type and specific procedures. Once a water source has been recovered, different accident management strategies can be used: send water into the core and/or cool the reactor pressure vessel (RPV) externally. It should be noticed that, depending on the amount of water available, these strategies might conflict with other uses of water such as for instance activating spray systems in the containment or may have deleterious effects as for instance an increase in the production of hydrogen. Generally, for in-vessel reflooding, the models used for evaluation of accident management measures suffer from a lack of validation. Given this background, the objectives of the workshop were: -) to exchange information on different Severe Accident Management strategies used or contemplated for the in-vessel coolability issue; -) to review recent, ongoing and planned experimental programmes on reflooding; -) to review models used for reflooding in severe accident calculation tools, either simplified or sophisticated; -) to exchange information on the treatment of reflooding in different safety studies such as Probabilistic Safety Assessment; and -) to provide recommendations for future work, as necessary

  9. Experimental studies on the coolability of packed beds. Flooding of hot dry packed beds

    International Nuclear Information System (INIS)

    Leininger, S.; Kulenovic, R.; Laurien, E.

    2013-01-01

    In case of a severe accident in a nuclear power plant meltdown of the reactor core can occur and form a packed bed in the lower plenum of the reactor pressure vessel (RPV) after solidification due to contact with water. The removal of after-heat and the long-term coolability is of essential interest. The efficient injection of cooling water into the packed bed has to be assured without endangering the structural integrity of the reactor pressure vessel. The experiments performed aimed to study the dry-out and the quenching (flooding) of hot dry packed beds. Two different inflow variants, bottom- and top-flooding including the variation of the starting temperature of the packed bed and the injection rate were studied. In case of bottom flooding the quenching time increases with increasing packed bed temperature and decreasing injection rate. In case of top flooding the flow pattern is more complex, in a first phase the water flows preferentially toward the RPV wall, the flow paths conduct the water downwards. The flow resistance of the packed bed increases with increasing bed temperatures. The quenching temperatures increase significantly above average.

  10. KATS experiments to simulate corium spreading in the EPR core catcher concept

    International Nuclear Information System (INIS)

    Eppinger, B.; Fieg, G.; Schuetz, W.; Stegmaier, U.

    2001-01-01

    In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher de-vices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent cooling by flooding with water. Therefore a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were chemically inert ceramic layers, dry concrete and concrete with a shallow water layer on top. (authors)

  11. Corium Spreading Over Concrete: The Vulcano VE-U7 and VE-U8 Tests

    International Nuclear Information System (INIS)

    Journeau, Christophe; Boccaccio, Eric; Fouquart, Pascal; Jegou, Claude; Piluso, Pascal

    2002-01-01

    Two experiments have been performed in the VULCANO facility in which prototypic corium has been spread over concrete. In the VE-U7 test, a mixture representative of what can be expected at the opening of EPR reactor-pit gate has been spread on siliceous concrete and on a reference channel in inert refractory ceramic. The spreading progression was not much affected by the presence of concrete and sparging gases. In the VE-U8 test, a UO 2 -ZrO 2 mixture, prototypic of in-vessel corium, has been spread over a lime-siliceous concrete. Although residual power was not simulated in this experiment, up to 2 cm of concrete have been eroded during the test. Results in terms of spreading behaviour, effects of gases, concrete erosion and thermal attack are presented and discussed. (authors)

  12. Numerical analysis of vapor explosion in the system 'corium-water'

    International Nuclear Information System (INIS)

    Melikhov, O.I.; Melikhov, V.I.; Sokolin, A.V.

    2000-01-01

    The thermal detonation taking into account the microinteraction processes model has been applied to study thermal detonation wave escalation and propagation in the corium-water mixture. Transient escalation stage and subsequent steady-state propagation stage of the thermal detonation have been calculated. The essential decrease of the escalation length in comparison with the previous results calculated without microinteraction concept has been obtained. (author)

  13. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  14. Escalation and propagation of thermal detonation in the corium-water systems

    International Nuclear Information System (INIS)

    Melikhov, O.I.; Melikhov, V.I.; Sokolin, A.V.

    2001-01-01

    The thermal detonation taking into account micro-interaction processes model has been applied to study thermal detonation wave escalation and propagation in the corium-water mixture. Transient escalation stage and subsequent steady-state propagation stage of the thermal detonation have been calculated. The essential decrease of the escalation length in comparison with the previous results calculated without micro-interaction concept has been obtained. (authors)

  15. Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

    International Nuclear Information System (INIS)

    An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho

    2016-01-01

    For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B 4 C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

  16. Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B{sub 4}C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

  17. Proposal of In-vessel corium retention concept for Paks NPP

    International Nuclear Information System (INIS)

    Elter, J.; Toth, E.; Matejovic, P.

    2011-01-01

    The in-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) seems to be a promising severe accident management strategy not only for new generation of advanced PWRs, but also for VVER-440/V213 reactors, which were designed several years ago. The basic idea of in-vessel retention of corium is to prevent RPV failure by flooding the reactor cavity so that the reactor pressure vessel is submerged in water up to its support structures, and thus the decay heat can be transferred from the corium pool through the vessel wall and into the water surrounding the vessel. An IVR concept with simple ECVR loop based only on minor modifications of existing plant technology was proposed for the Paks Nuclear Power Plant. 2 severe accident (LB and SB LOCA) without availability of HP and LP safety injection in power upgrade (108%) conditions were simulated using the ASTEC code. The analyses show that the proposed solution is effective in preserving RPV integrity in the case of severe accident. Possible uncertainties in code predictions are covered by the applied conservative assumptions

  18. Experimental Study of Interactions Between Sub-oxidized Corium and Reactor Vessel Steel

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Granovsky, V.S.; Krushinov, E.V.; Vitol, S.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Tromm, W.; Miassoedov, A.; Bottomley, D.; Fischer, M.; Piluso, P.; Altstadt, E.; Willschutz, H.G.; Fichoti, F.

    2006-01-01

    One of the critical factors in the analysis of in-vessel melt retention is the vessel strength. It is, in particular, sensitive to the thickness of intact vessel wall, which, in its turn, depends on the thermal conditions and physicochemical interactions with corium. Physicochemical interaction of prototypic UO 2 -ZrO 2 -Zr corium melt and VVER vessel steel was examined during the 2. Phase of the ISTC METCOR Project. Rasplav-3 test facility was used for conducting four tests, in which the Zr oxidation degree and interaction front temperature were varied; in one of the tests, stainless steel was added to the melt. Direct experimental measurements and post-test analyses were used for determining corrosion kinetics and maximum corrosion depth (i.e. the physicochemical impact of corium on the cooled vessel steel specimens), as well as the steel temperature conditions during the interaction, and finally the structure and composition of crystallized ingots, including the interaction zone. The minimum temperature on the interaction front boundary, which determined its final position and maximum corrosion depth was ∼ 1090 deg. C. An empirical correlation for calculation of corrosion kinetics has been derived. (authors)

  19. A comparative analysis of molten corium-concrete interaction models employed in MELCOR and MAAP codes

    International Nuclear Information System (INIS)

    Park, Soo Yong; Song, Y. M.; Kim, D. H.; Kim, H. D.

    1999-03-01

    The purpose of this report are to identify the modelling differences by review phenomenological models related to MCCI, and to investigate modelling uncertainty by performing sensitivity analysis, and finally to identify models to be improved in MELCOR. As the results, the most important uncertain parameter in the MCCI area is the debris stratification/mixing, and heat transfer between molten corium and overlying water pool. MAAP has a very simple and flexible corium-water heat transfer model, which seems to be needed in MELCOR for evaluation of real plants as long as large phenomenological uncertainty still exists. During the corium-concrete interaction, there is a temperature distribution inside basemat concrete. This would affect the amount or timing of gas generation. While MAAP calculates the temperature distribution through nodalization methodology, MELCOR calculates concrete response based on one-dimensional steady-state ablation, with no consideration given to conduction into the concrete or to decomposition in advanced of the ablation front. The code may be inaccurate for analysis of combustible gas generation during MCCI. Thus there is a necessity to improve the concrete decomposition model in MELCOR. (Author). 12 refs., 5 tabs., 42 figs

  20. Behavior of concrete in contact with molten corium in the case of a hypothetical core melt accident

    International Nuclear Information System (INIS)

    Peehs, M.; Skokan, A.; Reimann, M.

    1979-01-01

    The temperature-dependent properties of basaltic and limestone concrete as needed for predicting Corium melt propagation in concrete (elongation behavior, specific heat and degradation enthalpy, thermal diffusivity, and conductivity) are determined experimentally together with the chemical and physical reactions occurring in heated concrete. The determined oxidation potential of -335 kJ/mole for molten Corium interacting with the concrete is in accordance with the observed H 2 generation due to the melt internal oxidation of zirconium, chromium, and iron. The liquefaction temperatures of the different concretes investigated are approx. 1300 to 1400 0 C. The relatively high degradation enthalpy of basaltic and limestone concrete is the reason for the barrier effect of concrete against propagating molten Corium

  1. Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Almjashev, V.I. [Alexandrov Scientific-Research Technology Institute (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [KTH, Stockholm (Sweden); Gusarov, V.V. [SPb State Technology University (SPbGTU), St. Petersburg (Russian Federation); Barrachin, M. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [EC-Joint Research Centre, Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Cadarache, St Paul lez Durance (France)

    2014-10-15

    Highlights: • The METCOR facility simulates vessel steel corrosion in contact with corium. • Steel corrosion rates in UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} coria accelerate above 1050 K. • However corrosion rates can also be limited by melt O{sub 2} supply. • The impact of this on in-vessel retention (IVR) strategy is discussed. - Abstract: During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 °C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR)

  2. Assessment of models for steam release from concrete and implications for modeling corium behavior in reactor cavities

    International Nuclear Information System (INIS)

    Washington, K.E.; Carroll, D.E.

    1988-01-01

    Models for concrete outgassing have been developed and incorporated into a developmental version of the CONTAIN code for the assessment of corium behavior in reactor cavities. The resultant code, referred to as CONTAIN/OR in order to distinguish it from the released version of CONTAIN, has the capability to model transient heat conduction and concrete outgassing in core-concrete interaction problems. This study focused on validation and assessment of the outgassing model through comparisons with other concrete response codes. In general, the model is not mechanistic; however, there are certain important processes and feedback effects that are treated rigorously. The CONTAIN outgassing model was compared against two mechanistic concrete response codes (USINT and SLAM). Gas release and temperature profile predictions for several concrete thicknesses and heating rates were performed with acceptable agreement seen in each case. The model was also applied to predict corium behavior in a reactor cavity for a hypothetical severe accident scenario. In this calculation, gases evolving from the concrete during nonablating periods fueled exothermic Zr chemical reactions in the corium. Higher corium temperatures and more concrete ablation were observed when compared with that seen when concrete outgassing was neglected. Even though this result depends somewhat upon the makeup of the corium sources and the concrete type in the cavity, it does show that concrete outgassing can be important in the modeling of corium behavior in reactor cavities. In particular, the need to expand the traditional role of CORCON from steady-state ablation to the consideration of more transient events is clearly evident as a result of this work. 5 refs., 11 figs., 1 tab

  3. Thermophysical properties of liquid UO2, ZrO2 and corium by molecular dynamics and predictive models

    International Nuclear Information System (INIS)

    Kim, Woong Kee; Shim, Ji Hoon; Kaviany Massoud

    2016-01-01

    The analysis of such accidents (fate of the melt), requires accurate corium thermophysical properties data up to 5000 K. In addition, the initial corium melt superheat melt, determined from such properties, are key in predicting the fuel-coolant interactions (FCIs) and convection and retention of corium in accident scenarios, e.g., core-melt down corium discharge from reactor pressure vessels and spreading in external core-catcher. Due to the high temperatures, data on molten corium and its constituents are limited, so there are much data scatters and mostly extrapolations (even from solid state) have been used. Here we predict the thermophysical properties of molten UO 2 and ZrO 2 using classical molecular dynamics (MD) simulations (properties of corium are predicted using the mixture theories and UO 2 and ZrO 2 properties). The thermophysical properties (density, compressibility, heat capacity, viscosity and surface tension) of liquid UO 2 and ZrO 2 are predicted using classical molecular dynamics simulations, up to 5000 K. For atomic interactions, the CRG and the Teter potential models are found most appropriate. The liquid behavior is verified with the random motion of the constituent atoms and the pair-distribution functions, starting with the solid phase and raising the temperature to realize liquid phase. The viscosity and thermal conductivity are calculated with the Green-Kubo autocorrelation decay formulae and compared with the predictive models of Andrade and Bridgman. For liquid UO 2 , the CRG model gives satisfactory MD predictions. For ZrO 2 , the density is reliably predicted with the CRG potential model, while the compressibility and viscosity are more accurately predicted by the Teter model

  4. Molten Corium-Concrete Interaction Behavior Analyses for Severe Accident Management in CANDU Reactor

    International Nuclear Information System (INIS)

    Choi, Y.; Kim, D. H.; Song, Y. M.

    2014-01-01

    After the last few severe accidents, the importance of accident management in nuclear power plants has increased. Many countries, including the United States (US) and Canada, have focused on understanding severe accidents in order to identify ways to further improve the safety of nuclear plants. It has been recognized that severe accident analyses of nuclear power plants will be beneficial in understanding plant-specific vulnerabilities during severe accidents. The objectives of this paper are to describe the molten corium behavior to identify a plant response with various concrete specific components. Accident analyses techniques using ISSAC can be useful tools for MCCI behavior in severe accident mitigation

  5. The modeling and analysis of in-vessel corium/structure interaction in boiling water reactors

    International Nuclear Information System (INIS)

    Podowski, M.Z.; Kurul, N.; Kim, S.-W.; Baltyn, W.; Frid, W.

    1997-01-01

    A complete stand-alone state-of-the-art model has been developed of the interaction between corium debris in the lower plenum and the RPV walls and internal structures, including the vessel failure mechanisms. This new model has been formulated as a set of consistent computer modules which could be linked with other existing models and/or computer codes. The combined lower head and lower plenum modules were parametrically tested and applied to predict the consequences of a hypothetical station blackout in a Swedish BWR. (author)

  6. Vulcano: a dedicated R and D program to master corium recuperation for future reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bouchter, J.C.; Cognet, G.

    1994-12-31

    In the field of Severe Accident studies for future Nuclear Power Plants, the CEA (Commissariat a l`Energie Atomique) has launched an important program operating with UO{sub 2} materials. General objectives cover the qualification of industrial core-catcher concepts as well as the improvement of the understanding of corium behaviour inside the pressure vessel. After a presentation of the general scope of the project, the paper focuses on the first experimental phase (VULCANO E-30) which deals with major questions of core-catcher concepts based on spreading and flooding principles. (authors). 3 refs., 6 figs.

  7. Vulcano: a dedicated R and D program to master corium recuperation for future reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Cognet, G.

    1994-01-01

    In the field of Severe Accident studies for future Nuclear Power Plants, the CEA (Commissariat a l'Energie Atomique) has launched an important program operating with UO 2 materials. General objectives cover the qualification of industrial core-catcher concepts as well as the improvement of the understanding of corium behaviour inside the pressure vessel. After a presentation of the general scope of the project, the paper focuses on the first experimental phase (VULCANO E-30) which deals with major questions of core-catcher concepts based on spreading and flooding principles. (authors). 3 refs., 6 figs

  8. Improved Design of PECS to reduce Flow Instability for EU-APR1400

    International Nuclear Information System (INIS)

    Hwang, Do Hyun; Lee, Keunsung

    2014-01-01

    For EU-APR1400, PECS (Passive Ex-vessel corium retaining and Cooling System), so-called core catcher, was adopted to keep the integrity of basemat in containment by preventing MCCI (Molten Core Concrete Interaction) through retaining core debris and cooling corium outside the reactor vessel. In this paper, the improved design of PECS is presented to increase coolability by reducing flow instability in the region of cooling channel. In this paper, flow instability analysis was carried out using CFD code to find out the most improved design of PECS, which is to increase coolability by reducing bubble entrainment in the region of cooling channel. The reduction of bubble entrainment in the downcomer facilitates higher mass flow rates in the downcomer. Among presented four designed for the downcomer of PECS, the superstep design shows the highest mass flow rate and the lowest gas holdup in the downcomer as well as in the cooling channel. Compared with the existing design, the elimination of the horizontal part and the addition of an extra space above the vertical entrance to the downcomer seem to help the separation of the vapor

  9. VULCANO: a large scale U O2 program to study corium behaviour and cooling for future reactors

    International Nuclear Information System (INIS)

    Cognet, G.; Bouchter, J.C.

    1994-01-01

    The CEA has launched the VULCANO project, a large experimental facility whose objectives are the understanding of corium behaviour from core melting up to vessel melt-through, and the qualification of core-catcher concepts. This paper deals with the strategy adopted to overcome the difficulties of such experiments (use of real materials such as U O 2 , controlled temperature and flowrate...); in particular, it describes the feasibility studies undertaken on corium production, and on sustained heating within the melt (micro-waves). Some indications are also given on scaling studies for experiments devoted to vessel integrity. 7 figs., 3 refs

  10. Numerical Analysis of Molten Corium Dispersion during Hypothetical High-Pressure Accidents in APR1400 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kim, Jong Tae; Ha, Kwang Soon; Kim, Sang Baik; Kim, Hee Dong; Jeong, Jae Sik

    2010-01-01

    During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by the following jet of a high pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet with very high velocity and is released into the upper compartment of the NPP by an overpressure in the cavity. The heat-carrying fragments of the corium transfer the thermal energy to the ambient air in the containment and react chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. New generation NPPs such as APR1400 and EPR have been designed in consideration of reducing the possibility of the containment failure from the DCH. In order for that, APR1400 has a convolute-type corium chamber connected to the reactor cavity. In the case of EPR, severe-accident dedicated depressurization valves are installed to preclude a high pressure melt ejection (HPME). DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical reaction. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. The corium dispersion rates for many types of the NPP containments had been obtained by experiments in 90s. And some correlations from the experimental data were developed. As mentioned above, APR1400 has a corium chamber to reduce the corium dispersion rate. But there is no experimental data for the dispersion rate specific to the APR1400 cavity geometry. So its performance for capturing of the dispersed corium

  11. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.

    1987-01-01

    This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals and gas bubble nucleation in molten metals are relevant problems which are addressed in this work. Models are developed for jet expansion, primary breakup of the jet and secondary fragmentation of melt droplets resulting from violent effervescence of dissolved gas. The jet expansion model is based on a general relation for bubble growth which includes both inertia-controlled and diffusion-controlled growth phases. The jet expansion model is able to predict the jet void fraction, jet radius as a function of axial distance from the pressure vessel, bubble size and bubble pressure. The number density of gas bubbles in the melt, which is a basic parameter in the model, was determined experimentally and is about 10 8 per m 3 of liquid. The primary breakup of the jet produces a spray of droplets, about 2-3 mm in diameter. Parametric calculations for a TMLB' reactor accident sequence show that the corium jet is disrupted within a few initial jet diameters from the reactor vessel and that the radius of corium spray at the level of the reactor cavity floor is in the range of 0.8 to 2.6 m. (orig./HP)

  12. Investigation of molten corium-concrete interaction phenomena and aerosol release

    International Nuclear Information System (INIS)

    Spencer, B.W.; Thompson, D.H.; Armstrong, D.R.; Fink, J.K.; Gunther, W.H.; Kilsdonk, D.J.; Sehgal, B.R.

    1987-01-01

    The Electric Power Research Institute is sponsoring a program of laboratory investigations at Argonne National Laboratory to study the interaction between molten core materials and reactor concrete basemats during postulated severe reactor accidents, with particular emphasis on measurements of the magnitude and chemical species present in the aerosol releases. The approach in this program is to sustain internal heat generation in reactor-material corium using direct electrical heating and to develop test operating and diagnostics capabilities with a series of small- and intermediate-scale scoping tests followed by fully instrumented large-scale testing. Real reactor materials (UO 2 , ZrO 2 , oxides of stainless steel, plus metallics) are used, with small amounts of La 2 O 3 , BaO, and SrO added to simulate nonvolatile fission products. In intermediate-scale scoping tests completed to date, corium inventories of up to 29 kg have been heated with power inputs in excess of 1 kW/kg melt. The measured concrete ablation rates have ranged from 0.9 to 3.9 mm/minute. Aerosol samples have been examined using a scanning electron microscope and show submicron particles, 2-6 micrometer spheres, and agglomerates that range from a few micrometers to string 13 micrometers in length

  13. Evaluation of In-Vessel Corium Retention under a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon; Kang, Kyung-Ho; Ha, Kwang-Soon; Kim, Jong-Tae; Koo, Kil-Mo; Cho, Young-Ro; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

    2008-02-15

    The current study on In-Vessel corium Retention and its application activities to the actual nuclear power plant have been reviewed and discussed in this study. Severe accident sequence which determines an initial condition of the IVR has been evaluated and late phase melt progression, heat transfer on the outer reactor vessel, and in-vessel corium cooling mechanism have been estimated in detail. During the high pressure sequence of the reactor coolant system, a natural circulation flow of the hot steam leads to a failure of the pressurizer surge line before the reactor vessel failure, which leads to a rapid decrease of the reactor coolant system pressure. The results of RASPLAV/MASCA study by OECD/NEA have shown that a melt stratification has occurred in the lower plenum of the reactor vessel. In particular, laver inversion has occurred, which is that a high density of the metal melt moves to the lower part of the oxidic melt layer. A method of heat transfer enhancement on the outer reactor vessel is an optimal design of the reactor vessel insulation for an increase of the natural circulation flow between the outer reactor vessel and the its insulation, and an increase of the critical Heat flux on the outer reactor vessel by using various method, such as Nono fluid, coated reactor vessel, and so on. An increase method of the in-vessel melt cooling is a development of the In-vessel core catcher and a decrease of focusing effect in the metal layer.

  14. Interaction between molten corium UO2+x-ZrO2-FeOy and VVER vessel steel

    International Nuclear Information System (INIS)

    Bechta, S. V.; Granovsky, V. S.; Khabensky, V. B.; Krushinov, E. V.; Vitol, S. A.; Sulatsky, A. A.; Gusarov, V. V.; Almiashev, V. I.; Lopukh, D. B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

    2010-01-01

    In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO 2+x -ZrO 2 -FeO y melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction Zone. The available experimental data have been used to develop a correlation for the corrosion rate as a function of temperature and heat flux. (authors)

  15. A study on corium melt pool behavior under external vessel cooling : investigation of the first phase research results in the OECD RASPLAV project

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Kim, Sang Baik; Kim, Hee Dong; Yoo, Kun Joong

    1998-04-01

    The scope and contents of the OECD RASPLAV program are to investigate natural convection heat transfer in the corium, chemical and mechanical interaction between the corium and the reactor vessel, crust formation of the corium, and thermal behaviour of the corium by experiments and model development during external vessel cooling to prevent reactor vessel failure in severe accidents of nuclear power plant. This study includes evaluation and analysis of the RASPLAV V phase I results for three years between July 1, 1994 and June 30, 1997. These results supply technical basis for our experimental program on severe accident research. Two large-scale experiments of RASPLAV-AW-between the corium and the reactor vessel. Several small-scale experiments were conducted to analyze thermal stratification in the corium. The salt experiments were conducted to estimate the crust and the mushy region formation, and natural convection heat transfer in the corium. In the analytical studies, pre and post analysis of the RASPLAV-AW-200 experiments and evaluation of the salt test results have been performed using CONV 2 and 3D computer codes, which were developed during RASPLAV program phase I. Low density corium was separated from the high density corium during the RASPLAV-AW-200 tests and the TULPAN test, which was a new finding in the RASPLAV project phase I. From the salts test, heat flux distribution in the side wall heating case is similar to the direct internal heat generation case, and the crust formation is a little effect on heat transfer rate. The results of CONV 2 and 3 D were very well with with the experimental results. The results of RASLAV project phase I, such as furnace design and the techniques on fuel melting, are very helpful to our severe accident experimental program. (author). 57 refs., 13 tabs., 52 figs.

  16. New set of convective heat transfer coefficients established for pools and validated against CLARA experiments for application to corium pools

    Energy Technology Data Exchange (ETDEWEB)

    Michel, B., E-mail: benedicte.michel@irsn.fr

    2015-05-15

    Highlights: • A new set of 2D convective heat transfer correlations is proposed. • It takes into account different horizontal and lateral superficial velocities. • It is based on previously established correlations. • It is validated against recent CLARA experiments. • It has to be implemented in a 0D MCCI (molten core concrete interaction) code. - Abstract: During an hypothetical Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) severe accident with core meltdown and vessel failure, corium would fall directly on the concrete reactor pit basemat if no water is present. The high temperature of the corium pool maintained by the residual power would lead to the erosion of the concrete walls and basemat of this reactor pit. The thermal decomposition of concrete will lead to the release of a significant amount of gases that will modify the corium pool thermal hydraulics. In particular, it will affect heat transfers between the corium pool and the concrete which determine the reactor pit ablation kinetics. A new set of convective heat transfer coefficients in a pool with different lateral and horizontal superficial gas velocities is modeled and validated against the recent CLARA experimental program. 155 tests of this program, in two size configurations and a high range of investigated viscosity, have been used to validate the model. Then, a method to define different lateral and horizontal superficial gas velocities in a 0D code is proposed together with a discussion about the possible viscosity in the reactor case when the pool is semi-solid. This model is going to be implemented in the 0D ASTEC/MEDICIS code in order to determine the impact of the convective heat transfer in the concrete ablation by corium.

  17. Analysis on the Multiplication Factor with the Change of Corium Mass and Void Fraction

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Hae Sun; Park, Chang Je; Song, Jin Ho; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The neutron absorbing materials and fuel rods would be separately arranged and relocated, since the control materials in metallic structures have lower melting points than that of the oxide fuel (UO{sub 2}) rod materials. In addition, core reflood for a BWR is normally accomplished by supplying unborated water unlikely for a PWR. Therefore, a potential for a recriticality event to occur may exist, if unborated coolant injection is initiated with this configuration in the reactor core. The re-criticality in this system, however, brings into question what the uranium mass is required to achieve a critical level. Furthermore, the additional decay heat from molten fuel (corium) will produce an increase of void and eventually results in under-moderation of neutrons. The prior verification of these consequential physical variations in criticality eigenvalue (effective multiplication factor, k{sub eff}) should be greatly contributed to control and termination of re-criticality. Therefore, this study addresses what uranium mass of corium could achieve re-criticality of an accident core, and how effect the coolant void fraction has on eigenvalue (k{sub eff}) and its reactivity. To analyze the critical mass and the effect on criticality upon changing coolant density, k{sub eff} values were calculated using the MCNPX 2.5.0 code, and the reactivity change was also investigated. As a result, a large change in corium mass leads to a little change in k{sub eff} value, nevertheless, only about 60 kg of uranium is necessary to achieve a critical level. Thus, the amounts to reach a re-criticality are not fairly large, considering the actual uranium quantities loaded in the reactor core. Based on the condition with k{sub eff} greater than unity, the absolute values of k{sub eff} decrease rate and the coolant density coefficient were gradually increased due to the steady increments of coolant void (i.e., decrease in coolant density). In addition, the k{sub eff} value approaches the

  18. Differences between silica and limestone concretes that may affect their interaction with corium

    International Nuclear Information System (INIS)

    Journeau, C.; Haquet, J. F.; Piluso, P.; Bonnet, J. M.

    2008-01-01

    Recent Molten Core Concrete Interaction tests performed at Argonne National Laboratory and at CEA Cadarache have shown that, whereas the ablation of limestone-rich concretes is almost isotropic, the ablation of silica-rich concretes is much faster towards the sides than towards the bottom of the cavity. The following differences exists between limestone-rich and silica-rich concretes: limestone concretes liberate about twice as much gas, at a given ablation rate than siliceous concretes (more than 50% more at constant heat flux) and this can affect pool hydraulics and crust stability: limestone concrete has a higher liquidus temperature than siliceous concrete and molten limestone concrete has a larger diffusion coefficient and can more easily dissolve a corium crust than siliceous melt; limestone aggregates are destroyed by de-carbonation at around 1000 K while silica aggregates melt only above 2000 K, so that floating silica aggregates can form cold spots increasing corium solidification near the interface; de-carbonation of limestone leads to a significant shrinkage of concrete melt volume compared to the cold solid that hampers the mechanical stability of overlying crusts; the chemical composition of molten mortar (sand + cement) and concrete (sand + gravel + cement) is close for limestone-rich concretes while it is different for siliceous concretes, so that the melt composition may vary significantly in case of non-simultaneous melting of the siliceous concrete constituents; molten silicates have a large viscosity, so that transport properties are different for the two types of concretes. The small range of plant concrete compositions that have been considered for MCCI experiments has not yet been found sufficient to determine which of the above-mentioned differences is paramount to explain the observed difference in ablation patterns. Separate Effect Tests using specially-designed 'artificial concretes' and prototypic corium would provide the necessary

  19. Influence of corium oxidation on fission product release from molten pool

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V., E-mail: bechta@sbor.spb.s [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Krushinov, E.V.; Vitol, S.A.; Khabensky, V.B.; Kotova, S.Yu.; Sulatsky, A.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almyashev, V.I. [Grebenschikov Institute of Silicate Chemistry of the Russian Academy of Sciences (ISC RAS), St. Petersburg (Russian Federation); Ducros, G.; Journeau, C. [CEA, DEN, Cadarache, F-13108 St. Paul lez Durance (France); Bottomley, D. [Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Clement, B. [Institut de Radioprotection et Surete Nucleaire (IRSN), St. Paul lez Durance (France); Herranz, L. [CIEMAT, Madrid (Spain); Guentay, S. [PSI, Wuerenlingen (Switzerland); Trambauer, K. [GRS, Muenchen (Germany); Auvinen, A. [VTT, Espoo (Finland); Bezlepkin, V.V. [SPbAEP, St. Petersburg (Russian Federation)

    2010-05-15

    Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in a cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions.

  20. An interpretation of the observations performed on rasplav AW200-1 corium

    International Nuclear Information System (INIS)

    Froment, K.; Seiler, J.M.; CEA Centre d'Etudes de Grenoble, 38

    1997-01-01

    The RASPLAV test AW-200-1, performed in the Kurchatov Institute in 1996, showed unexpected results: elevated measured temperatures and stratification of the C-22 corium. Thermalhydraulic and thermodynamic calculations allowed us to give some explanation of the phenomenon which took place in the device: due to the large range between the solidus and the liquidus temperature of the initial mixture, and due to the density difference between the liquid and the solid phase in this temperature domain separation of these two phases had happened during the melting of the mixture (we have no explanation why this separation occurred). GEMINI2 calculations of the solidification paths are consistent with metallographic analyses which were carried out in these two separated layers after solidification. (author)

  1. Influence of corium oxidation on fission product release from molten pool

    International Nuclear Information System (INIS)

    Bechta, S.V.; Krushinov, E.V.; Vitol, S.A.

    2009-01-01

    Release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate and aerosol particle composition. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA are set. (author)

  2. Assessment of In-vessel corium retention for VVER-440/V213

    International Nuclear Information System (INIS)

    Matejovic, P.; Barnak, M.; Bachraty, M.; Berky, R.

    2011-01-01

    In-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) has been recognised as a feasible and promising severe accident management strategy for VVER-440/V213 reactors. In general, the avoiding of boiling crisis on outer (cooled) RPV (reactor pressure vessel) surface is sufficient condition for preserving the RPV integrity. The crucial point of the proposed IVR concept for VVER-440/V213 is the narrow gap between elliptical lower head and thermal and biological shield. In the cold conditions the width of this gap is only about 2 cm and would be even lower in hot IVR conditions, when the reactor wall is subjected to large thermal gradients due to temperature difference between the hot inner surface (loaded by corium) and cold outer surface (which is cooled by water in flooded cavity). Sufficient gap should remain free for coolant flow for the success of the proposed IVR concept. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width are of primarily importance. Two different approaches were used for the estimation of the thermal load: a conservative approach and a transient approach, both were computed with the ASTEC code. The structural analysis of RPV subjected to IVR load was performed using the finite element method (FEM) code ANSYS release 10.0. From the results obtained it follows, that even when the RPV is subjected to limiting loading conditions during severe accident, there should be sufficient gap width (∼ 1 cm) between RPV wall and thermal/biological shield for the coolant flow in natural circulation regime alongside the outer surface of the RPV wall

  3. Molten corium concrete interaction: investigation of heat transfer in two-phase flow

    International Nuclear Information System (INIS)

    Amizic, Milan

    2014-01-01

    In the context of severe accident research for the second and the third generation of nuclear power plants, there are still open issues concerning some aspects of the concrete cavity ablation during the molten corium - concrete interaction (MCCI). The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat melt through. For the purpose of experimental investigation of thermal hydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the small pool configuration (50 cm * 25 cm * 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s and the superficial gas velocity is varied up to 8 cm/s. This thesis comprises a brief description of MCCI phenomenology, literature reviews on the existing heat transfer correlations for two phase flow and the void fraction, a description of CLARA setup, experimental results and their interpretation. The experimental results are compared with existing models and some new models for the assessment of heat transfer coefficient in two-phase flow. (author) [fr

  4. Thermo-physical properties of corium: development of an assessed data base for severe accident applications

    Energy Technology Data Exchange (ETDEWEB)

    Strizhov, V.F.; Galimov, R.G.; Ozrin, V.D. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation); Yu Zitserman, V.; Kobzev, G.I.; Fokin, L.R. [Institute of high temperatures, Russian Academy of Sciences, Moscow (Russian Federation); Piluso, P. [CEA Cadarache (DEN/DTN/STRI), Lab. d' essais pour la Maitrise des Accidents graves, 13 - Saint Paul lez Durance (France); Chalaye, H. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif sur Yvette (France)

    2007-07-01

    In a hypothetical case of a core melt-down scenarios a very high temperature would be reached (up to 3000 K). In this case, the materials of the core and structural materials (fuel, cladding, metallic alloys, concrete, etc.) could melt to form complex and aggressive mixtures called corium. Modelling of severe accident phenomena, code development and assessments of nuclear safety require a reliable knowledge of the thermophysical properties of corium at wide temperature range (below solidus temperature, between solidus and liquidus temperature and above the liquidus temperature). Common Russian-French project ISTC 3078, has been devoted to the development, assessment and recommendation for the establishment of a reliable thermophysical data base for severe accident applications. The project consists of two tasks related to properties of pure metallic (U, Zr, Fe, Cr, Ni) and oxide (UO{sub 2}, U{sub 3}O{sub 8}, U{sub 4}O{sub 9}, NiO, ZrO{sub 2}, Cr{sub 2}O{sub 3}, FeO, Fe{sub 2}O{sub 3}, Fe{sub 3}O{sub 4}, Al{sub 2}O{sub 3}, CaO, MgO, SiO{sub 2}, HfO{sub 2}, CeO{sub 2}) components, and mixtures relevant to severe accident conditions. Three categories of data (on UPAK classification) were considered: experimental data, critically evaluated data, and predicted data. The data of the first category is a result of specific experiment, data of the second category is a result of the analysis of data consistency and co-processing (expert and statistical) obtained in several experiments, data of the third category are based on model estimates, using correlations between different physical properties. The process of assessing, review and development of recommendation is described in the paper and illustrated by examples on thermophysical properties. (authors)

  5. Analysis of B4C influences on thermodynamic properties and phase separation of molten corium with ionic liquid U-Zr-Fe-O-B-C-FPs database

    International Nuclear Information System (INIS)

    Fukasawa, Masanori; Tamura, Shigeyuki; Saito, Masaki

    2009-01-01

    Boron carbide influences on thermodynamic properties and phase separation of molten corium such as liquidus temperature were estimated with our U-Zr-Fe-O-B-C-FPs thermodynamic database. The liquidus temperature of the oxide for the typical corium was estimated to increase by a hundred degrees with B 4 C addition when the corium included up to 10 wt% Fe. On the other hand, the liquidus temperature was hardly changed when the corium included 50 wt% Fe. The interaction temperature between the steel and the corium with B 4 C was estimated at 1130 K. We define the interaction temperature as the lowest temperature where the solid Fe and the liquid phase of a corium are in equilibrium, at which interactions such as microstructure change of the vessel were observed in test studies. Although it is 180 K lower than that without B 4 C, the estimated temperature is still over 200 K higher than the criterion temperature where the vessel loses its structural strength, which has been used in the feasibility evaluation of the in-vessel retention. Other thermodynamic influences of B 4 C were also estimated as not having a negative impact on the in-vessel retention. (author)

  6. Experimental study on coolability of particulate core-metal debris bed with oxidization, (2). Fragmentation and enhanced heat transfer in zircaloy debris bed

    International Nuclear Information System (INIS)

    Su, Guanghui; Sugiyama, Ken-ichiro; Aoki, Hiroomi; Kimura, Iichi

    2006-01-01

    The oxidization and coolability characteristics of the particulate Zircaloy debris bed, which is deposited under the hard debris and through which first vapor penetrates and then water penetrates, are studied in the present paper. In the vapor penetration experiments, it is found that Zircaloy debris particles are effectively broken into small pieces after making thick oxidized layer with deep clacks by rapid oxidization under the condition that vapor with 20 cm/s penetrates for 30 to 70 min at an initial debris bed temperature of 1,030degC. It is also confirmed in the water penetration experiments that the oxidized particle debris bed has potentially of high coolability when water penetrates through the fully oxidized particle bed because of a high capillary force originating from those particles with deep cracks on their surfaces. Based on the present study, a new scenario for the appearance and disappearance of the hot spot in the TMI-2 accident is possible. The particulate core-metal core-metal debris bed is first heated up by rapid oxidization with heat generation when vapor can penetrate through the debris bed with porosities. This corresponds to the appearance of the hot spot. The resultant oxidized particulate debris bed causes a high coolability due to its high capillary force when the water can touch the debris bed at wet condition. This corresponds to the disappearance of the hot spot. (author)

  7. A phenomenological analysis of melt progression in the lower head of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, J.M., E-mail: jean-marie.seiler@cea.fr [CEA, DEN, DTN, F-38054 Grenoble (France); Tourniaire, B. [EDF/Septen, Lyon (France)

    2014-03-15

    Highlights: • We propose a phenomenological description of melt progression into the lower head. • We examine changes in heat loads on the vessel. • Heat loads are more severe than emphasized by the bounding situation assumption. • Both primary circuit and ex-vessel reflooding are necessary for in-vessel retention. • Vessel failure conditions are examined. - Abstract: The analysis of in-vessel corium cooling (IVC) and retention (IVR) involves the description of very complex and transient physical phenomena. To get round this difficulty, “bounding” situations are often emphasized for the demonstration of corium coolability, by vessel flooding and/or by reactor pit flooding. This approach however comes up against its own limitations. More realistic melt progression scenarios are required to provide plausible corium configurations and vessel failure conditions. Work to develop more realistic melt progression scenarios has been done at CEA, in collaboration with EDF. Development has concentrated on the French 1300 MWe PWR, considering both dry scenarios and the possibility of flooding of the RPC (reactor primary circuit) and/or the reactor pit. The models used for this approach have been derived from the analysis of the TMI2 accident and take benefit from the lessons derived from several programs related to pool thermal hydraulics (BALI, COPO, ACOPO, etc.), material interactions (RASPLAV, MASCA), critical heat flux (CHF) on the external surface of the vessel (KAIST, SULTAN, ULPU), etc. Important conclusions of this work are as follows: (a)After the start of corium melting and onset of melt formation in the core at low pressure (∼1 to 5 bars), it seems questionable that RPV (reactor pressure vessel) reflooding alone would be sufficient to achieve corium retention in the vessel; (b)If the vessel is not cooled externally, it may fail due to local heat-up before the whole core fuel inventory is relocated in the lower head; (c)Even if the vessel is

  8. Modelling of multicomponent diffusion in a two-phase oxide-metal corium pool by a diffuse interface method

    International Nuclear Information System (INIS)

    Cardon, Clement

    2016-01-01

    This Ph.D. topic is focused on the modelling of stratification kinetics for an oxide-metal corium pool (U-O-Zr-steel system) in terms of multicomponent and multiphase diffusion. This work is part of a larger research effort for the development of a detailed corium pool modelling based on a CFD approach for thermal hydraulics. The overall goal is to improve the understanding of the involved phenomena and obtain closure laws for integral macroscopic models. The phase-field method coupled with an energy functional using the CALPHAD method appears to be relevant for this purpose. In a first part, we have developed a diffuse interface model in order to describe the diffusion process in the U-O system. This model has been coupled with a CALPHAD thermodynamic database and its parameterization has been developed with, in particular, an up-scaling procedure related to the interface thickness. Then, within the framework of a modelling for the U-O-Zr ternary system, we have proposed a generalization of the diffuse interface model through an assumption of local equilibrium for redox mechanisms. A particular attention was paid to the model analysis by 1D numerical simulations with a special focus on the steady state composition profiles. Finally we have applied this model to the U-O-Zr-Fe system. For that purpose, we have considered a configuration close to small-scale experimental tests of oxide-metal corium pool stratification. (author) [fr

  9. Integral experiments on in-vessel coolability and vessel creep: results and analysis of the FOREVER-C1 test

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A. [Division of Nuclear Power Safety, Royal Institute of Technology, Drottning Kristinas Vaeg., Stockholm (Sweden)

    1999-07-01

    This paper describes the FOREVER (Failure Of REactor VEssel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The paper presents the experimental results and analysis of the first FOREVER-C1 test. During this experiment, the 1/10th scale pressure vessel, heated to about 900degC and pressurized to 26 bars, was subjected to creep deformation in a non-stop 24-hours test. The vessel wall displacement data clearly shows different stages of the vessel deformation due to thermal expansion, elastic, plastic and creep processes. The maximum displacement was observed at the lowermost region of the vessel lower plenum. Information on the FOREVER-C1 measured thermal characteristics and analysis of the observed thermal and structural behavior is presented. The coupled nature of thermal and mechanical processes, as well as the effect of other system conditions (such as depressurization) on the melt pool and vessel temperature responses are analyzed. (author)

  10. Design feasibility study on corium stabilization in bottom end-fitting for AHWR under accident condition

    International Nuclear Information System (INIS)

    Gokhale, Onkar; Mukhopadhyay, D.; Chatterjee, B.; Singh, R.K.

    2015-01-01

    Advanced Heavy Water Reactor (AHWR) is being designed in a robust way to cater both Design and Beyond Design Basis Accidents to meet all the safety functions. All the functions are met by passive means with special emphasis on 'residual heat removal' which is catered by passive natural circulation mode. In context to Design Basis Accidents, several features are designed to handle worst kind of scenario like Station Black Out. For Design Extension Conditions (DEC), the means of passive natural circulation is adopted as a design means to meet the DEC-A conditions like cooling of moderator by natural circulation means with GDWP inventory. Under the DEC-B condition where large scale of fuel melting is envisaged, a core catcher is designed with active/passive cooling modes to take care of the residual heat of the core. All the mentioned features utilizes the natural mode of heat transfer to meet one of the safety function i.e. 'residual heat removal'. The analysis shows that the tube sheet as well as lattice tube temperatures remain low and are able to take out the heat from corium through sub-cooled nucleate boiling. The ES cooling is sufficient to maintain the cooling water in subcooled condition. The integrity of tube sheet and lattice tube is maintained

  11. Experimental study of the fragmentation and quench behavior of corium melts in water

    International Nuclear Information System (INIS)

    Wang, S.K.; Blomquist, C.A.; Spencer, B.W.; McUmber, L.M.; Schneider, J.P.; Illinois Univ., Urbana, IL

    1989-01-01

    The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a crucial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues which arise during the molten core relocation include: (1) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (2) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (3) the quench rate of the molten fuel through the water in the lower plasma, (4) the steam generation and hydrogen generation during the interaction, (5) the transient pressurization of the primary system, and (6) the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through -6) was performed in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam/hydrogen generation, and the posttest debris characteristics. 9 refs., 29 figs

  12. Analysis of a thermite experiment to study low pressure corium dispersion

    International Nuclear Information System (INIS)

    Wilhelm, D.

    2001-08-01

    The report describes the recalculation of a thermite experiment in a reduced scale which simulates the discharge of molten core materials out of the pressure vessel of a light water reactor into the open compartments and the dome of the containment. The experiment was performed in the framework of a multinational effort at the Sandia National Laboratory, U.S.A. It is being followed by the DISCO program at the Forschungszentrum Karlsruhe. A computational fluid dynamics code was supplemented with specific models to recalculate the Sandia experiment in order to identify problem areas which need to be addressed in the future. Therefore, a first attempt was undertaken to extrapolate to reactor conditions. This was done in two steps to separate geometric from material scaling relationships. The study shows that important experimental results can be extrapolated according to general scaling laws but that there are sensitivities, especially when replacing thermite by corium. The results show a considerable scatter and a dependence on geometric resolution and dynamics of energy transfer between participating components. (orig.) [de

  13. Coupled thermo-mechanical analysis of corium-loaded lower head of pressure vessel

    International Nuclear Information System (INIS)

    Mishra, J.; Balasubramaniyan, V.

    2016-01-01

    A severe accident in the pressurised water reactor may lead to the relocation of core materials to the lower head of Reactor Pressure Vessel (RPV). The core debris at the bottom of RPV forms a melt pool of corium due to decay heat. The understanding of behaviour of pressure vessel, characterised by failure mode and time to failure, in this scenario is one of the important steps in predicting the accident progression. The most predominant failure mode is multi-axial creep deformation of the vessel with a non-uniform temperature field. Towards this, a numerical analysis methodology is developed for the prediction of pressure vessel deformation during the severe accidents. The methodology involves 2-D finite element modelling under multi-physics environment, which account the creep phenomena using Norton-Bailey creep law with a typical damage model of RPV material. The validation of the methodology is carried out using the results from OLHF experiment carried out in Sandia National Laboratory (SNL), USA, within the framework of an OECD. (author)

  14. An analysis of molten-corium-induced failure of drain pipes in BWR Mark 2 containments

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.; Podowski, M.Z.

    1991-01-01

    This study has focused on mechanistic simulation and analysis of potential failure modes for inpedestal drywell drain pipes in the Limerick boiling water reactor (BWR) Mark 2 containment. Physical phenomena related to surface tension breakdown, heatup, melting, ablation, crust formation and failure, and core material relocation into drain pipes with simultaneous melting of pipe walls were modeled and analyzed. The results of analysis have been used to assess the possibility of drain pipe failure and the resultant loss of pressure-suppression capability. Estimates have been made for the timing and amount of molten corium released to the wetwell. The study has revealed that significantly different melt progression sequences can result depending upon the failure characteristics of the frozen metallic crust which forms over the drain cover during the initial stages of debris pour. Another important result is that it can take several days for the molten fuel to ablate the frozen metallic debris layer -- if the frozen layer has cooled below 1100 K before fuel attack. 10 refs., 3 figs., 4 tabs

  15. Coolability of the melt in the reactor tank. A compilation and evaluation of the state of the art and suggestions for experiments in the area; Smaeltans kylbarhet i reaktortanken. En sammanstaellning och vaerdering av kunskapslaeget och foerslag till experiment inom omraadet

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Ferenc [ES-konsult Energi och Saekerhet AB, Stockholm (Sweden)

    2002-04-01

    This study is limited to experiments about phenomena and mechanisms that affect the coolability of core debris in the reactor tank that may delay the tank rupture. The goal of the study is to get an overview of the phenomena that are important for the in-vessel coolability, and to evaluate the need for new experiments. Both theoretical and experimental projects are suggested.

  16. Numerical models for the analysis of thermal behavior and coolability of a particulate debris bed in reactor lower head

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Sang Baik; Kim, Byung Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This report provides three distinctive, but closely related numerical models developed for the analysis of thermal behavior and coolability of a particulate debris bed that is may be formed inside the reactor lower head during severe accident late phases. The first numerical module presented in the report, MELTPRO-DRY, is used to analyze numerically heat-up and melting process of the dry particle bed, downward- and sideward-relocation of the liquid melt under gravity force and capillary force acting among porous particles, and solidification of the liquid melt relocated into colder region. The second module, MELTPROG-WET, is used to simulate numerically the cooling process of the particulate debris bed under the existence of water, which is subjected to two types of numerical models. The first type of WET module utilizes distinctive models that parametrically simulate the water cooling process, that is, quenching region, dryout region, and transition region. The choice of each parametric model depends on temperature gradient between the cooling water and the debris particles. The second type of WET module utilizes two-phase flow model that mechanically simulates the cooling process of the debris bed. For a consistent simulation from the water cooling to the dryout debris bed, on the other hand, the aforementioned two modules, MELTPROG-DRY and MELTPROG-WET, were integrated into a single computer program DBCOOL. Each of computational models was verified through limited applications to a heat-generating particulate bed contained in the rectangular cavity. 22 refs., 5 figs., 2 tabs. (Author)

  17. Experimental investigations on the coolability of prototypical particle beds with respect to reactor safety; Experimentelle Untersuchungen der Kuehlbarkeit prototypischer Schuettungskonfigurationen unter dem Aspekt der Reaktorsicherheit

    Energy Technology Data Exchange (ETDEWEB)

    Leininger, Simon

    2017-02-22

    In case of a severe accident in a light water reactor, continuous unavailability of cooling water to the reactor core may result in overheating of the fuel elements and finally the loss of core integrity. Under such conditions, a structure of heat-releasing particles of different size and shape may be formed by fragmentation of molten core material in several stages of the accident. The long-term coolability of such beds is of prime im-portance to avoid any damage to the reactor pressure vessel or even a release of fission products to the environment. In the frame of this work, specific experiments were con-ducted under prototypical conditions employing the existing DEBRIS test facility in order to gain further knowledge about the thermohydraulic behavior of such beds. In steady state boiling experiments, the pressure gradients in particle beds were meas-ured both for one- and multi-dimensional cooling water flow conditions and compared with one another in order to assess the flow behavior inside the bed. For these different flow conditions as well as for stratified bed configurations, the maximum removable heat flux densities were determined in the dryout experiments. E. g., it was found that an axial stratification of the permeability can significantly reduce the bed's coolability. For the first time, the quenching behavior of dry, superheated beds was investigated at elevated system pressure up to 0.5 MPa. In these experiments, the effect of system pressure on the coolability was quantified by means of the quenching time (time period to cool down the bed to saturation temperature). The investigated particle beds mainly consisted of non-spherical particles with well-defined geometry (cylinders and screws). It was shown that the effect of the particles geometry on the flow in a particle bed can be best estimated by using an equivalent particle diameter calculated for monodisperse particle beds from the product of the Sauter diameter and a shape factor and for

  18. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    International Nuclear Information System (INIS)

    Phung, Viet-Anh; Galushin, Sergey; Raub, Sebastian; Goronovski, Andrei; Villanueva, Walter; Kööp, Kaspar; Grishchenko, Dmitry; Kudinov, Pavel

    2016-01-01

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small ( 100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input

  19. Characteristics of debris in the lower head of a BWR in different severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2016-08-15

    Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

  20. Thermophysical properties of liquid UO{sub 2}, ZrO{sub 2} and corium by molecular dynamics and predictive models

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Kee; Shim, Ji Hoon [Pohang University of Science and Technology, Pohang (Korea, Republic of); Kaviany Massoud [University of Michigan, Ann Arbor (United States)

    2016-10-15

    The analysis of such accidents (fate of the melt), requires accurate corium thermophysical properties data up to 5000 K. In addition, the initial corium melt superheat melt, determined from such properties, are key in predicting the fuel-coolant interactions (FCIs) and convection and retention of corium in accident scenarios, e.g., core-melt down corium discharge from reactor pressure vessels and spreading in external core-catcher. Due to the high temperatures, data on molten corium and its constituents are limited, so there are much data scatters and mostly extrapolations (even from solid state) have been used. Here we predict the thermophysical properties of molten UO{sub 2} and ZrO{sub 2} using classical molecular dynamics (MD) simulations (properties of corium are predicted using the mixture theories and UO{sub 2} and ZrO{sub 2} properties). The thermophysical properties (density, compressibility, heat capacity, viscosity and surface tension) of liquid UO{sub 2} and ZrO{sub 2} are predicted using classical molecular dynamics simulations, up to 5000 K. For atomic interactions, the CRG and the Teter potential models are found most appropriate. The liquid behavior is verified with the random motion of the constituent atoms and the pair-distribution functions, starting with the solid phase and raising the temperature to realize liquid phase. The viscosity and thermal conductivity are calculated with the Green-Kubo autocorrelation decay formulae and compared with the predictive models of Andrade and Bridgman. For liquid UO{sub 2}, the CRG model gives satisfactory MD predictions. For ZrO{sub 2}, the density is reliably predicted with the CRG potential model, while the compressibility and viscosity are more accurately predicted by the Teter model.

  1. Experimental investigation of 150-KG-scale corium melt jet quenching in water

    Energy Technology Data Exchange (ETDEWEB)

    Magallon, D.; Hohmann, H.

    1995-09-01

    This paper compares and discusses the results of two large scale FARO quenching tests known as L-11 and L-14, which involved, respectively, 151 kg of W% 76.7 UO{sub 2} + 19.2 ZrO{sub 2} + 4.1 Zr and 125 kg of W% 80 UO{sub 2} + 20 ZrO{sub 2} melts poured into 600-kg, 2-m-depth water at saturation at 5.0 MPa. The results are further compared with those of two previous tests performed using a pure oxidic melt, respectively 18 and 44 kg of W% 80 UO{sub 2} + 20 ZrO{sub 2} melt quenched in 1-m-depth water at saturation at 5.0 MPa. In all the tests, significant breakup and quenching took place during the melt fall through the water. No steam explosion occurred. In the tests performed with a pure oxide UO{sub 2}-ZrO{sub 2} melt, part of the corium (from 1/6 to 1/3) did not breakup and reached the bottom plate still molten whatever the water depth was. Test L-11 data suggest that full oxidation and complete breakup of the melt occurred during the melt fall through the water. A proportion of 64% of the total energy content of the melt was released to the water during this phase ({approximately}1.5 s), against 44% for L-14. The maximum temperature increase of the bottom plate was 330 K (L-14). The mean particle size of the debris ranged between 2.5 and 4.8mm.

  2. Validation of ASTEC v2.0 corium jet fragmentation model using FARO experiments

    International Nuclear Information System (INIS)

    Hermsmeyer, S.; Pla, P.; Sangiorgi, M.

    2015-01-01

    Highlights: • Model validation base extended to six FARO experiments. • Focus on the calculation of the fragmented particle diameter. • Capability and limits of the ASTEC fragmentation model. • Sensitivity analysis of model outputs. - Abstract: ASTEC is an integral code for the prediction of Severe Accidents in Nuclear Power Plants. As such, it needs to cover all physical processes that could occur during accident progression, yet keeping its models simple enough for the ensemble to stay manageable and produce results within an acceptable time. The present paper is concerned with the validation of the Corium jet fragmentation model of ASTEC v2.0 rev3 by means of a selection of six experiments carried out within the FARO facility. The different conditions applied within these six experiments help to analyse the model behaviour in different situations and to expose model limits. In addition to comparing model outputs with experimental measurements, sensitivity analyses are applied to investigate the model. Results of the paper are (i) validation runs, accompanied by an identification of situations where the implemented fragmentation model does not match the experiments well, and discussion of results; (ii) its special attention to the models calculating the diameter of fragmented particles, the identification of a fault in one model implemented, and the discussion of simplification and ad hoc modification to improve the model fit; and, (iii) an investigation of the sensitivity of predictions towards inputs and parameters. In this way, the paper offers a thorough investigation of the merit and limitation of the fragmentation model used in ASTEC

  3. Experimental investigation on molten pool representing corium composition at Fukushima Daiichi nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang Mo, E-mail: sangmoan@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yueong-gu, Daejeon, 305-353 (Korea, Republic of); Song, Jin Ho [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yueong-gu, Daejeon, 305-353 (Korea, Republic of); Kim, Jong-Yun [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yueong-gu, Daejeon, 305-353 (Korea, Republic of); Radiochemistry & Nuclear Nonproliferation, University of Science & Technology, Gajeong-ro 217, Yuseong-gu, Daejeon, 34113 (Korea, Republic of); Kim, HwanYeol [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yueong-gu, Daejeon, 305-353 (Korea, Republic of); Naitoh, Masanori [The Institute of Applied Energy, 1-14-2 Nishi-shimbashi, 1-Chome, Minato-ku, Tokyo, 105-0003 (Japan)

    2016-09-15

    A configuration of molten core in the Fukushima Daiichi NPP (nuclear power plant) was investigated by a melting and solidification experiment. About 5 kg of a mixture, whose composition in terms of weight is UO{sub 2} (60%), Zr + ZrO{sub 2} (25%), stainless steel (14%), B{sub 4}C (1%), was melted in a cold crucible using an induction heating technique. It was shown that the solidified melt consists of upper crust and lower solidified ingot. The solidified ingot was separated into two layers. A physical and chemical analysis was performed for the samples taken from the solidified melt to investigate the morphology and chemical characteristics. It was found that the solidified ingot consists of a metal-rich layer on the top and an oxide-rich layer at the bottom. In addition, the oxide layer at the bottom has composition close to the initial charge composition and surrounded by a thin crust layer. It turned out that B{sub 4}C was more concentrated in the upper metal-rich layer. These findings provide important insights for understanding the core melt progression and taking proper post-accident recovery actions for the Fukushima Daiichi NPP. - Highlights: • A configuration of molten core in the Fukushima Daiich NPP unit 1 is investigated. • Corium ingot consists of metallic layer on the top and oxidic layer at the bottom. • Boron carbide was more concentrated in the upper metallic layer. • Two layered configuration would contribute to the post-accident recovery actions.

  4. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2001-03-15

    In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

  5. A Heat Transfer Model for a Stratified Corium-Metal Pool in the Lower Plenum of a Nuclear Reactor

    International Nuclear Information System (INIS)

    Sohal, M.S.; Siefken, L.J.

    1999-01-01

    This preliminary design report describes a model for heat transfer in a corium-metal stratified pool. It was decided to make use of the existing COUPLE model. Currently available correlations for natural convection heat transfer in a pool with and without internal heat generation were obtained. The appropriate correlations will be incorporated in the existing COUPLE model. Heat conduction and solidification modeling will be done with existing algorithms in the COUPLE. Assessment of the new model will be done by simple energy conservation problems

  6. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    International Nuclear Information System (INIS)

    Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin; Park, Jae Hong

    2001-03-01

    In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report

  7. Recent severe accident research synthesis of the major outcomes from the SARNET network

    Energy Technology Data Exchange (ETDEWEB)

    Van Dorsselaere, J.-P., E-mail: jean-pierre.van-dorsselaere@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Saint-Paul-lez-Durance (France); Auvinen, A. [VTT Technical Research Centre, Espoo (Finland); Beraha, D. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), Köln (Germany); Chatelard, P. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Saint-Paul-lez-Durance (France); Herranz, L.E. [Centro de Investigaciones Energéticas MedioAmbientales y Tecnológicas (CIEMAT), Madrid (Spain); Journeau, C. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Paris (France); Klein-Hessling, W. [Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS), Köln (Germany); Kljenak, I. [Jozef Stefan Institute (JSI), Ljubljana (Slovenia); Miassoedov, A. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Paci, S. [University of Pisa, Pisa (Italy); Zeyen, R. [European Commission Joint Research Centre, Institute for Energy (JRC/IET), Petten (Netherlands)

    2015-09-15

    Highlights: • SARNET network of excellence integration mid-2013 in the NUGENIA Association. • Progress of knowledge on corium behaviour, hydrogen explosion and source term. • Further development of ASTEC integral code to capitalize knowledge. • Ranking of next R&D high priority issues accounting for international research. • Dissemination of knowledge through education courses and ERMSAR conferences. - Abstract: The SARNET network (Severe Accident Research NETwork of excellence), co-funded by the European Commission from 2004 to 2013, has allowed to significantly improve the knowledge on severe accidents and to disseminate it through courses and ERMSAR conferences. The major investigated topics, involving more than 250 researchers from 22 countries, were in- and ex-vessel corium/debris coolability, molten-core–concrete-interaction, steam explosion, hydrogen combustion and mitigation in containment, impact of oxidising conditions on source term, and iodine chemistry. The ranking of the high priority issues was updated to account for the results of recent international research and for the impact of Fukushima nuclear accidents in Japan. In addition, the ASTEC integral code was further developed to capitalize the new knowledge. The network has reached self-sustainability by integration in mid-2013 into the NUGENIA Association. The main activities and outcomes of the network are presented.

  8. Recent severe accident research synthesis of the major outcomes from the SARNET network

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.-P.; Auvinen, A.; Beraha, D.; Chatelard, P.; Herranz, L.E.; Journeau, C.; Klein-Hessling, W.; Kljenak, I.; Miassoedov, A.; Paci, S.; Zeyen, R.

    2015-01-01

    Highlights: • SARNET network of excellence integration mid-2013 in the NUGENIA Association. • Progress of knowledge on corium behaviour, hydrogen explosion and source term. • Further development of ASTEC integral code to capitalize knowledge. • Ranking of next R&D high priority issues accounting for international research. • Dissemination of knowledge through education courses and ERMSAR conferences. - Abstract: The SARNET network (Severe Accident Research NETwork of excellence), co-funded by the European Commission from 2004 to 2013, has allowed to significantly improve the knowledge on severe accidents and to disseminate it through courses and ERMSAR conferences. The major investigated topics, involving more than 250 researchers from 22 countries, were in- and ex-vessel corium/debris coolability, molten-core–concrete-interaction, steam explosion, hydrogen combustion and mitigation in containment, impact of oxidising conditions on source term, and iodine chemistry. The ranking of the high priority issues was updated to account for the results of recent international research and for the impact of Fukushima nuclear accidents in Japan. In addition, the ASTEC integral code was further developed to capitalize the new knowledge. The network has reached self-sustainability by integration in mid-2013 into the NUGENIA Association. The main activities and outcomes of the network are presented

  9. Code development for debris bed coolability problem. Final report for the period 1997-05-01 - 1999-08-14

    International Nuclear Information System (INIS)

    Loboiko, A.I.

    2000-03-01

    The study was devoted to the problem of debris bed coolability arising from severe accident at nuclear power reactor. After reactor core melting occurs and subsequent debris bed is formed in the lower plenum of reactor pressure vessel (RPV) it is important to confine this debris bed inside RPV boundary. One of the possible accident scenarios assumes the interaction between coolant and molten core materials resulting from rapid melt quenching, freezing and fragmentation. Particulated fuel and steel may subsequently settle on available surfaces within the reactor vessel, forming debris porous beds which produce radioactive decay heating. In case of severe core degradation, such heat transfer mechanisms as radiation, conduction and natural single-phase convection may appear to be insufficient and coolant boiling may happen on the surface or inside the bed. Depending on rate of heat generation there may be sufficient debris cool down or its 'dryout' which pose a danger for RPV integrity. The study considers development of 2D numerical code capable to predict coolant saturation as a function of different parameters. Analysis of previous activities on one-dimensional and multi-dimensional models was done. On the basis of the analysis it was concluded that the correct prediction of the debris saturation on dryant power requires two-dimensional numerical simulation considering the processes like two-phase convection, capillary effects, different models of permeability, different models of heat transfer between solid debris and coolant, non-homogeneity of parameters porous medium, heat and mass transfer between debris bed and a highly porous gap along the inner RPV surface. Particular attention was given to consideration of boundary conditions for debris bed. Introduction of the analytical model for dependence of gap properties on heat flux from debris bed allowed to create an algorithm for use in numerical calculations and finally to develop a code which allowed for stable

  10. Synthetic analyses of the LAVA experimental results on in-vessel corium retention through gap cooling

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Cho, Young Ro; Koo, Kil Mo; Park, Rae Joon; Kim, Jong Hwan; Kim, Jong Tae; Ha, Kwang Sun; Kim, Sang Baik; Kim, Hee Dong

    2001-03-01

    LAVA(Lower-plenum Arrested Vessel Attack) has been performed to gather proof of gap formation between the debris and lower head vessel and to evaluate the effect of the gap formation on in-vessel cooling. Through the total of 12 tests, the analyses on the melt relocation process, gap formation and the thermal and mechanical behaviors of the vessel were performed. The thermal behaviors of the lower head vessel were affected by the formation of the fragmented particles and melt pool during the melt relocation process depending on mass and composition of melt and subcooling and depth of water. During the melt relocation process 10.0 to 20.0 % of the melt mass was fragmented and also 15.5 to 47.5 % of the thermal energy of the melt was transferred to water. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation between the debris and the lower head vessel in case there was an internal pressure load across the vessel abreast with the thermal load induced by the thermite melt. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were determined by the possibilities of the water ingression into the gap depending on the melt composition of the corium simulant. The enhanced cooling capacity through the gap was distinguished in the Al 2 O 3 melt tests. It could be inferred from the analyses on the heat removal capacity through the gap that the lower head vessel could effectively cooldown via heat removal in the gap governed by counter current flow limits(CCFL) even if 2mm thick gap should form in the 30 kg Al 2 O 3 melt tests, which was also confirmed through the variations of the conduction heat flux in the vessel and rapid cool down of the vessel outer surface in the Al 2 O 3 melt tests. In the case of large melt mass of 70 kg Al 2 O 3 melt, however, the infinite possibility of heat removal through the

  11. Interaction between molten corium UO{sub 2+x}-ZrO{sub 2}-FeO{sub y} and VVER vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S. V.; Granovsky, V. S.; Khabensky, V. B.; Krushinov, E. V.; Vitol, S. A.; Sulatsky, A. A. [Alexandrov Sci Res Technol Inst, Sosnovyi Bor (Russian Federation); Gusarov, V. V.; Almiashev, V. I. [Russian Acad Sci, Inst Silicate Chem, St Petersburg (Russian Federation); Lopukh, D. B. [SPb State Electrotech Univ LETI SPbGETU, St Petersburg (Russian Federation); Bottomley, D. [Joint Res Ctr, Inst Transurane, Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA Saclay, DEN, DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Dresden Rossendorf, Dresden (Germany); Fichot, F. [CEA Cadarache, SEMCA, DPAM, IRSN, St Paul Les Durance (France); Kymalainen, O. [FORTUM Nucl Serv Ltd, Espoo (Finland)

    2010-07-01

    In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO{sub 2+x}-ZrO{sub 2}-FeO{sub y} melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction Zone. The available experimental data have been used to develop a correlation for the corrosion rate as a function of temperature and heat flux. (authors)

  12. Ex-vessel remote maintenance for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) requires the use of remote handling technology to carry out maintenance operations on the machine. These operations consist in removing and repairing such components as diagnostic modules by using remotely operated maintenance equipment. The major equipment being developed for maintenance external to the plasma chamber includes a bridge-mounted manipulator system for test cell operations, decontamination (decon) equipment, hot cell equipment, and solid-radiation-waste-handling equipment. Wherever possible, the project will use commercially available equipment. Several areas of the maintenance system design were addressed in fiscal year (FY) 1987, including conceptual designs of manipulator systems, the start of a remote equipment research and development (RandD) program, and definition of the hot cell, decon, and equipment repair facility requirements. R and D work included preliminary demonstrations of remote handling operations on full-size, partial mock-ups of the CIT machine at the Oak Ridge National Laboratory (ORNL) Remote Operations and Maintenance Development (ROMD) Facility. 1 ref., 6 figs

  13. High temperature zirconia binders for ex-vessel catcher brickwork

    International Nuclear Information System (INIS)

    Mineev, V.N.; Borovkova, L.B.; Akopov, F.A.; Akopyan, A.A.; Barykin, B.M.; Borodina, T.I.; Val'yano, G.E.; Bel'maz, N.S.; Bel'maz, K.N.

    2000-01-01

    The studies on selection of compositions of binding materials (mortars) on the zirconium dioxide basis with two types of binders - the barium monoaluminate and zirconium dioxide binding suspension - are accomplished. The bases of technology for their fabrication and application are developed; the density, porosity, shrinkage and behavior in contact with the steel melts and iron oxide are specified. The mortars developed are recommended for application in the external trap fireproof protection on the basis of the zirconium dioxide refractory materials [ru

  14. Design of ex-vessel neutron monitor for ITER

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Yamauchi, Michinori; Kasai, Satoshi; Ebisawa, Katsuyuki; Walker, Chris

    2002-07-01

    A neutron flux monitor has been designed by using 235 U fission chambers to be installed outside the vacuum vessel of ITER. We investigated moderator materials to get flat energy response the responses of 235 U fission chambers. Here we employed graphite and beryllium with a ratio of Be/C=0.25 as moderator, which materials are stable in ITER relevant temperature in a horizontal port. Based on the neutronics calculations, a fission chamber with 200 mg of 235 U is adopted for the neutron flux monitor. Three detectors are mounted in a stainless steel housing with moderation material. Two fission chamber assemblies will be installed in a horizontal port; one is for D-D and calibration operation, and another is for D-T operation. The assembly for the D-D operation and the calibration are installed just outside the port plug in the horizontal port. The assembly for the D-T operation is installed just behind the additional shield in the port. Combining of those assemblies with both pulse counting mode and Campbelling mode in the electronics, a dynamic range of 10 7 can be obtained with 1 ms temporal resolution. Effects of gamma-rays and magnetic fields on the fission chamber are negligible in this arrangement. The neutron flux monitor can meet the required 10% accuracy for a fusion power monitor. (author)

  15. The new Ex-Vessel Magnetic Diagnostics System for JET

    International Nuclear Information System (INIS)

    Coccorese, V.; Artaserse, G.; Quercia, A.; Chitarin, G.; Peruzzo, S.; Edlington, T.; Gerasimov, S.; Sowden, C.

    2006-01-01

    A new system of magnetic probes was installed during the 2005 shutdown and was commissioned during the 2005/06 restart phase of JET. The system has been developed in the framework of the JET enhancement project on Magnetic Diagnostics, which aims to improve the equilibrium reconstruction and the real time control in JET, by means of 98 new field measurements as well as of new software tools. The subsystem presented in the paper includes probes located outside the vessel and it is made of 8 pickup coils, 8 Hall probes and 6 flux loops. The objective of this subsystem is twofold: i) provide experimental data for a better modelling of the iron in the axisymmetric codes for plasma equilibrium reconstruction; ii) test the reliability of direct field measurements. The new sensors are located very near to the iron structure, so to provide useful information for the online tuning of the code parameters representing the iron characteristics. Direct field measurements from Hall probes are used to correct the drift of the integrators of the pickup coils signals. This feature will be crucial for future ITER-like devices, where long lasting flat top phases are expected, in a high neutron yield and a high temperature environment. After a general overview of the system, the paper describes the major manufacturing and installation issues, including the construction of the supports and probes as well as the acceptance tests before and after installation. The functional commissioning of the system, which was successfully concluded during the restart phase, is also illustrated. It includes the integration of the new signals in the JET CODAS system and the analysis of several discharges with and without plasma. The critical aspects of the assessment of the reliability of the signals are shown and commented on. (author)

  16. Analysis of coolability of the control rods of a Savannah River Site production reactor with loss of normal forced convection cooling

    International Nuclear Information System (INIS)

    Easterling, T.C.; Hightower, N.T.; Smith, D.C.; Amos, C.N.

    1992-01-01

    An analytical study of the coolability of the control rods in the Savannah River Site (SRS) K-Production Reactor under conditions of loss of normal forced convection cooling has been performed. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven flow over the control rods that occurs when forced cooling is lost, and the limit of critical heat flux that sets the acceptance criteria for the study. The objective of the study is to demonstrate that the control rods will remain cooled at powers representative of those anticipated for restart of the reactor. The study accomplishes this objective with a very tractable simplified analysis for the modest restart power. In addition, a best-estimate calculation is performed, and the results are compared to results from sub-scale scoping experiments. 5 refs

  17. Modeling of spreading of the melted corium jet inside the pool of emergency heat removal during severe accidents at NPP

    Directory of Open Access Journals (Sweden)

    I. V. Kazachkov

    2012-03-01

    Full Text Available Important nuclear power safety problem in touch with modeling of melted corium jet spreading inside the coolant pool is considered in the paper. It appears by development of the passive protection systems against se-vere accidents. The non-linear mathematical developed model is presented for the jet under reactor vessel pool for one of the perspective passive protection systems and the results of its analysis and studies are given. The performed analysis and the results of the numerical simulation done on the base of the model have allowed estab-lishing the interesting behaviors of the system, which may be useful for the scientists, as well as the engineers-constructors of the passive protection systems against severe accidents.

  18. Analysis of materials in connection with corium melt retention in WWER reactor vessels. Final report for the period 15 October 1995 - 14 October 1996

    International Nuclear Information System (INIS)

    Efanov, A.D.

    1997-02-01

    Analysis of the state of severe accident codes being developed in Russia describing processes of corium - reactor pressure vessel interaction during severe accidents showed that at present there is no reliable validated and verified code. This study considered some of the most advanced severe accident codes which include models of heat generating liquid convections and RPV tolerance capacity however the possibility of physico-chemical interactions is not being considered. The final report demonstrates one of the examples for the settlement modelling of processes of reactor core debris cooling and corium pool mirror cooling with the use of the KOSTER 2 Code developed at IPPE. It was shown that cooling by water spray within some definite region of drop dimensions (0.5 / 4 mm in diameter) could provide the opportunity of the WWER type RPV integrity preservation under the accepted conditions. Due to some uncertainties in calculation modules obtained results are recommended to be treated as preliminary. (author). 26 refs, figs, tabs

  19. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool

    International Nuclear Information System (INIS)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L.

    2007-01-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  20. Analysis of top flooding during molten corium concrete interaction (MCCI) with the code MEDICIS using a simplified approach for the combined effect of crust formation and boiling

    International Nuclear Information System (INIS)

    Spengler, C.

    2012-01-01

    The objective of this work is to provide adequate models in the code MEDICIS for the molten corium concrete interaction (MCCI) phase in a severe accident. Here, the multidimensional distribution of heat fluxes from the molten pool of corium to the sidewall and bottom wall concrete structures in the reactor pit and to the top surface is a persistent subject of international research activities on MCCI. In recent experi-ments with internally heated oxide melts it was observed that the erosion progress may be anisotropic - with an apparent preference of the sidewall compared to the bottom wall - or isotropic, in dependence of the type of concrete with which the cori-um interacts. The lumped parameter code MEDICIS, which is part of the severe accident codes ASTEC and COCOSYS - developed and used at IRSN/GRS respectively GRS for the latter one -, is dedicated to simulate the phenomenology during MCCI. In this work a simplified modelling in MEDICIS is tested to account for the observed ablation behaviour during MCCI, with focus on the heat transfer to the top surface under flooded conditions. This approach is assessed by calculations for selected MCCI experiments involving the top flooding of the melt. (orig.)

  1. OECD MCCI project 2-D Core Concrete Interaction (CCI) tests : CCI-3 test data report-thermalhydraulic results. Rev. 0 October 15, 2005.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of a third long-term 2-D Core-Concrete Interaction (CCI) experiment designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-3 experiment, which was conducted on September 22, 2005. Test specifications for CCI-3 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 375

  2. OECD MMCI 2-D Core Concrete Interaction (CCI) tests : CCCI-1 test data report-thermalhydraulic results. Rev 0 January 31, 2004.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten coreconcrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-1 experiment, which was conducted on December 19, 2003. Test specifications for CCI-1 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

  3. OECD MCCI 2-D Core Concrete Interaction (CCI) tests : CCI-2 test data report-thermalhydraulic results, Rev. 0 October 15, 2004.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-2 experiment, which was conducted on August 24, 2004. Test specifications for CCI-2 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

  4. Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

    International Nuclear Information System (INIS)

    Ishii, M.; Zhang, G.; No, H. C.; Eltwila, F.

    1994-01-01

    Accident sequences which lead to severe core damage and to possible radioactive fission products into the environment have a very low probability. However, the interest in this area increased significantly due to the occurrence of the small break loss-of-coolant accident at TMI-2 which led to partial core damage, and of the Chernobyl accident in the former USSR which led to extensive core disassembly and significant release of fission products over several countries. In particular, the latter accident raised the international concern over the potential consequences of severe accidents in nuclear reactor systems. One of the significant shortcomings in the analyses of severe accidents is the lack of well-established and reliable scaling criteria for various multiphase flow phenomena. However, the scaling criteria are essential to the severe accident, because the full scale tests are basically impossible to perform. They are required for (1) designing scaled down or simulation experiments, (2) evaluating data and extrapolating the data to prototypic conditions, and (3) developing correctly scaled physical models and correlations. In view of this, a new scaling method is developed for the analysis of severe accidents. Its approach is quite different from the conventional methods. In order to demonstrate its applicability, this new stepwise integral scaling method has been applied to the analysis of the corium dispersion problem in the direct containment heating. ((orig.))

  5. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool; L'essai Plinius/Colima CA-U3 sur le relachement des aerosols de produits de fission au-dessus d'un bain de corium de type VVER

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L

    2007-07-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  6. Effect of UV-B and high visual radiation on photosynthesis in freshwater (nostoc spongiaeforme) and marine (Phormidium corium) cyanobacteria.

    Science.gov (United States)

    Bhandari, Rupali; Sharma, Prabhat Kumar

    2007-08-01

    Human activity is causing depletion of ozone in stratosphere, resulting in increased UV-B radiation and global warming. However, impact of these climatic changes on the aquatic organism (especially marine) is not fully understood. Here, we have studied the effect of excess UV-B and visible radiation on photosynthetic pigments, fatty acids content, lipid peroxidation, nitrogen content, nitrogen reductase activity and membrane proteins, induction of mycosporine-like amino acids (MAAs) and antioxidant enzymes superoxide dismutase (SOD) and ascorbate peroxidase (APX) in freshwater (Nostoc spongiaeform) and marine (Phormidium corium) cyanobacteria. UV-B treatment resulted in an increase in photosynthetic pigments in Nostoc and decrease in Phormidium, but high light treatment caused photobleaching of most of the pigments in both the species. Unsaturation level of fatty acids of both total and glycolipids remained unchanged in both the cyanobacteria, as a result of UV-B and high light treatments. Saturated fatty acids of total and glycolipids declined slightly in Nostoc by both the treatments. but remained unchanged in Phormidium. No changes in the unsaturated lipid content in our study probably suggested adaptation of the organism to the treatments. However, both treatments resulted in peroxidation of membrane lipids, indicating oxidative damage to lipids without any change in the level of unsaturation of fatty acid in the cell membrane. Qualitative and quantitative changes were observed in membrane protein profile due to the treatments. Cyanobacteria were able to synthesize MAAs in response to the UV-B treatment. Both treatments also increased the activities of SOD and APX. In conclusion, the study demonstrated induction of antioxidants such as SOD and APX under visible light treatment and screening pigment (MAAs) under UV-B treatment, which might protect the cyanobacteria from oxidative damage caused by high light and UV-B radiation.

  7. Premixing of corium into water during a Fuel-Coolant Interaction. The models used in the 3 field version of the MC3D code and two examples of validation on Billeau and FARO experiments

    Energy Technology Data Exchange (ETDEWEB)

    Berthoud, G.; Crecy, F. de; Duplat, F.; Meignen, R.; Valette, M. [CEA/Grenoble, DRN/DTP, 17 Avenue des Martyrs, 38054 Grenoble Cedex 9 (France)

    1998-01-01

    This paper presents the <> application of the multiphasic 3D computer code MC3D. This application is devoted to the premixing phase of a Fuel Coolant Interaction (FCI) when large amounts of molten corium flow into water and interact with it. A description of the new features of the model is given (a more complete description of the full model is given in annex). Calculations of Billeau experiments (cold or hot spheres dropped into water) and of a FARO test (<> corium dropped into 5 MPa saturated water) are presented. (author)

  8. Level 2 probabilistic event analyses and quantification

    International Nuclear Information System (INIS)

    Boneham, P.

    2003-01-01

    In this paper an example of quantification of a severe accident phenomenological event is given. The performed analysis for assessment of the probability that the debris released from the reactor vessel was in a coolable configuration in the lower drywell is presented. It is also analysed the assessment of the type of core/concrete attack that would occur. The coolability of the debris ex-vessel evaluation by an event in the Simplified Boiling Water Reactor (SBWR) Containment Event Tree (CET) and a detailed Decomposition Event Tree (DET) developed to aid in the quantification of this CET event are considered. The headings in the DET selected to represent plant physical states (e.g., reactor vessel pressure at the time of vessel failure) and the uncertainties associated with the occurrence of critical physical phenomena (e.g., debris configuration in the lower drywell) considered important to assessing whether the debris was coolable or not coolable ex-vessel are also discussed

  9. Improvements in modelling (by ESCADRE mod1.0) radiative heat losses through gas and aerosols generated by molten corium-concrete interactions

    International Nuclear Information System (INIS)

    Passalacqua, R.

    1996-01-01

    Aerosols generated during the molten core-concrete interaction (MCCI) influence the reactor cavity thermal hydraulics: the cloud of aerosols, located inside the reactor cavity, restrains the upward-directed heat exchange consequently the cool-down of the high-temperature molten corium for a considerable period of time. IPSN is developing a computer code system for source predictions in severe accident scenarios. This code system is named ESCADRE. WECHSL/CALTHER is internal module dealing with MCCI (it is also a stand-alone code): it models the heat transfers involving the superior volume of the cavity. When modelling the upward-directed power distribution by WECHSL/CALTHER, a faster concrete basemat penetration takes place due to the low heat losses of the closed MCCI cavity enclosure. The model, here presented, is going to be validated with data from the AEROSTAT experiment. This experiment, planned at CEA Cadarache, will evaluate the influence of aerosols on the global power distribution in the reactor cavity. Radiative heat losses are important especially for cavity configurations such as those of new plant designs (equipped with a core-catcher) where the upward power losses are promoted by the corium spreading in a flat cavity

  10. OECD MCCI project long-term 2-D molten core concrete interaction test design report, Rev. 0. September 30, 2002

    International Nuclear Information System (INIS)

    Farmer, M.T.; Kilsdonk, D.J.; Lomperski, S.; Aeschliman, R.W.; Basu, S.

    2011-01-01

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following two technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of the first program objective, the Small-Scale Water Ingression and Crust Strength (SSWICS) test series has been initiated to provide fundamental information on the ability of water to ingress into cracks and fissures that form in the debris during quench, thereby augmenting the otherwise conduction-limited heat transfer process. A test plan for Melt Eruption Separate Effects Tests (MESET) has also been developed to provide information on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions. In terms of the second program objective, the project Management Board (MB) has approved startup activities required to carry out

  11. OECD MCCI project Melt Eruption Test (MET) design report, Rev. 2. April 15, 2003

    International Nuclear Information System (INIS)

    Farmer, M.T.; Lomperski, S.; Kilsdonk, D.J.; Aeschlimann, R.W.; Basu, S.

    2011-01-01

    satisfy these PRG recommendations. Specifically, the revised plan focuses on providing data on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions, including a floating crust boundary condition. The overall objective of MET is to determine to what extent core debris is rendered coolable by eruptive-type processes that breach the crust that rests upon the melt. The specific objectives of this test are as follows: (1) Evaluate the augmentation in surface heat flux during periods of melt eruption; (2) Evaluate the melt entrainment coefficient from the heat flux and gas flow rate data for input into models that calculate ex-vessel debris coolability; (3) Characterize the morphology and coolability of debris resulting from eruptive processes that transport melt into overlying water; and (4) Discriminate between periods when eruptions take the form of particle ejections into overlying water, leading to a porous particle bed, and single-phase extrusions, which lead to volcano-type structures.

  12. A fast running method for predicting the efficiency of core melt spreading for application in ASTEC

    International Nuclear Information System (INIS)

    Spengler, C.

    2010-01-01

    The integral Accident Source Term Evaluation Code (ASTEC) is jointly developed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) and the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor, from the initial event until the possible radiological release of fission products out of the containment. In the frame of the new series of ASTEC V2 versions appropriate model extensions to the European Pressurised Water Reactor (EPR) are under development. With view to assessing with ASTEC the proper operation of the ex-vessel melt retention and coolability concept of the EPR with regard to melt spreading an approximation of the area finally covered by the corium and of the distance run by the corium front before freezing is required. A necessary capability of ASTEC is in a first step to identify such boundary cases, for which there is a potential that the melt will freeze before the spreading area is completely filled. This paper presents a fast running method for estimating the final extent of the area covered with melt on which a simplified criterion in ASTEC for detecting such boundary cases will be based. If a boundary case is detected the application of a more-detailed method might be necessary to assess further the consequences for the accident sequence. The major objective here is to provide a reliable method for estimating the final result of the spreading and not to provide highly detailed methods to simulate the dynamics of the transient process. (orig.)

  13. Overview of IRSN R and D on NPP safety, with focus on severe accident

    International Nuclear Information System (INIS)

    Van Dorsselaere, Jean-Pierre

    2015-01-01

    IRSN contributes to the continuous improvement of safety level of Gen.II and III reactors, with the aim to approach for Gen.II the target safety level of Gen.III. This needs to build the necessary knowledge to appreciate margins for safety important systems, structures and components in the frame of plant operation life extension beyond 40 years. Research is a major IRSN mission that is tightly linked to expertise needs: it involves 40% of overall budget and, out of radiation protection and safety of waste disposal, around 280 scientists. IRSN has acquired a huge experience in the last 30 years on severe accidents, both on experimental and theoretical aspects, in particular through management of large international research programmes like the Phébus. FP integral experiments in the last 20 years and the coordination of the SARNET network of excellence that continues now in the frame of the NUGENIA European association. Besides, IRSN is developing, in collaboration with GRS (Germany), the integral system code ASTEC that is considered now as the European reference code due to the continuous capitalization of all the international knowledge. The presentation summarizes the ongoing IRSN research on the different phenomena involved in severe accidents, with more focus in the last years on mitigation devices or measures, i.e. for in-vessel and ex-vessel corium coolability, hydrogen explosion risk and source term. IRSN leads several international projects in Euratom frame (such as CESAM on ASTEC, PASSAM on source term mitigation, and IVMR on in-vessel corium retention) or OECD/NEA/CSNI (such as STEM). Moreover, several national projects on the above issues are ongoing with the French actors in this domain. Collaboration between IRSN and India is very active and efficient on ASTEC code with BARC and AERB, in particular through PHWR model development and assessment, and could be extended in the future to other issues either on severe accidents or on other Topics. (author)

  14. Accident analyses on TMLB' and LOCA for KNGR using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Choi, Y.; Ahn, K.I

    2000-11-01

    Plant specific phenomenological analyses for the Korean Next Generation Reactor, using MELCOR program, are described in this report. The most important two accident sequences, a station blackout and a loss of coolant scenario, are selected. Complete coverage of corium behavior both in-vessel and ex-vessel, and the corresponding containment responses, are analyzed. The in-vessel progression includes the thermal hydraulics in the primary system, core heat up, hydrogen generation, and melt progression up to the reactor vessel breach. The ex-vessel progression describes molten corium - concrete interaction phenomena and the pressure behavior in the containment atmosphere.

  15. Severe accidents and nuclear containment integrity (SANCY). SANCY summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I. [VTT Processes, Espoo (Finland)

    2004-07-01

    SANCY project investigates physical phenomena related to severe nuclear accidents with importance to Finnish nuclear power plants. Currently the major topics are the ex-vessel coolability issues, long-term severe accident management and containment leak tightness and adoption and development of new calculation tools considering also the needs of the future Olkiluoto 3 plant. SANCY employs both experimental and analytical methods. (orig.)

  16. Presentation of the Vulcano installation which uses a plasma transferred arc rotary furnace for corium melting; Utilisation d`un four tournant a arc plasma transfere pour fondre et couler des melanges d`oxydes autour de 2000 C. Presentation du film Vulcano

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G.; Laffont, G.; Jegou, C.; Pierre, J.; Journeau, C.; Sudreau, F.; Roubaud, A. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d`Etudes des Reacteurs

    1998-06-01

    In the case of loss coolant accident, the reactor core could melt and turn into a mixture of uranium oxides, zirconium, iron and steel called corium. A large experimental program has been launched to study corium behaviour, to qualify solutions to stabilize it and to confine it in the reactor containment. The Vulcano installation has been designed to that purpose. It is made up of: i) a plasma transferred arc rotary furnace, ii) a testing surface covered with refractory materials, iii) an induction heating system in order to simulate the residual power of corium, iv) instrumentation devices such as video cameras, thermocouples, infra-red pyrometers and flowmeters, and v) a laboratory to perform chemical analysis of corium samples. The first experimental results show that a mixture of corium and concrete spreads better than expected. It seems that a low initial height of matter can produce a great distance flowing while having a chaotic behaviour. This characteristic suggests that the mixture acts as a Bingham type threshold fluid. (A.C.) 5 refs.

  17. Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

    Energy Technology Data Exchange (ETDEWEB)

    M. Anderson; M. Corradini; K.Y. Bank; R. Bonazza; D. Cho

    2005-04-26

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications.

  18. Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

    International Nuclear Information System (INIS)

    Anderson, M.; Corradini, M.; Bank, K.Y.; Bonazza, R.; Cho, D.

    2005-01-01

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications

  19. Ex-vessel remote maintenance design for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Macdonald, D.

    1987-01-01

    The use of deuterium-tritium (D-T) fuel for operation of the Compact Ignition Tokamak (CIT) imposes a requirement for remote handling technology to carry out maintenance operations on auxiliary machine components. These operations consist of removing and repairing components such as diagnostics and radio frequency (rf) heating modules using remotely operated maintenance equipment. The major equipment that is being developed to accomplish maintenance external to the plasma chamber includes the bridge-mounted manipulator system for test cell operations, decontamination (decon) equipment, hot cell equipment, and solid rad-waste handling equipment. Wherever possible, the project will use commercially available equipment. Several areas of the maintenance system design have been addressed in fiscal year (FY) 1987. These included conceptual designs of manipulator systems, the start of a remote equipment research and development (R and D) program, and definition of the hot cell, decon, and equipment repair facility requirements. The manipulator work included investigating transporters and viewing/lighting subsystems. In each case, existing commercial units are being assessed initially, along with viable alternative approaches. R and D work also included demonstrations of remote handling operations on full-size, partial mock-ups of the CIT machine at the Oak Ridge National Laboratory (ORNL) Remote Operations and Maintenance Development Facility

  20. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M J; Oyinloye, J O; Chambers, I [Electrowatt Consulting Engineers and Scientists, Warrington, Cheshire (United Kingdom); Scott, C K [Atlantic Nuclear Services, Fredericton, NB (Canada); Omar, A M [Atomic Energy Control Board, Ottawa, ON (Canada)

    1991-12-31

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs.

  1. Ex-vessel molten core debris interactions at CANDU nuclear power plants

    International Nuclear Information System (INIS)

    Lewis, M.J.; Oyinloye, J.O.; Chambers, I.; Scott, C.K.; Omar, A.M.

    1990-01-01

    Currently, the Atomic Energy Control Board (AECB) of Canada is sponsoring a project with a long term objective of obtaining an evaluation, independent of the industry, of the consequences to the public and the environment of postulated severe accidents at a Canadian nuclear power plant. Phase 1 of this project is a scoping study conducted to establish the relative consequences of a number of postulated event sequences. The studies in this paper model a multi-unit CANDU reactor at which pre-defined initiating events and their consequences could lead to severe core damage and relocation of the core debris onto the floor of the concrete reactor vault. Depending on the accident sequence assumptions made, an overlying pool of water may or may not be present. The US-NRC computer code CORCON Mod 2.0 was used to calculate the behaviour of the core material interacting with the concrete. The code calculates the decomposition of concrete by the molten core, and also the gases produced, which are released into the containment. The challenges to containment integrity are described, from the viewpoint of foundation decomposition and failure due to overpressure. The containment thermal-hydraulic behaviour is examined using an in-house computer code (CREM) written for this purpose. It is found that the containment envelope, in the absence of mitigating operator actions or design safety features, even for a case involving early core disassembly with the vacuum building unavailable, is unlikely to be failed within the 48 hours time frame examined. The paper identifies several areas for improvement in the models for future studies of core-concrete interactions for CANDU reactor plants. (author). 8 refs., 1 tab., 5 figs

  2. Ex-vessel remote maintenance development plans for the Burning Plasma Experiment

    International Nuclear Information System (INIS)

    Burgess, T.W.; Davis, F.C.

    1991-01-01

    Remote maintenance (RM) is fundamental to the basic design requirements of the Burning Plasma Experiment (BPX), and an extensive RM development and demonstration program is planned to meet these requirements. The program first draws from the experience base that exists in the fission community and Europe's Joint European Torus (JET) Project. Successful solutions are applied where possible and, in many cases, improved in order to achieve the performance demanded by a multiyear program that must be capable of efficiently executing RM procedures. Early, concurrent efforts in the design and fabrication of prototype remote handling (RH) equipment, remote tooling, and maintainable machine components will precede an extensive use of mock-up equipment in order to test, develop, and demonstrate the technology. 7 refs,. 5 figs

  3. Development on the ex-vessel transfer machine for Monju, (1)

    International Nuclear Information System (INIS)

    Inoue, Takashi; Kinuta, Kenji; Tomita, Takaaki

    1980-01-01

    As for the fast breeder reactors being developed as the national project, the construction of the prototype reactor ''Monju'' is started in 1980 based on the results of development of the experimental reactor ''Joyo''. In the fuel-handling facility, the development of which is promoted by Fuji Electric Co., Ltd., the development of the fuel transfer machines which take out and charge the fuel for the reactors and carry out the transport between fuel storage facilities is important subject. In this paper, the outline of the development of a fuel transfer machine is described, centering around the handling of fuel in sodium and the removal of the heat of spent fuel. ''Monju'' with 300 MW electric output is being developed for the purposes of putting fast breeder reactors in practical use, increasing plant power output, improving the rate of operation, and securing the safety. Plutonium is used as the fuel, and liquid metallic sodium is used as the coolant in fast breeder reactors, accordingly the fuel must be handled safely in gas-tight, shielded vessels while it is cooled, and securely in sodium and sodium cover gas by remote operation. Gripper, gripper-driving mechanism, coffin, movable block, door valves, cooling system and carriage compose the fuel transfer machine, and these are described. The main results of development and the tests for the development, such as the trial manufacture of gripper and decay heat removal test, are reported. (Kako, I.)

  4. Overview of the ACEX project iodine work

    Energy Technology Data Exchange (ETDEWEB)

    Merilo, M

    1996-12-01

    The ACEX project is an internationally sponsored research program that focuses on several aspects of severe accidents. The areas addressed are iodine behavior in containments, pool scrubbing, molten corium concrete interactions, and ex-vessel core debris coolability. These areas all represent extensions to the previous and current ACE and MACE programs respectively. The ACE-Phase B (iodine) project, and other recent research efforts, have clarified the roles of the important phenomena that influence iodine volatility in reactor containments during severe accidents. The ACE Iodine Chemistry Subcommittee concluded that even though enough data has been generated to support reasonably good quantification of the important phenomena, a few important areas remain where quantification is still uncertain. This is due to a lack of agreement on how to utilize the existing database, as well as the possible absence of critical test and/or property data. Technical resolution of the overall iodine behavior issue is therefore not feasible until these uncertainties are fully assessed and practical solutions have been identified, implemented, and verified. The overall objectives of the ACEX iodine research program are to ensure that the iodine database can be used to predict the airborne concentration of iodine, the conditions for iodine reservoir stability, and to provide a mechanistic understanding for these phenomena. The first phase of this work involves a comprehensive review and interpretation of the existing database in order to formulate practical strategies for dealing with significant uncertainties and/or deficiencies. Several projects are underway involving the effects of organic reactions and structural surface interactions. In addition effort is being expended on standardizing the aqueous iodine kinetics database, specifying useful mass transfer models, and defining methodology for pH prediction. (Abstract Truncated)

  5. Overview of the ACEX project iodine work

    International Nuclear Information System (INIS)

    Merilo, M.

    1996-01-01

    The ACEX project is an internationally sponsored research program that focuses on several aspects of severe accidents. The areas addressed are iodine behavior in containments, pool scrubbing, molten corium concrete interactions, and ex-vessel core debris coolability. These areas all represent extensions to the previous and current ACE and MACE programs respectively. The ACE-Phase B (iodine) project, and other recent research efforts, have clarified the roles of the important phenomena that influence iodine volatility in reactor containments during severe accidents. The ACE Iodine Chemistry Subcommittee concluded that even though enough data has been generated to support reasonably good quantification of the important phenomena, a few important areas remain where quantification is still uncertain. This is due to a lack of agreement on how to utilize the existing database, as well as the possible absence of critical test and/or property data. Technical resolution of the overall iodine behavior issue is therefore not feasible until these uncertainties are fully assessed and practical solutions have been identified, implemented, and verified. The overall objectives of the ACEX iodine research program are to ensure that the iodine database can be used to predict the airborne concentration of iodine, the conditions for iodine reservoir stability, and to provide a mechanistic understanding for these phenomena. The first phase of this work involves a comprehensive review and interpretation of the existing database in order to formulate practical strategies for dealing with significant uncertainties and/or deficiencies. Several projects are underway involving the effects of organic reactions and structural surface interactions. In addition effort is being expended on standardizing the aqueous iodine kinetics database, specifying useful mass transfer models, and defining methodology for pH prediction. The results of this work are expected to identify where additional data

  6. Experimental Study for Effects of the Stud shape of the Core Catcher System

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Seo, Gwang Hyeok; Shin, Doyoung; Jeun, Gyoodong; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2015-05-15

    In preparation of potential severe accidents, a nuclear power plant is equipped with diverse systems of engineering safety features or mitigation system dedicated to the severe accidents conditions. As a common strategy, a number of nuclear power plants adopt the in-vessel retention (IVR) and/or external reactor vessel cooling (ERVC) strategies. With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The core catcher system is for Ex-vessel in the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while keeping coolable geometry in case that the RPV failure occurs. The system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. As shown in Fig 1, the engineered cooling channel of the ex-vessel core catcher was adopted to remove sensible heat and decay heat of the molten corium using cooling water flooded from the In-Containment Refueling Water Storage Tank (IRWST) by gravity. A large number of studs are placed in the cooling channel to support the core catcher body. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the temperature distribution and overall coolability of the system. Thus, it is of importance to investigate the effects of studs on the coolability of the core catcher system. In the current research, to evaluate the effect of a stud on the streamline and natural convective boiling performance, numerical and experimental approaches were taken. As a part of numerical approach, CFD simulation using ANSYS/FLUENT was carried out. The objective was to predict disturbance of the streamline and temperature distribution due to the interference of the studs. Through the CFD

  7. OECD/MCCI 2-D Core Concrete Interaction (CCI) tests : final report February 28, 2006.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    reactor material database for dry cavity conditions is solely one-dimensional. Although the MACE Scoping Test was carried out with a two-dimensional concrete cavity, the interaction was flooded soon after ablation was initiated to investigate debris coolability. Moreover, due to the scoping nature of this test, the apparatus was minimally instrumented and therefore the results are of limited value from the code validation viewpoint. Aside from the MACE program, the COTELS test series also investigated 2-D CCI under flooded cavity conditions. However, the input power density for these tests was quite high relative to the prototypic case. Finally, the BETA test series provided valuable data on 2-D core concrete interaction under dry cavity conditions, but these tests focused on investigating the interaction of the metallic (steel) phase with concrete. Due to these limitations, there is significant uncertainty in the partition of energy dissipated for the ablation of concrete in the lateral and axial directions under dry cavity conditions for the case of a core oxide melt. Accurate knowledge of this 'power split' is important in the evaluation of the consequences of an ex-vessel severe accident; e.g., lateral erosion can undermine containment structures, while axial erosion can penetrate the basemat, leading to ground contamination and/or possible containment bypass. As a result of this uncertainty, there are still substantial differences among computer codes in the prediction of 2-D cavity erosion behavior under both wet and dry cavity conditions. In light of the above issues, the OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program was initiated at Argonne National Laboratory. The project conducted reactor materials experiments and associated analysis to achieve the following technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focused on providing both confirmatory evidence and test data for the

  8. Coolability in the frame of core melt accidents in light water reactors. Model development and validation for ATHLET-CD and ASTEC. Final report; Kuehlbarkeit im Rahmen von Kernschmelzunfaellen bei Leichtwasserreaktoren. Modellentwicklung und Validierung fuer ATHLET-CD und ASTEC. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Michael; Pohlner, Georg; Rahman, Saidur; Berkhan, Ana

    2015-07-15

    The code system ATHLET/ATHLET-CD is being developed in the frame of the reactor safety research of the German Federal Ministry for Economic Affairs and Energy (BMWi) within the topic analysis of transients and accident sequences. It serves for simulation of transients and accidents to be used in safety analyses for light water reactors. In the present project the development and validation of models for ATHLET-CD for description of the processes during severe accidents are continued. These works should enable broad safety analyses by a mechanistic description of the processes even during late phases of a degrading core and by this a profound estimation on coolability and accident management options during every phase. With the actual status of modelling in ATHLET-CD analyses on coolability are made to give a solid base for estimates about stabilization by cooling or accident progression, dependent on the scenario. The modeling in the MEWA module, describing the processes in a severely degraded core in ATHLET-CD, is extended on the processes in the lower plenum. For this, the model on melt pool behavior is extended and linked to the RPV wall. The coupling between MEWA and the thermal-hydraulics of ATHLET-CD is improved. The validation of the models is continued by calculations on new experiments and comparing analyses done in the frame of the European Network SARNET-2. For the European integral code ASTEC contributions from the modeling for ATHLET-CD will be done, especially by providing a model for the melt behavior in the lower plenum of a LWR. This report illustrates the work carried out in the frame of this project, and shows results of calculations and the status of validation by recalculations on experiments for debris bed coolability, melt pool behavior as well as jet fragmentation and debris bed formation.

  9. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident; Contribution des essais en materiaux prototypiques sur la plate-forme Plinius a l'etude des accidents graves de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch

    2008-01-15

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  10. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Journeau, Ch.

    2008-01-01

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  11. Coolability of severely degraded CANDU cores

    International Nuclear Information System (INIS)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Mijhawan, S.

    1995-07-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually re solidify. Thus, the calandria vessel would act inherently as a core-catcher as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author). 48 refs., 3 tabs., 18 figs

  12. Coolability of severely degraded CANDU cores. Revised

    International Nuclear Information System (INIS)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S.

    1996-01-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)

  13. High temperature chemical reactivity in the system (U, Zr,Fe, O). A contribution to the study of zirconia as a ``core catcher``; Reactivite chimique a haute temperature dans le systeme (U, Zr, Fe, O) contribution a l`etude de la zircone comme recuperateur de ``corium``

    Energy Technology Data Exchange (ETDEWEB)

    Maurizi, A [CEA Centre d` Etudes Nucleaires de Saclay, 91 -Gif-sur-Yvette (France); [CEA Centre d` Etudes de Saclay, 91 - Gif-sur-Yvette (France). Direction des Technologies Avancees

    1996-12-11

    Within the framework of the improvement of nuclear reactor safety, a device to recover corium is proposed to be installed under the reactor vessel to limit the consequences of a core melting. According to our bibliographic study, stabilised zirconia seems to be the best refractory material to play this role and to support the physicochemical, mechanical and thermal requirements imposed to the corium catcher. The nature of the chemical interactions between zirconia and iron of high temperature were established and experimental data on the (U, Fe, Zr, O) quaternary system which stands for the corium were determined. First of all, the Knudsen effusion mass-spectrometric method was used to establish the liquidus position for a (U, Zr, O) alloy representative of the corium (U/Zr = 1,5) at 2000 deg C. The oxygen solubility limit in a (U, Zr, O) liquid alloy is about 7 atomic %. In oxidising conditions, the reaction between zirconia and iron leads to the formation of a stabilised zirconia-iron oxide solid solution. Up to 10 atomic % of iron can be incorporated in the structure, leading to the stabilisation of cubic zirconia and a modification of lattice constants. The valence and localisation of those iron measured as a function of time and temperature from 1500 to 2400 deg C, after high frequency inductive heating, both on laboratory materials are commercial bricks. The reaction rate is governed by an activation energy of about 80 kJ/mol. Our results demonstrate that stabilised zirconia is able to efficiently absorb oxidised iron. (author). 169 refs.

  14. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

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

    1984-11-01

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

  15. Heat transfer between relocated materials and the RPV lower head

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Knudson, D.L. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States); Kohriyama, T. [INSS, Fukui (Japan)

    2001-07-01

    Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

  16. Heat transfer between relocated materials and the RPV lower head

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Kohriyama, T.

    2001-01-01

    Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

  17. Liquid entrainment through orifices by sparging gas

    International Nuclear Information System (INIS)

    Bonnet, J.M.; Malara, M.; Amblard, M.; Seiler, J.M.

    2001-01-01

    Corium Coolability by water flood during an MCCI (Molten Corium Concrete Interaction) is still an open problem. Several physical mechanisms have been identified which may reduce significantly and finally stop the ablation of concrete. Among these mechanisms, corium ejection by sparging gas into the overlying water may represent an important contribution. This mechanism was at the origin of a large and coolable debris bed and volcano formation in the MACE M3B test. This mechanism has also been observed in simulant material tests performed at UCSB and at FZK. The objective of the work, which is described in the present paper, is to model this mechanism and to quantify the liquid entrainment rate by sparging gas. (author)

  18. Cooling and spreading of corium during its fall into water in a pressurised water nuclear plant severe accident: description of mechanical and thermal interactions in a three phase flow during spreading of cold or heated spheres in a liquid pool; Refroidissement et dispersion du corium lors de sa chute dans l'eau pendant un accident severe de reacteur nucleaire a eau pressurisee: description des interactions mecaniques et thermiques en ecoulement triphasique lors de la dispersion de spheres solides froides ou chaudes dans un bain liquide

    Energy Technology Data Exchange (ETDEWEB)

    Duplat, F

    1998-10-26

    In the frame of nuclear safety studies about corium and water interactions, we address spreading and cooling stage of corium fragments in a liquid pool. Considering the complexity of encountered flow regimes and mechanical and thermal interactions coupling, modelling validation is based on a thermal-hydraulic computer code (MC3D). A bibliographical study shows that classical modelling of three phase flow is based on constitutive laws already established in the case of two phase flow. The present study states a complete analysis of BILLEAU experiments and defines a characterisation method for a sphere cloud. Some complementary QUEOS experiments are also described. Mechanical interaction terms such as added mass, lift and turbulent dispersion have been presented in the frame of a three phase flow and their influence has been tested in numerical simulations of BILLEAU tests. The effect of film vapour overheat, as well as particle diameter evolution have been studied. Moreover a radiative heat transfer modelling developed in Karlsruhe research centre (FZK) has been analysed and completed. Numerical simulations achieved for this study show that mechanical and thermal behaviour of the system are actually coupled. Taking into account lift and turbulent dispersion terms as well as heat transfer modifications all wed better results. This study also presents some considerations about flow regimes identification as a preliminary for studies about numerical diffusion that was already estimated in the present state of the computer code MC3D. (author)

  19. Status of direct containment heating in CSNI member countries. Report of task group on ex-vessel thermal-hydraulics

    International Nuclear Information System (INIS)

    1989-03-01

    The status of activities on direct containment heating in the light water reactor program in OECD/CSNI countries is presented. Experimental and analytical studies are reviewed. Approaches or measures are discussed for accident management in relation to direct containment heating. A discussion is given of common and diverging views among the countries based, in part, on response to a questionnaire. The key issues are discussed and recommendations are provided for future CSNI work on direct containment heating

  20. Ex-vessel melt-coolant interactions in deep water pool: Studies and accident management for Swedish BWRs

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Chu, C.C.; Spencer, B.W.; Frid, W.; Loewenhielm, G.

    1993-01-01

    In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL analyses

  1. Severe accident simulation and analysis for a CAREM-like integral nuclear reactor: ex-vessel phase

    International Nuclear Information System (INIS)

    Caputo, M.; García, J.M.; Giménez, M.; Sánchez, S.

    2013-01-01

    The main phenomena and processes involved in the progression of a hypothetical nuclear severe accident in an integral type reactor like CAREM are studied, quantifying the most relevant parameters, in order to contribute to the plant design and the development of an appropriate severe accident management program. A computational plant model was developed using Melcor code, including the reactor pressure vessel and the containment. A loss of coolant accident caused by a double guillotine pipe break in the steam dome zone of the pressure vessel (1.5 inches diameter) was simulated. Along this work the analysis were focused in the containment dynamics. As a consequence of the postulated loss of coolant accident the water inventory boils off leading to the core uncovery and fuel heat-up. At high temperatures the zircaloy steam oxidation becomes relevant, with hydrogen generation as one of the reaction products. The hydrogen produced is release into the containment and the possibility of hydrogen combustion in presence of enough oxygen makes relevant the analysis of containment hydrogen distribution. It is assumed that there is not any hydrogen control system. Due to the postulated loss of coolant a big amount of steam and energy is released into the containment, with a consequent fast pressurization of the dry well which makes possible air and steam discharging into the wet well (suppression pool). At the beginning the flow discharged into the pool is mainly composed of air, a non-condensable gas that pressurizes the wet well. After most of the containment air is pushed into the atmosphere wet well the pressurization rate decreases because the flow discharge is mainly composed by steam, which condensates in the pool. Also some other containment pressure peaks were observed as a consequence of hydrogen deflagrations. (author)

  2. A remote in-vessel and ex-vessel force-reflecting telerobotic system for the burning plasma experiment

    International Nuclear Information System (INIS)

    Kuban, D.P.; Busko, N.

    1992-01-01

    The Burning Plasma Experiment (BPX) has made an applaudable commitment to total remote maintenance which will undoubtedly move fusion energy closer to commercial reality. This commitment poses new and formidable challenges due to the dimensional constraints, diversity of maintenance operations, and the geometrically intricate equipment arrangements. These challenges must be addressed for successful hot operation of the Princeton Plasma Physics Laboratory BPX. This paper reports on a new manipulator system, called the TeleMate, which is under development to contribute to this needed capability. This system combines enhancements to a proven mechanical design with state-of-the-art controls technology, to produce a flexible system that can be configured to address the numerous remote fusion applications. The mechanical portion of the system has many years of operation in existing radioactive facilities. This paper presents a system description, the development status, initial test data, and control system performance

  3. Ex-vessel melt-coolant interactions in deep water pool: studies and accident management for Swedish BWRs

    International Nuclear Information System (INIS)

    Chu, C.C.; Sienicki, J.J.; Spencer, B.W.; Frid, W.; Loewenhielm, G.

    1995-01-01

    In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL-1 code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL-1 analyses. A description of the modeling incorporated in THIRMAL-1 is also provided. (orig.)

  4. Temperature-dependent attenuation of ex-vessel flux measurements at the Hanford Fast Flux Test Facility

    International Nuclear Information System (INIS)

    McLane, F.E.; Wood, M.R.; Rathbun, J.L.

    1982-01-01

    Indicated nuclear power, developed by measuring leakage neutrons, has been found to be temperature dependent at the Hanford Fast Flux Test Facility (FFTF). The magnitude, sense and speed of response of the effect suggest that hot sodium above th core and shield is a significant cause. Future designs which may minimize this effect are discussed

  5. Melt cooling by bottom flooding. The COMET core-catcher concept

    International Nuclear Information System (INIS)

    Foit, Jerzy Jan; Alsmeyer, Hans; Tromm, Walter; Buerger, Manfred; Journeau, Christophe

    2009-01-01

    The COMET concept has been developed to cool an ex-vessel corium melt in case of a hypothetical severe accident leading to vessel melt-through. After erosion of a sacrificial concrete layer the melt is passively flooded by bottom injection of coolant water. The open porosities and large surface that are generated during melt solidification form a porous permeable structure that is permanently filled with the evaporating water and thus allows an efficient short-term as well as long-term removal of the decay heat. The advantages of this concept are the fast cool-down and complete solidification of the melt within less than one hour typically. This stops further release of fission products from the corium. A drawback may be the fast release of steam during the quenching process. Several experimental series have been performed by FZK (Germany) to test and optimise the functionality of the different variants of the COMET concept. Thermite generated melts of iron and aluminium oxide were used. The large scale COMET-H test series with sustained inductive heating includes nine experiments performed with an array of water injection channels embedded in a sacrificial concrete layer. Variation of the water inlet pressure and melt height showed that melts up to 50 cm height can be safely cooled with an overpressure of the coolant water of 0.2 bar. The CometPC concept is based on cooling by flooding the melt from the bottom through layers of porous, water filled concrete. The third variant of the COMET design, CometPCA, uses a layer of porous, water filled concrete CometPCA from which flow channels protrude into the layer of sacrificial concrete. This modified concept combines the advantages of the original COMET concept with flow channels and the high resistance of a water-filled porous concrete layer against downward melt attack. Four large scale CometPCA experiments (FZK, Germany) have demonstrated an efficient cooling of melts up to 50 cm height using the recommended water

  6. Analysis of two-phase flow and boiling heat transfer in inclined channel of core-catcher

    International Nuclear Information System (INIS)

    Tahara, M.; Suzuki, Y.; Abe, N.; Kurita, T.; Hamazaki, R.; Kojima, Y.

    2008-01-01

    Passive Corium Cooling System (CCS) provides a function of ex-vessel debris cooling and molten core stabilization during a severe accident. CCS features inclined cooling channels arranged axi-symmetrically below the core-catcher basin. In order to estimate the coolability of the inclined cooling channel, it is indispensable to identify the flow pattern of the two-phase flow in the cooling channel. Several former studies for the two-phase flow pattern in the inclined channel are referred. Taitel and Dukler (1976) developed a prediction method of the flow pattern transition in horizontal and near horizontal tubes. Barnea et al. (1980) showed the flow pattern map of upward flow with 10 degrees inclination. Sakaguti et al. (1996) observed the two-phase flow patterns in the horizontal pipe connected with slightly upward pipe, in which the flow pattern in the pipe with a bending part was expressed by the combination of a basic flow pattern and some auxiliary flow patterns. Then we investigated these studies In order to identify the flow patterns observed in the inclined cooling channel of CCS. Furthermore we experimentally observed the flow patterns in the inclined cooling channel with various inlet conditions. As a result of the investigation and observation, typical flow patterns in the inclined cooling channel were identified. Two typical flow patterns were observed depending on the steam flow rate, one of which is 'elongated bubble 'flow, and the other is 'churn with collapsing backward and upward slug 'flow The flow and heat transfer in the inclined channel of CCS is analyzed by using a two-phase analysis code employing two-fluid model in which the constitutive equations for the two-phase flow in inclined channels are incorporated. That is, drift flux parameter for each of the elongated bubble flow, and the churn with collapsing backward and upward slug flow are incorporated to the two-phase analysis code, which are based on the rising velocity of the long bubble in

  7. Results and exploitation of FP-4 and FP-5 research in the area 'Safety of existing installations'. Part II

    International Nuclear Information System (INIS)

    Goethem, G. van; Zurita, A.; Manolatos, P.; Casalta, S.

    2004-01-01

    An overview is given of the most important achievements of the research programme co-financed by the European Union (EU) in the area of LWR safety over the FP-4 and FP-5 periods from the end-users point of view. The end-users are: the contracting organisations (i.e. utilities and associated engineering companies, regulatory bodies and associated technical safety organisations, manufacturing industry and associated services), the non-contracting organisations (including decision makers and opinion leaders) and the European Commission. Besides Community research strategy and programme implementation aspects in general, this paper is focusing on the S/T achievements obtained by multi-partner projects in the - 7 clusters of multi-partner projects in Euratom FP-4 (1994-1998): AGE for structural ageing, INV and EXV for in-vessel core degradation and ex-vessel molten corium coolability, ST for radiological source term, CONT for containment integrity, AMM for accident management measures and INNO for innovative safety features - the total cost of the 67 multipartner projects comprised in this Community research was Euro 71.3 million, out of which Euro 35.9 million was contributed by the EU budget - 3 clusters of multi-partner projects in Euratom FP-5 (1998-2002): PLEM for plant life extension and management; SAM for severe accident management and EVOL for evolutionary safety concepts - the total cost of the 71 multipartner projects comprised in this Community research is Euro 85.4 million, out of which Euro 43.5 million is contributed by the EU budget. The objectives of this Community research are discussed and a number of FP-4 and FP-5 projects are selected to demonstrate to what extent the proposed objectives were indeed met. Besides technological requirements, socio-economic aspects are becoming increasingly important due to the level of public and political acceptance and to the economic pressure of deregulated electricity markets; this is also discussed. Finally the

  8. Procedures for initiation, cost-sharing and management of OECD projects in nuclear safety

    International Nuclear Information System (INIS)

    2002-01-01

    The OECD (CSNI) projects aim to produce results relevant for the safe operation of nuclear power plants through international collaborative projects. In general, the projects consist of advanced experimental programmes that are conducted at specialized facilities. At present, the following OECD (CSNI) projects are in operation: - The Halden Project, covering fuel/materials and I and C/Human Factors issues; - The Cabri Project, addressing reactivity transients on high burnup fuels; - The MASCA Project, which deals with in-vessel corium phenomena; - The OLHF Project, dealing with lower head failure mechanisms; - The SETH Project addressing thermal-hydraulics issues, started in 2001; - The MCCI Project on ex-vessel coolability and melt-concrete interaction. There are significant differences among these projects in terms of their motivation, size and scope. The Halden Project and the Cabri Water Loop Project are large undertakings where the host organisations assume full and direct responsibility for the project establishment and administration - as well as for the negotiation with relevant parties on the terms of participation. In the other cases, instead, the NEA secretariat has a more direct responsibility, conferred by the CSNI, in establishing the project technical and financial basis, as well as for its implementation and administration. The objective of this procedure is to provide a common basis for the establishment and management of the OECD projects in the area of nuclear safety. It is a follow-up of a recommendation expressed by the CSNI Bureau during its meeting in October 2001, where the procedures for the establishment and management of the OECD (CSNI) projects in nuclear safety were addressed. While this procedure attempts at defining general guidelines for project initiation, financing and management, one should bear in mind that each project has its own motivation, background and framework. Thus, some degree of flexibility in project structure

  9. Evaluation of molten lead mixing in sodium coolant by diffusion for application to PAHR

    International Nuclear Information System (INIS)

    Chawla, T.C.; Pedersen, D.R.; Leaf, G.; Minkowycz, W.J.

    1983-01-01

    In post-accident heat removal (PAHR) applications the use of a lead slab is being considered for protecting a porous bed of steel shots in ex-vessel cavity from direct impingement of molten steel or fuel upon vessel failure following a hypothetical core dissembly accident in an LMFBR. The porous bed is provided to increase coolability of the fuel debris by the sodium coolant. The objectives of the present study are (1) to determine melting rates of lead slabs of various thicknesses in contact with sodium coolant and (2) to evaluate the extent of penetration and mixing rates of molten lead into sodium coolant by molecular diffusion alone

  10. Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

    Energy Technology Data Exchange (ETDEWEB)

    Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

    1998-01-01

    This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

  11. Current position on severe accident phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Henry, Robert E [Fauske and Associates, Inc., Burr Ridge, IL (United States)

    2004-07-01

    The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors.

  12. Current position on severe accident phenomena

    International Nuclear Information System (INIS)

    Henry, Robert E.

    2004-01-01

    The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors

  13. Thermal-Hydraulic Effects of Stud Shape and Size on the Safety Margin of Core Catcher System

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2015-10-15

    With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The newly engineered corium cooling system, that is, an ex-vessel core catcher system has been designed and adapted in some nuclear power plants such as VVER-1000, EPR, ESBWR, EU-APR1400 to mention a few. For example, Russia adopted a crucible-type core catcher for VVER-1000. On the other hand, a way to catch melt spreading is adopted by several countries, such as EPR in France, ESBWR in USA, ABWR in japan, and EU-APR1400 in Korea In Korea, the core catcher system has been designed and implemented for the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while maintaining a coolable geometry in case that RPV failure occurs. The core catcher system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the overall thermal-hydraulic performance including two-phase heat transfer coefficient and critical heat flux (CHF) of the system. Thus, it is of importance to investigate the thermal-hydraulic effects of studs on the coolability, especially the CHF of the core catcher system. With aforementioned importance, pool boiling experiments were carried out with stud shape of, rectangular, cylinder, and elliptic and for stud sizes of 10, 15, 20, and 25 mm under the condition of atmospheric saturated water. A particular attention was focused on observing local vapor behavior around the studs and finding any hot spots, where the vapors are accumulated. The occurrence of the CHF is anticipated at the back side of the studs. The visual observation and CHF measurements indicate that the

  14. Thermal-Hydraulic Effects of Stud Shape and Size on the Safety Margin of Core Catcher System

    International Nuclear Information System (INIS)

    Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Kim, Sung Joong

    2015-01-01

    With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The newly engineered corium cooling system, that is, an ex-vessel core catcher system has been designed and adapted in some nuclear power plants such as VVER-1000, EPR, ESBWR, EU-APR1400 to mention a few. For example, Russia adopted a crucible-type core catcher for VVER-1000. On the other hand, a way to catch melt spreading is adopted by several countries, such as EPR in France, ESBWR in USA, ABWR in japan, and EU-APR1400 in Korea In Korea, the core catcher system has been designed and implemented for the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while maintaining a coolable geometry in case that RPV failure occurs. The core catcher system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the overall thermal-hydraulic performance including two-phase heat transfer coefficient and critical heat flux (CHF) of the system. Thus, it is of importance to investigate the thermal-hydraulic effects of studs on the coolability, especially the CHF of the core catcher system. With aforementioned importance, pool boiling experiments were carried out with stud shape of, rectangular, cylinder, and elliptic and for stud sizes of 10, 15, 20, and 25 mm under the condition of atmospheric saturated water. A particular attention was focused on observing local vapor behavior around the studs and finding any hot spots, where the vapors are accumulated. The occurrence of the CHF is anticipated at the back side of the studs. The visual observation and CHF measurements indicate that the

  15. Research and development strategy on the behavior of containments during severe accidents

    International Nuclear Information System (INIS)

    Lecomte, C.

    1990-06-01

    In case of an hypothetical severe accident leading to core melting, the last barrier preventing radionucleide release in the environnment is the containment of the main reactor building. The French research and development programmes aimed at understanding the containment behavior during severe accidents relate to several domains; some of them are: - assessment of hydrogen behavior - corium behavior and coolability - ultimate resistance of the containments and leaktightness - caracterization of filtered venting procedure. All these aspects are covered by code calculations and experimental developments

  16. In-vessel coolability and retention of a core melt

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Liu, C.; Additon, S.

    1997-01-01

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. The technical treatment in this assessment includes: (a) new data on energy flow from either volumetrically heated pools or non-heated layers on top, boiling and critical heat flux in inverted, curved geometries, emissivity of molten (superheated) samples of steel, and chemical reactivity proof tests, (b) a simple but accurate mathematical formulation that allows prediction of thermal loads by means of convenient hand calculations, (c) a detailed model programmed on the computer to sample input parameters over the uncertainty ranges, and to produce probability distributions of thermal loads and margins for departure from nucleate boiling at each angular position on the lower head, and (d) detailed structural evaluations that demonstrate that departure from nucleate boiling is a necessary and sufficient criterion for failure. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is open-quotes physically unreasonable.close quotes Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings

  17. Study of ex-vessel steam explosion risk of Reactor Pit Flooding System and structural response of containment for CPR1000"+ Unit

    International Nuclear Information System (INIS)

    Zhang Juanhua; Chen Peng

    2015-01-01

    Reactor Pit Flooding System is one of the special mitigation measures for severe accident for CPR1000"+ Unit. If the In-Vessel Relocation function of Reactor Pit Flooding System is failed, there is the steam explosion risk in reactor cavity. This paper firstly adopts MC3D code to build steam explosion model in order to calculate the pressure load and impulses of steam explosion that are as the input data of containment structural response analysis. The next step is to model the containment structure and analyze the structural response by ABAQUS code. The analysis results show that the integral damage induced by steam explosion to the external containment wall is shallow, and the containment structural integrity can be maintained. The risk and damage to the containment integrity reduced by steam explosion of RPF is small, and it does not influence the design and implementation of RPF. (author)

  18. United States Nuclear Regulatory Commission research program on core debris/concrete interactions and ex-vessel fission-product release

    International Nuclear Information System (INIS)

    Burson, S.B.

    1987-01-01

    The study of core debris/concrete interaction phenomena has been a significant element of the NRC's Severe Accident Research Program for a number of years. The CORCON and VANESA codes used to predict the consequences of high-temperature debris attack on concrete and fission-product aerosol release are state-of-the-art computational tools. The major thrust of current NRC sponsored research focuses on the refinement, verification, and validation of these codes. An overview of the analytical and experimental aspects of the NRC research program is presented

  19. Phenomenological Studies on Melt-Structure-Water Interactions (MSWI) during Postulated Severe Accidents: Year 2004 Activity. APRI 5 report

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Park, H.S.; Nayak, A.K.; Hansson, R.C.; Chiferaw, D.; Stepanyan, A.; Rao, R.S.; Karbojian, A. [Royal Inst. of Technology, Stockholm (Sweden). Div. of Nuclear Power Safety

    2005-04-01

    This report presents descriptions of the major results obtained in the research program 'Melt-Structure-Water Interaction (MSWI)' at NPS/RIT during the year 2004. The primary objectives of the MSWI Project in year 2004 were to study (1) the in-vessel and exvessel melt/debris bed coolability process when melt is flooded with water, and (2) the energetics and characteristics of steam explosions. Our general approaches are to establish scaling relationships so that the data obtained in the experiments could be extended to prototypical accident geometries and conditions, develop phenomenological or computational models for the processes under investigation and validate the existing and newly developed models against data obtained at RIT and at other laboratories. In 2004, several experimental programs, such as the COMECO (Corium MElt COolability), POMECO (POrous MEdia COolability) and MISTEE (Micro-Interactions in STeam Explosion Experiments) programs were continued. The SIMECO (Simulation of MElt Coolability) program was restarted in 2004. The construction of the POMECO-GRAND (POrous MEdia COolability) facility was delayed due to lack of finances. However, existing POMECO facility was modified to study 3-D effects on debris coolability. In this report, the results from the COMECO experiment with high temperature oxidic melt, from the POMECO experiments for the multi-dimensional effects on debris bed coolability, from the SIMECO experiment for three-layer pool configuration and from the MISTEE experiments for steam explosion characteristics and loads are described. For analytical efforts, results from the COMETA code for the entire process of the steam explosions are discussed.

  20. Physical and numerical modelling of corium spreading with solidification in safety studies of pressurized water reactors; Modelisation physique et numerique de l`etalement d`un fluide avec solidification dans le cadre des etudes de surete pour les reacteurs eau sous pression

    Energy Technology Data Exchange (ETDEWEB)

    Eberle Patrick [Service d`Etude et de Modelisation en Thermohydrolique, CEA/DRN/DTP/SMTH, Grenoble (France)]|[Grenoble-1 Univ., 74 Annecy (France)

    1997-12-12

    In the frame of severe accidents of nuclear pressurized water reactor, it is important to understand and to model phenomena of corium spreading with solidification. The first part of the study describes experiments with simulating materials as well as simple models of the literature. We deduce a model where the equations of conservation are averaged over the volume. This model gives interesting results for continuous spreading but it is not convenient for discontinuous phenomena. A more precise model is then necessary. In the second part of this study, we present a complete model from which the basic idea is to average the conservation equations over the fluid height, supposing the characteristic fluid thickness is small in comparison with the characteristic spreading length. This model describes the thermalhydraulic aspects of the spreading as well as the mechanical behaviour of the upper crust. The liquid phases are supposed to be stratified and have a Newtonian fluid behaviour. The dynamical crust model takes into account a non-linear behaviour law. This law depends on the deformation tensor whereas the liquid behaviour low, depends on the rate of deformation tensor, so it is necessary to link this two notions by supplementary equations. The operation of averaging the equations gives terms at the interfaces which must be determined by constitutive laws. We deduce laws by fixing the velocity and temperature profile in the fluid height. The previous system of equations is discretized by finite volumes and semi-implicit methods. The discretized models are included in the specific code THEMA. The results of the model show good agreement with available experimental results. (author) 9 refs., 45 figs., 42 tabs.

  1. An overview of past and present activities in the severe accident domain within the framework of WGAMA

    International Nuclear Information System (INIS)

    Guentay, S.

    2011-01-01

    The objectives of the NEA-CSNI’s Working Group on the Analysis and Management of Accidents (WGAMA) are to assess and where necessary strengthen the technical basis needed for the prevention, mitigation, and management of potential accidents in nuclear power plants, and to facilitate international convergence on safety issues and accident management analyses and strategies. In order to fulfill this objective, the working group undertakes: Exchange technical experience and information relevant for resolving current or emerging safety issues; Promote the development of phenomena-based models and codes used for the safety analysis, including the performance of benchmarking exercises; Assess the state of knowledge in areas relevant for the accident analysis and where needed; Promote research activities aimed to improve such understanding, while supporting the maintenance of expertise and infrastructure in nuclear safety research. Continuing to be active in the severe accident field as the successor of the previous principle working group 4 (PWG4) of CSNI, WGAMA has accumulated an immense consolidated knowledge, which has been created along the years and reflects the improved understanding in very complex severe accident phenomenology, their modeling and their risk and safety relevance. WGAMA activities related to severe accidents include exclusively the following technical areas: progression of accident into core damage and associated in-vessel phenomena; coolability of over-heated cores; ex-vessel corium interaction with concrete and coolant; in-containment combustible gas control; physical-chemical behavior of radioactive species in the containment. The activities mainly focus on existing reactors, but will also comprise applications for some advanced reactor designs. Being established in 2000, WGAMA carried out the activities which had been initiated by the former PWG 4 group for the first few years, and initiated several more afterwards. A more targeted approach

  2. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  3. KAPOOL experiments to simulate molten corium - sacrificial concrete interaction

    International Nuclear Information System (INIS)

    Eppinger, B.; Fieg, G.; Tromm, W.

    2001-01-01

    In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. In the planned European Pressurized Reactor (EPR) the core melt is retained in the reactor cavity for ∼ 1 h to pick up late melts after the failure of the reactor pressure vessel. The reactor cavity is protected by a layer of sacrificial concrete and closed by a melt gate at the bottom towards the spreading compartment. After erosion of the sacrificial concrete and melt-through of the gate the core melt should be distributed homogeneously into the spreading compartment. There the melt is cooled by flooding with water. The knowledge of the sacrificial concrete erosion phase in the reactor cavity is essential for the severe accident assessment. Several KAPOOL experiments have been performed to investigate the erosion of two possible compositions of sacrificial concretes using alumina-iron thermite melts as a simulant for the core melt. Erosion rates as a function of the melt temperature and the inhomogeneity of the melt front are presented in this paper. (authors)

  4. Stress analysis and scaling studies of corium crusts

    International Nuclear Information System (INIS)

    Feng, Z.; Engelstad, R.L.; Lovell, E.G.; Corradini, M.L.

    1992-01-01

    In the event of a severe accident in a LWR, water may be input to cool the molten mixture of fuel and concrete. A number of structural models are developed and used to predict whether a crust will be formed and remain stable between the melt and water. Bending stresses and membrane stresses due to pressure loadings and the temperature differential are considered in the analyses to investigate the stability of the crust as a function of the time, thickness and span. The results from parametric studies show the conditions under which a crust could develop, and how such structural models could be used to determine scaling effects and provide correlations to prototypic accident situations. (orig.)

  5. Computational simulation of turbulent natural convection in a corium pool

    International Nuclear Information System (INIS)

    Vieira, Camila B.; Su, Jian; Niceno, Bojan

    2013-01-01

    After a severe accident in a nuclear power plant, the total thermal loading on the vessel of a nuclear reactor is controlled by the convective heat transfer. Taking that fact into account, this work aimed to analyze the turbulent natural convection inside a representative lower head cavity. By means of an open-source CFD code, OpenFOAM (Open Field Operation and Manipulation), numerical simulations were performed to investigate a volumetrically heated fluid (Pr = 7.0) at internal Rayleigh (Ra) numbers ranging from 10 8 to 10 15 . Bearing in mind that severe accident scenario and the physical-chemical effects are many and complex, the fluid analyzed was considered Newtonian, with constant physical properties, homogeneous and single phase. Even working with that simplifications, the modeling of turbulent natural convection has posed a considerable challenge for the Reynolds Averaged Navier-Stokes (RANS) equations based models, not only because of the complete unsteadiness of the flow and the strong turbulence effects in the near wall regions, but also because of the correct treatment of the turbulent heat fluxes (θu i ). So, this work outlined three approaches for treating the turbulent heat fluxes: the Simple Gradient Diffusion Hypothesis (SGDH), the Generalized Gradient Diffusion Hypothesis (GGDH) and the Algebraic Flux Model (AFM). Simulations performed at BALI test based geometry with a four equations model, k-ε-v 2 -f (commonly called as v 2 -f and V2-f), showed that despite of AFM and GGDH have provided reasonable agreement with experimental data for turbulent natural convection in a differentially heated cavity, they proved to be very unstable for buoyancy-driven flows with internal source in comparison to SGDH model. (author)

  6. Computational simulation of turbulent natural convection in a corium pool

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Camila B.; Su, Jian, E-mail: camila@lasme.coppe.ufrj.br, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Cursos de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Niceno, Bojan, E-mail: bojan.niceno@psi.ch [Paul Scherrer Institut (PSI), Villigen (Switzerland). Nuclear Energy and Safety

    2013-07-01

    After a severe accident in a nuclear power plant, the total thermal loading on the vessel of a nuclear reactor is controlled by the convective heat transfer. Taking that fact into account, this work aimed to analyze the turbulent natural convection inside a representative lower head cavity. By means of an open-source CFD code, OpenFOAM (Open Field Operation and Manipulation), numerical simulations were performed to investigate a volumetrically heated fluid (Pr = 7.0) at internal Rayleigh (Ra) numbers ranging from 10{sup 8} to 10{sup 15}. Bearing in mind that severe accident scenario and the physical-chemical effects are many and complex, the fluid analyzed was considered Newtonian, with constant physical properties, homogeneous and single phase. Even working with that simplifications, the modeling of turbulent natural convection has posed a considerable challenge for the Reynolds Averaged Navier-Stokes (RANS) equations based models, not only because of the complete unsteadiness of the flow and the strong turbulence effects in the near wall regions, but also because of the correct treatment of the turbulent heat fluxes (θu{sub i}). So, this work outlined three approaches for treating the turbulent heat fluxes: the Simple Gradient Diffusion Hypothesis (SGDH), the Generalized Gradient Diffusion Hypothesis (GGDH) and the Algebraic Flux Model (AFM). Simulations performed at BALI test based geometry with a four equations model, k-ε-v{sup 2} -f (commonly called as v{sup 2}-f and V2-f), showed that despite of AFM and GGDH have provided reasonable agreement with experimental data for turbulent natural convection in a differentially heated cavity, they proved to be very unstable for buoyancy-driven flows with internal source in comparison to SGDH model. (author)

  7. Analyses of corium spreading in Mark I containment geometry

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Chu, C.C.; Farmer, M.T.

    1991-01-01

    An assessment of melt spreading in the Mark I system has been carried out using the MELTSPREAD-1 computer code together with supporting analyses. Application of MELTSPREAD-1 confirms the calculation of shell survival in a wet containment for the most probable melt release conditions from NUREG/CR-5423. According to MELTSPREAD-1, a dry containment also may not be threatened by melt spreading. This reflects the heat losses undergone by the melt in the process of spreading to the shell conservatively neglected in NUREG/CR-5423. However, there exist parameter ranges outside the most probable set where shell failure may be calculated. Accounting for the breakup and quenching of melt relocating through a deep layer of subcooled water also conservatively neglected in NUREG/CR-5423 can reduce the set of parameter variations for which containment failure is calculated in the wet case

  8. Numerical predictions of natural convection in a uniformly heated pool

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Cho, D.H.

    1993-01-01

    In the event of a core meltdown accident, one of the accident progression paths is fuel relocation to the lower reactor plenum. In the heavy-water new production reactor (NPR-HWR) design, the reactor cavity is flooded with water. In such a design, decay heat removal to the water in the reactor cavity and thence to the containment may be adequate to keep the reactor vessel temperature below failure limits. If this is the case, the accident progression can be arrested by retaining a coolable corium configuration in the lower reactor plenum. The strategy of reactor cavity flooding to prevent reactor vessel failure from molten corium relocation to the reactor vessel lower head has also been considered for commercial pressurized water reactors

  9. Interaction between the radiative flux emitted by a corium melt and aerosols from corium/concrete interaction

    Energy Technology Data Exchange (ETDEWEB)

    Zabiego, M.; Cognet, G. [CEA-DRN/DER/SERA - CE Cadarache, Saint-Paul-Lez-Durance (France); Henderson, D. [Univ. of Wisconsin, Madison, WI (United States)

    1995-09-01

    In this paper we present a one-dimensional numerical model that deals with radiative transfer in a medium where aerosols are present. This model is written with the aim of performing radiative transfer calculations in the framework of severe Pressurized Water Reactor accidents, especially during the last stage of such an accident Molten Core Concrete Interaction (MCCI) when aerosols are very numerous. We explain the theoretical basis of our model, writing the general radiative transfer equation, knowing that aerosol droplets participate in radiation transport. We then simplify this equation for a one-dimensional medium and we propose to solve it using the spherical harmonics approximation. This gives us the radiative intensity and we can then deduce the radiative flux. Aerosol optical properties (extinction and scattering coefficients) are also required in such a calculation. They are determined using Rayleigh or Mie theory, depending, depending on the aerosol size. In order to provide an example of results one can expect from such a calculation, we applied our model to a test problem with given aerosol size and concentration distributions. Our example does not model any experiment explicitly but the physical conditions used are very close to the L4 test from the Advanced Containment Experiment (ACE) program.

  10. Prediction of mass fraction of agglomerated debris in a LWR severe accident

    International Nuclear Information System (INIS)

    Kudinov, P.; Davydov, M.

    2011-01-01

    Ex-vessel termination of accident progression in Swedish type Boiling Water Reactors (BWRs) is contingent upon efficacy of melt fragmentation and solidification in a deep pool of water below reactor vessel. When liquid melt reaches the bottom of the pool it can create agglomerated debris and “cake” regions that increase hydraulic resistance of the bed and affect coolability of the bed. This paper discusses development and application of a conservative-mechanistic approach to quantify mass fractions of agglomerated debris. Experimental data from the DEFOR-A (Debris Bed Formation and Agglomeration) tests with high superheat of binary oxidic simulant material melt is used for validation of the methods. Application of the approach to plant accident analysis suggests that melt superheat has less significant influence on agglomeration of the debris than jet penetration depth. The paper also discusses the impact of the uncertainty in the jet disintegration and penetration behavior on the agglomeration mode map. (author)

  11. The LEONAR code: a new tool for PSA Level 2 analyses

    International Nuclear Information System (INIS)

    Tourniaire, B; Spindler, B.; Ratel, G.; Seiler, J.M.; Iooss, B.; Marques, M.; Gaudier, F.; Greffier, G.

    2011-01-01

    The LEONAR code, complementary to integral codes such as MAAP or ASTEC, is a new severe accident simulation tool which can calculate easily 1000 late phase reactor situations within a few hours and provide a statistical evaluation of the situations. LEONAR can be used for the analysis of the impact on the failure probabilities of specific Severe Accident Management measures (for instance: water injection) or design modifications (for instance: pressure vessel flooding or dedicated reactor pit flooding), or to focus the research effort on key phenomena. The starting conditions for LEONAR are a set of core melting situations that are separately calculated from a core degradation code (such as MAAP, which is used by EDF). LEONAR describes the core melt evolution after flooding in the core, the corium relocation in the lower head (under dry and wet conditions), the evolution of corium in the lower head including the effect of flooding, the vessel failure, corium relocation in the reactor cavity, interaction between corium and basemat concrete, possible corium spreading in the neighbour rooms, on the containment floor. Scenario events as well as specific physical model parameters are characterised by a probability density distribution. The probabilistic evaluation is performed by URANIE that is coupled to the physical calculations. The calculation results are treated in a statistical way in order to provide easily usable information. This tool can be used to identify the main parameters that influence corium coolability for severe accident late phases. It is aimed to replace efficiently PIRT exercises. An important impact of such a tool is that it can be used to make a demonstration that the probability of basemat failure can be significantly reduced by coupling a number of separate severe accident management measures or design modifications despite each separate measure is not sufficient by itself to avoid the failure. (authors)

  12. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

    This report describes and discusses the containment accident progression and the important severe accident containment thermohydraulic phenomena. The overall objective of the report is to provide a rather detailed presentation of the present status of phenomenological knowledge, including an account of relevant experimental investigations and to discuss, to some extent, the modelling approach used in the MAAP 3.0 computer code. The MAAP code has been used in Sweden as the main tool in the analysis of severe accidents. The dependence of the containment accident progression and containment phenomena on the initial conditions, which in turn are heavily dependent on the in-vessel accident progression and phenomena as well as associated uncertainties, is emphasized. The report is in three parts dealing with: * Swedish reactor containments, the severe accident mitigation programme in Sweden and containment accident progression in Swedish PWRs and BWRs as predicted by the MAAP 3.0 code. * Key non-energetic ex-vessel phenomena (melt fragmentation in water, melt quenching and coolability, core-concrete interaction and high temperature in containment). * Early containment threats due to energetic events (hydrogen combustion, high pressure melt ejection and direct containment heating, and ex-vessel steam explosions). The report concludes that our understanding of the containment severe accident progression and phenomena has improved very significantly over the parts ten years and, thereby, our ability to assess containment threats, to quantify uncertainties, and to interpret the results of experiments and computer code calculations have also increased. (au)

  13. Cooling of an internal-heated debris bed with fine particles

    International Nuclear Information System (INIS)

    Yang, Z.L.; Sehgal, B.R.

    2001-01-01

    In this paper, an analytical model on dryout heat flux of ex-vessel debris beds with fines particles under top flooding conditions has been developed. The parametric study is performed on the effect of the stratification of the debris beds on the dryout heat flux. The calculated results show that the stratification configuration of the debris beds with smaller particles and lower porosity layer resting on the top of another layer of the beds has profound effect on the dryout heat flux for the debris beds both with and without a downcomer. The enhancement of the dryout heat flux by the downcomer is significant. The efficiency of the single downcomer on the enhancement of the dryout heat flux is also analyzed. This, in general, agrees well with experimental data. The model is also employed to perform the assessment on the coolability of the ex-vessel debris bed under representative accidental conditions. One conservative case is chosen, and it is found that the downcomer could be efficient measure to cool the debris bed and hence terminate the severe accident. (authors)

  14. Low pressure injection sequence sensitivity study of the M1 module of MEDICI

    International Nuclear Information System (INIS)

    Corradini, M.L.; Moses, G.A.; Norkus, J.K.; Welzbacker, R.T.

    1985-01-01

    In order to assess the consequences of a PWR containment failure and the ensuing radiological source term following a severe reactor accident, it is necessary to understand the ex-vessel behavior of the molten core. The M1 module of MEDICI models the dynamic fuel-coolant mixing, energetic interaction, and ejection of fuel and coolant from the reactor cavity following such an accident. A sensitivity study of the low pressure injection sequence was performed utilizing a Box-Behnken statistical design to treat five sets of input variables considered to be significant in the mixing and steam explosion processes. The low pressure injection sequence was studied in which the molten corium is modeled as a pour stream entering the cavity without entraining or sweeping out fuel or coolant

  15. Nuclear reactor cavity floor passive heat removal system

    Science.gov (United States)

    Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

    2018-03-06

    A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

  16. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y.; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1997-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  17. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1998-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  18. Analysis of Two Phase Natural Circulation Flow in the Cooling Channel of the PECS

    Energy Technology Data Exchange (ETDEWEB)

    Park, R. J; Ha, K. S; Rhee, B. W; Kim, H. Y [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Decay heat and sensible heat of the relocated and spread corium are removed by the natural circulation flow at the bottom and side wall of the core catcher and the top water cooling of the corium. The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. As shown in Fig. 1, the engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting in the PECS. To maintain the integrity of the ex-vessel core catcher, however, it is necessary that the coolant be sufficiently circulated along the inclined cooling channel to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. For this reason, a verification experiment on the cooling capability of the EU-APR1400 core catcher has been performed in the CE (Cooling Experiment)-PECS facility at KAERI. Preliminary simulations of two-phase natural circulation in the CE-PECS were performed to predict two-phase flow characteristics and to determine the natural circulation mass flow rate in the flow channel. In this study, simulations of two-phase natural circulation in a real core catcher of the PECS have been performed to determine the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code.

  19. Simulation of core melt spreading with lava: theoretical background and status of validation

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Breest, A.; Spengler, C.

    2000-01-01

    The goal of this paper is to present the GRS R and D achievements and perspectives of its approach to simulate ex-vessel core melt spreading. The basic idea followed by GRS is the analogy of core melt spreading to volcanic lava flows. A fact first proposed by Robson (1967) and now widely accepted is that lava rheologically behaves as a Bingham fluid, which is characterized by yield stress and plastic viscosity. Recent experimental investigations by Epstein (1996) reveal that corium-concrete mixtures may be described as Bingham fluids. The GRS code LAVA is based on a successful lava flow model, but is adapted to prototypic corium and corium-simulation spreading. Furthermore some detailed physical models such as a thermal crust model on the free melt surface and a model for heat conduction into the substratum are added. Heat losses of the bulk, which is represented by one mean temperature, are now determined by radiation and by temperature profiles in the upper crust and in the substratum. In order to reduce the weak mesh dependence of the original algorithm, a random space method of cellular automata is integrated, which removes the mesh bias without increasing calculation time. LAVA is successfully validated against a lot of experiments using different materials spread. The validation process has shown that LAVA is a robust and fast running code to simulate corium-type spreading. LAVA provides all integral information of practical interest (spreading length, height of the melt after stabilization) and seems to be an appropriate tool for handling large core melt masses within a plant application. (orig.)

  20. Review of IVR-ERVC and using flooding concept for application to high power reactor

    International Nuclear Information System (INIS)

    Lee, Min ho; Heo, Hyo; Bang, In Cheol

    2016-01-01

    Accident scope will be limited in the RPV. For example, in case of Fukushima, they have difficulties for cleanup the accident and even catching the location of the melt-through corium. Therefore, IVR-ERVC is the right strategy for mitigation of the severe accident. However, in case of high power reactors, there is a Critical Heat Flux (CHF) problem in its application to high power reactor. If CHF occurred, boiling regime changes from effective nucleate boiling to ineffective film boiling, so temperature of the RPV goes up and finally the RPV fails. To solve the CHF problem, here have been a lot of works for IVR-ERVC. In the point of in-vessel heat transfer, Theofanous suggested risk oriented accident analysis methodology which is a combination of probabilistic and deterministic approach. A lot of experiments have been done using simulants of corium in various experimental apparatus. Their simulants were usually water due to simulate large Rayleigh number and natural circulation of corium. IVR-ERVC concept has been researched for a long time. For in-vessel heat transfer, simulants or real corium was used to get a heat flux distribution to the outer wall. And based on those results, ex-vessel cooling has been researched in various geometry to get cooling limit as CHF. Material flooding is suggested as improvement of ERVC in APR 1400 to secure safety margin for CHF. Regardless of Prandtl number of the flooding material, the focusing effect of heat flux was mitigated; the maximum heat flux was reduced less than half of the maximum heat flux in bare condition

  1. Modes of heat removal from a heat-generating debris bed

    International Nuclear Information System (INIS)

    Squarer, D.; Hochreiter, L.E.; Piecznski, A.T.

    1984-01-01

    In the worst hypothetical accident in a light water reactor, when all protection systems fail, the core could be converted into a deep particulate bed either in-vessel or ex-vessel. The containment of such an accident depends on the coolability of a heat-generating debris bed. Some recent experimental and analytical studies that are concerned with heat removal from such a particulate bed are reviewed. Studies have indicated that bed dryout flux and, therefore, the heat removal rate from the particulate bed increases with the particle diameter (i.e., the permeability) for pool boiling conditions and can exceed the critical heat flux of a flat plate. Bed dryout in a large particle bed (i.e., a few millimetres) was found to be closely related to the ''flooding'' limit of the bed. Dryout under forced flow conditions was found to be affected by both forced and natural convection for mass flow rate smaller than m /SUB cr/ , whereas above this mass flow rate, bed dryout is proportional to the mass flow rate. Recent analyses were found to be in agreement with experimental data; however, additional research is needed to assess factors not accounted for in previous studies (e.g., effect of pressure, multidimensionality, stratification, etc.). Based on the expected pressure and particle sizes in a postulated severe accident sequence, a debris bed should be coolable, given a sufficient water supply

  2. Phenomena in the interaction among a core melt and protective and sacrificial materials

    International Nuclear Information System (INIS)

    Steinwarz, W.; Koller, W.; Dyllong, N.; Fischer, M.; Hellmann, S.; Lansmann, V.; Nie, M.; Haefner, W.; Alkan, Z.; Andrae, P.; Rensing, B.

    2000-01-01

    In a postulated core meltdown accident in a light water reactor there are bound to be interactions, in the ex-vessel phase, among the core melt and the structural materials within and below the reactor cavity. In existing plants, these structural materials normally are structural concrete, while future, evolutionary reactor lines are to have sacrificial and protective materials specially designed for this hypothetical case. To add to the state of knowledge about the phenomena occurring, experiments need to be conducted under conditions as realistic as possible. Within the research programs funded by the European Union, the German Federal Ministry for Economics, and the German nuclear power plant operators, experiments on a laboratory as well as an industrial scale on these problems are being carried out in the two projects called CORESA (COrium on REfractory and SAcrificial materials) and ECOSTAR (Ex-vessel COre melt STAbilization Research). The experiments are accompanied by an extensive analytical theoretical program also serving to advance and validate computer codes on the problems under investigation. The projects, which are carried out with international European participation, are expected to allow a concept to be developed for managing postulated accident scenarios involving core meltdown for innovative nuclear power plants, and to provide findings on risk evaluation of plants now in operation so as to further develop accident management measures. (orig.) [de

  3. Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

    International Nuclear Information System (INIS)

    Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min

    2016-01-01

    As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI

  4. Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI.

  5. Preliminary Design of Optimized Reactor Insulator for Severe Accident Mitigation of APR1400

    International Nuclear Information System (INIS)

    Heo, Sun; Lee, Jae-Gon; Kang, Yong-Chul

    2007-01-01

    APR1400, a Korean evolutionary advance light water reactor, has many advanced safety feature to prevent and mitigate of design basis accident (DBA) and severe accident. When reactor cooling system (RCS) fails to cooling its core, the core melted down and the molten core gathers together on bottom of reactor vessel. The molten core hurts reactor vessel and is released to containment, which raises the release of radioactive isotopes and the heating of the containment atmosphere. Finally, the corium is accumulated in the bottom of reactor cavity and it also raises the Molten Core and Concrete Interaction (MCCI) and the heating of containment atmosphere. There are two strategies to cooling molten core. Those are in-vessel retention and ex-vessel cooling. At the early stage of APR1400 design, only ex-vessel cooling which is cooling of the molten core outside the vessel after vessel failure is considered based on EPRI Utility Requirement Document (URD) for Evolutionary LWR. However, a need has been arisen to reflect current research findings on severe accident phenomena and mitigation technologies to Korean URD and IVRERVC (In-Vessel corium Retention using Ex-Reactor Vessel Cooling) was adopted APR1400. The ERVC is not considered as a licensing design basis but based on the defense-in-depth principle and safety margin basis, which is the top-tier requirement of the severe accident mitigation design as stated in the KURD. The Severe Accident Management strategy for APR1400 is intended to aid the plant operating staff to secure reactor vessel integrity in the early stage of the severe accident. As a part of a design implementation of IVR-ERVC for APR1400, we developed the preliminary design requirement, design specification and conceptual design

  6. Experimental study of heat transfer in the slotted channels at CTF facility

    International Nuclear Information System (INIS)

    Asmolov, V.; Kobzar, L.; Nickulshin, V.; Strizhov, V.

    1999-01-01

    During core melt accident significant amount of core may relocate in the reactor pressure vessel lower head. During its cooling it may form cracks inside the corium and gap between corium and reactor vessel. Gap also may appear due to deformation of the lower head if its temperature exceed creep limit. Slotted channels ensure ingress of the cooling water into the corium, and exit of the generated steam. Study of the cool-down mechanism of the solid core debris in the lower head of the reactor vessel through gap and cracks is the objective of experimental work on the CTF facility. Thermal hydraulics in the heated channels closed from the bottom and flooded with the saturated water from the top of the channel, is characterized by the counterflow of the steam and water, attended by such specific phenomena as the dry out when boiling, flooding and overturning of the coming down flow of water at the certain flow rates of the steam going up, partial dry out of the channel, and reflooding from the top of the heated channel with the saturated water. The above phenomena may reveal independently or in different combinations depending on geometric parameters of the channel, heat release, and coolant parameters. Interchange of these processes with a certain cyclic sequence is possible. Experimental study was performed at the CTF (Coolability Test Facility) facility, which is a part of the thermohydraulic KC test facility in the RRC 'Kurchatov Institute'. Presented results are obtained at the CTF-1 test section which represents a vertical flat channel modeling a single crack in the solidified corium or the gap between the corium and reactor vessel

  7. The VULCANO spreading programme

    Energy Technology Data Exchange (ETDEWEB)

    Cognet, G.; Laffont, G.; Jegou, C.; Journeau, C.; Sudreau, F.; Pierre, J.; Ramacciotti, M. [CEA (Atomic Energy Commission), DRN/DER - Bat. 212, CEA Cadarache, 13108 St. Paul Lez Durance (France)

    1999-07-01

    Among the currently studied core-catcher projects, some of them suppose corium spreading before cooling, in particular the EPR (European Pressurized Reactor) core-catcher concept is based on mixing the corium with a special concrete, spreading the molten mixture on a large multi-layer surface cooled from the bottom and subsequently cooling by flooding with water. Therefore, melt spreading deserves intensive investigation in order to determine and quantify key phenomena which govern the stopping of spreading. In France, for some years, the Nuclear Reactor Division of the Atomic Energy Commission (CEA/DRN) has undertaken a large program to improve knowledge on corium behaviour and coolability. This program is based on experimental and theoretical investigations which are finally gathered in scenario and mechanistic computer codes. In this framework, the real material experimental programme, VULCANO, conducted within an European frame, is currently devoted to the study of corium spreading. In 1997 and 1998, several tests have been performed on dry corium spreading with various composition of melts. Although all the observed phenomena, in particular the differences between simulant and real material melts have not been yet totally explained, these tests have already provided a lot of information about: The behaviour of complex mixtures including refractory oxides, silica, iron oxides and in one case iron metal; Spreading progression, which was never stopped in any of these tests by a crust formation at the front; The structure of spread melts (porosity, crusts,...); Physico-chemical interaction between melt and the refractory substratum which was composed of zirconia bricks. (authors)

  8. The VULCANO spreading programme

    International Nuclear Information System (INIS)

    Cognet, G.; Laffont, G.; Jegou, C.; Journeau, C.; Sudreau, F.; Pierre, J.; Ramacciotti, M.

    1999-01-01

    Among the currently studied core-catcher projects, some of them suppose corium spreading before cooling, in particular the EPR (European Pressurized Reactor) core-catcher concept is based on mixing the corium with a special concrete, spreading the molten mixture on a large multi-layer surface cooled from the bottom and subsequently cooling by flooding with water. Therefore, melt spreading deserves intensive investigation in order to determine and quantify key phenomena which govern the stopping of spreading. In France, for some years, the Nuclear Reactor Division of the Atomic Energy Commission (CEA/DRN) has undertaken a large program to improve knowledge on corium behaviour and coolability. This program is based on experimental and theoretical investigations which are finally gathered in scenario and mechanistic computer codes. In this framework, the real material experimental programme, VULCANO, conducted within an European frame, is currently devoted to the study of corium spreading. In 1997 and 1998, several tests have been performed on dry corium spreading with various composition of melts. Although all the observed phenomena, in particular the differences between simulant and real material melts have not been yet totally explained, these tests have already provided a lot of information about: The behaviour of complex mixtures including refractory oxides, silica, iron oxides and in one case iron metal; Spreading progression, which was never stopped in any of these tests by a crust formation at the front; The structure of spread melts (porosity, crusts,...); Physico-chemical interaction between melt and the refractory substratum which was composed of zirconia bricks. (authors)

  9. Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR

    Energy Technology Data Exchange (ETDEWEB)

    Galushin, Sergey, E-mail: galushin@kth.se; Kudinov, Pavel, E-mail: pkudinov@kth.se

    2016-12-15

    Highlights: • A data base of the debris properties in lower plenum generated using MELCOR code. • The timing of safety systems has significant effect on the relocated debris properties. • Loose coupling between core relocation and vessel failure analyses was established. - Abstract: Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario

  10. A review of steam explosions with special emphasis on the Swedish and Finnish BWRs. APRI 4, Phase 2 Report

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Haraldsson, H.O.; Yang, Z.L. [Sehgal Konsult, Stockholm (Sweden)

    2002-04-01

    , to develop and/or improve fine fragmentation models, and to assess these models for use in reactor applications. The effects of a steam explosion of large or small yield on the debris coolability have been assessed, a low-porosity deep debris bed, which may be hard to cool, may be generated. An engineering assessment of the steam explosion energetics has been performed. The assessment covers a broader scope of scenarios of ex-vessel FCIs, i.e. including metallic-melt release scenarios. Results of the assessment indicate that the dynamic-loadings resulting from ex-vessel steam explosions in the ABB BWR plants may be smaller than the values predicted in a previous assessment.

  11. A review of steam explosions with special emphasis on the Swedish and Finnish BWRs. APRI 4, Phase 2 Report

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Haraldsson, H.O.; Yang, Z.L.

    2002-04-01

    , to develop and/or improve fine fragmentation models, and to assess these models for use in reactor applications. The effects of a steam explosion of large or small yield on the debris coolability have been assessed, a low-porosity deep debris bed, which may be hard to cool, may be generated. An engineering assessment of the steam explosion energetics has been performed. The assessment covers a broader scope of scenarios of ex-vessel FCIs, i.e. including metallic-melt release scenarios. Results of the assessment indicate that the dynamic-loadings resulting from ex-vessel steam explosions in the ABB BWR plants may be smaller than the values predicted in a previous assessment

  12. Research and development with regard to severe accidents in pressurised water reactors: Summary and outlook

    International Nuclear Information System (INIS)

    2011-01-01

    This document reviews the current state of research on severe accidents in France and other countries. It aims to provide an objective vision, and one that's as exhaustive as possible, for this innovative field of research. It will help in identifying R and D requirements and categorising them hierarchically. Obviously, the resulting prioritisation must be completed by a rigorous examination of needs in terms of safety analyses for various risks and physical phenomena, especially in relation to Level 2 Probabilistic Safety Assessments. PSA-2 should be sufficiently advanced so as not to obscure physical phenomena that, if not properly understood, might result in substantial uncertainty. It should be noted that neither the safety analyses nor PSA-2 are presented in this document. This report describes the physical phenomena liable to occur during a severe accident, in the reactor vessel and the containment. It presents accident sequences and methods for limiting impact. The corresponding scenarios are detailed in Chapter 2. Chapter 3 deals with in-vessel accident progression, examining core degradation (3.1), corium behaviour in the lower head (3.2), vessel rupture (3.3) and high-pressure core meltdown (3.4). Chapter 4 focuses on phenomena liable to induce early containment failure, namely direct containment heating (4.1), hydrogen risk (4.2) and steam explosions (4.3). The phenomenon that could lead to a late containment failure, namely molten core-concrete interaction, is discussed in Chapter 5. Chapter 6 focuses on problems related to in-vessel and ex-vessel corium retention and cooling, namely in-vessel retention by flooding the primary circuit or the reactor pit (6.1), cooling of the corium under water during the corium-concrete interaction (6.2), corium spreading (6.3) and ex-vessel core catchers (6.4). Chapter 7 relates to the release and transport of fission products (FP), addressing the themes of in-vessel FP release (7.1) and ex-vessel FP release (7.3), FP

  13. In-vessel coolability and retention of a core melt. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T. [California Univ., Santa Barbara, CA (United States). Center for Risk Studies and Safety

    1996-10-01

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.

  14. Mechanistic study of fuel freezing, channel plugging, and continued coolability during fast reactor overpower excursions

    International Nuclear Information System (INIS)

    Wong, K.W.; Catton, I.; Kastenberg, W.E.

    1977-07-01

    A mechanistic model is presented which describes events following fuel pin failure which may lead to in-channel fuel plate-out. The thermal and hydraulic effects of the plate-out fuel are also evaluated. Given the amount and particle size of the fuel injected into the coolant channel during fuel pin failure, and the initial conditions of the interaction zone, the physical states of the fuel particles and the coolant in the interaction zone can be determined. The trajectories of the fuel particles in the coolant channel are determined by assuming a slip factor between the local tangential velocities of the coolant and the fuel particles. The time and distance after which a fuel particle hits a wire wrap are then determined and the impact stresses induced in the thin solid fuel crust can be evaluated

  15. Refined model for the coolability of core debris with flow entry from the bottom

    International Nuclear Information System (INIS)

    Schulenberg, T.; Mueller, U.

    1986-01-01

    Within the context of a hypothetical severe accident in light water reactors also heat generating debris beds of a coarse particle size are discussed. A refined model for two-phase flow in particle beds is presented. Compared to previous models this model takes into account the effect of interfacial drag forces between liquid and vapor. These effects are important in coarse debris beds. The model is based on the momentum equations for separated flow, which are closed by empirical relations for the wall shear stress and the interfacial drag. When the refined model is applied to LWR severe accident scenarios an increased dryout heat flux is predicted for debris beds with flow entry from the bottom driven by a moderate downcomer head

  16. In-vessel coolability and retention of a core melt. Volume 2

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T.

    1996-10-01

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided

  17. In-vessel coolability and retention of a core melt. Volume 1

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T.

    1996-10-01

    The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided

  18. An Evaluation of Triggering Timing for the TROI Tests

    International Nuclear Information System (INIS)

    Park, Ik Kyu; Kim, Jong Hwan; Hong, Seong Wan

    2009-01-01

    In the postulated reactor severe accident, the molten corium can be poured into the remained reactor coolant of the lower pressure vessel or the reactor cavity. This might severely threaten the containment integrity, and thus, the experimental and analytical efforts have been done to reveal this risk. It is generally received by the steam explosion experts that the in-vessel steam explosion steam explosion would not challenge the integrity of the vessel and the containment. The ex-vessel explosion, however, cannot be excluded from the factor to threaten the integrity of the cavity and more the reactor vessel. The worse thing of ex-vessel situation is that water is subcooled under a relative low pressure. The results of steam explosion experiments indicate that the subcooled water under a low pressure might be a good environment to make a strong steam explosion. Furthermore, the calculation results for evaluating ex-vessel steam explosion work are too scattered each other. Thus, the conversion ratio of ex-vessel explosion is still remained as a resolved issue. SERENA phase 2 project which has been conducted since 1st Oct. 2007 is aimed a resolution of the uncertainties on the void fraction, the solidification, and the melt composition effect by performing a limited number of well-designed tests with advanced instrumentations to clarify the nature of a prototypic material with mild steam explosion characteristics and to provide innovative experimental data for a computer code validation. The steam explosion results such as explosion pressure, conversion ratio, and the debris configuration are strongly affected by the initial conditions. Meanwhile, some events like triggering magnitude might not be important for the steam explosion results. The initial conditions of the steam explosion, called premixture, are determined by three factors: first one is melt and water condition, second one is a mixing process, and the other is the triggering timing. The mixing process is

  19. Validation of Code ASTEC with LIVE-L1 Experimental Results

    International Nuclear Information System (INIS)

    Bachrata, Andrea

    2008-01-01

    The severe accidents with core melting are considered at the design stage of project at Generation 3+ of Nuclear Power Plants (NPP). Moreover, there is an effort to apply the severe accident management to the operated NPP. The one of main goals of severe accidents mitigation is corium localization and stabilization. The two strategies that fulfil this requirement are: the in-vessel retention (e.g. AP-600, AP- 1000) and the ex-vessel retention (e.g. EPR). To study the scenario of in-vessel retention, a large experimental program and the integrated codes have been developed. The LIVE-L1 experimental facility studied the formation of melt pools and the melt accumulation in the lower head using different cooling conditions. Nowadays, a new European computer code ASTEC is being developed jointly in France and Germany. One of the important steps in ASTEC development in the area of in-vessel retention of corium is its validation with LIVE-L1 experimental results. Details of the experiment are reported. Results of the ASTEC (module DIVA) application to the analysis of the test are presented. (author)

  20. Post Analysis of Two Phase Natural Circulation Mass Flow Rate for CE-PECS

    Energy Technology Data Exchange (ETDEWEB)

    Park, R. J.; Ha, K. S.; Rhee, B. W.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is required that the coolant be circulated at a rate along the inclined cooling channel sufficient to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. In this study, post simulations of two phase natural circulation in the CEPECS have been performed to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. Post simulations of two phase natural circulation in the CE-PECS have been conducted to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is approximately 8.7 kg/s in the base case.

  1. Development of integrated computer code for analysis of risk reduction strategy

    International Nuclear Information System (INIS)

    Kim, Dong Ha; Kim, See Darl; Kim, Hee Dong

    2002-05-01

    The development of the MIDAS/TH integrated severe accident code was performed in three main areas: 1) addition of new models derived from the national experimental programs and models for APR-1400 Korea next generation reactor, 2) improvement of the existing models using the recently available results, and 3) code restructuring for user friendliness. The unique MIDAS/TH models include: 1) a kinetics module for core power calculation during ATWS, 2) a gap cooling module between the molten corium pool and the reactor vessel wall, 3) a penetration tube failure module, 4) a PAR analysis module, and 5) a look-up table for the pressure and dynamic load during steam explosion. The improved models include: 1) a debris dispersal module considering the cavity geometry during DCH, 2) hydrogen burn and deflagration-to-detonation transition criteria, 3) a peak pressure estimation module for hydrogen detonation, and 4) the heat transfer module between the molten corium pool and the overlying water. The sparger and the ex-vessel heat transfer module were assessed. To enhance user friendliness, code restructuring was performed. In addition, a sample of severe accident analysis results was organized under the preliminary database structure

  2. Post Analysis of Two Phase Natural Circulation Mass Flow Rate for CE-PECS

    International Nuclear Information System (INIS)

    Park, R. J.; Ha, K. S.; Rhee, B. W.; Kim, H. Y.

    2015-01-01

    The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is required that the coolant be circulated at a rate along the inclined cooling channel sufficient to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. In this study, post simulations of two phase natural circulation in the CEPECS have been performed to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. Post simulations of two phase natural circulation in the CE-PECS have been conducted to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is approximately 8.7 kg/s in the base case

  3. In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

    In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

  4. An overview of the severe accident research activities within the LACOMERA platform at the Forschungszentrum Karlsruhe

    International Nuclear Information System (INIS)

    Miassoedov, A.; Alsmeyer, H.; Meyer, L.; Steinbrueck, M.; Tromm, W.

    2006-01-01

    The LACOMERA project at the Forschungszentrum Karlsruhe, Germany, is a 4 year action within the 5th Framework Programme of the EU which started in September 2002. Overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO, and COMET. These facilities can be used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. The paper summarises the main results obtained in the following experiments performed up to now. QUENCH-L1: Impact of air ingression on core degradation. The test provides unique data for the investigation of air ingress phenomenology in conditions as representative of a spent fuel pool accident as possible; QUENCH-L2: Boil-off of a flooded bundle. The test is of a generic interest for all reactor types, provided a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. DISCO-L1: Thermal hydraulic behaviour of the corium melt dispersion neglecting the chemical effects such as hydrogen generation and combustion. COMET-L1: Long-term 2D concrete ablation in a siliceous concrete cavity at intermediate decay heat power level with a top flooding phase after a phase of dry concrete erosion. COMET-L2: Investigation of long-term melt-concrete interaction of metallic corium in a cylindrical siliceous concrete cavity under dry conditions with decay heat simulation of intermediate power during the first test phase, and subsequently at reduced power during the second test phase. (author)

  5. Results of MACE tests M0 and M1

    International Nuclear Information System (INIS)

    Spencer, B.W.; Farmer, M.T.; Armstrong, D.R.; Kilsdonk, D.J.; Aeschlimann, R.W.; Fischer, M.

    1992-01-01

    This document discusses the Melt Attack and Coolability Experiment (MACE) Program underway at Argonne National Laboratory under ACE/EPRI sponsorship. The program addresses the efficacy of water to terminate an accident situation if melt progression were to result in a molten core/concrete interaction (MCCI) in the reactor containment. Large-scale experiments are being conducted in parallel with related modeling efforts, involving the addition of water to an MCI already underway. The experiments utilize UO 2 /ZrO 2 /Zr corium mixtures, direct electrical heating for simulation of decay heating, and various types of concrete basemats. Currently the tests involve 430 kg corium mass, 25 cm depth, in a 50 cm square test section. Test MO was a successful scoping test, but the first full size test, Ml, failed to achieve melt-water contact owing to existence of a preexisting bridge crust of corium charge. A heat flux of 3.5 MW/m 2 was measured in MO which removed energy from the corium pool equivalent to its entire heat of solidification prior to abatement by formation of an interfacial crust. The crust subsequently limited heat extraction to 600 kW/m 2 and less. Both tests MO and Ml revealed physical evidence of large pool swelling events which resulted in extrusion (and ejection) of melt into water above the crust, significantly increasing the overall quench and reducing the remaining melt in contact with the concrete. Furthermore, test Ml provided evidence of occasional ''burst mode'' ablation events and one additional important benefit of overlying water -- aerosol capture

  6. Basic Boiling Experiments with An Inclined Narrow Gap Associated With In-Vessel Retention

    International Nuclear Information System (INIS)

    Terazu, Kuninobu; Watanabe, Fukashi; Iwaki, Chikako; Yokobori, Seiichi; Akinaga, Makoto; Hamazaki, Ryoichi; SATO, Ken-ichi

    2002-01-01

    In the case of a severe accident with relocation of the molten corium into the lower plenum of reactor pressure vessel (RPV), the successful in-vessel corium retention (IVR) can prevent the progress to ex-vessel events with uncertainties and avoid the containment failure. One of the key phenomena governing the possibility of IVR would be the gap formation and cooling between a corium crust and the RPV wall, and for the achievement of IVR, it would be necessary to supply cooling water to RPV as early as possible. The BWR features relative to IVR behavior are a deep and massive water pool in the lower plenum, and many of control rod drive guide tubes (CRDGT) installed in the lower head of RPV, in which water is injected continuously except in the case of station blackout scenario. The present paper describes the basic boiling experiment conducted in order to investigate the boiling characteristics in an inclined narrow gap simulating a part of the lower head curvature. The boiling experiments were composed of visualization tests and heat transfer tests. In the visualization tests, two types of inclined gap were constructed using the parallel plate and the V-shaped parallel plate with heating from the top plate, and the boiling flow pattern was observed with various gap width and heat flux. These observation results showed that water was easily supplied from the gap bottom of parallel plate even in a very narrow gap with smaller width than 1 mm, and water could flow continuously in the narrow gap by the geometric and thermal imbalance from the experiment results using the V-shaped parallel plate. In the heat transfer tests, the critical heat flux (CHF) data in an inclined narrow channel formed by the parallel plates were measured in terms of the parameters of gap width, heated length and inclined angle of a channel, and the effect of inclination was incorporated into the existing CHF correlation for a narrow gap. The CHF correlation modified for an inclined narrow gap

  7. OECM MCCI Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-2 final data report, Rev. 0 February 12, 2003

    International Nuclear Information System (INIS)

    Lomperski, S.; Farmer, M.T.; Kilsdonk, D.; Aeschlimann, B.

    2011-01-01

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼φ30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the second water ingression test, designated SSWICS-2. The test investigated the quench behavior of a 15 cm deep, fully

  8. OECD MMCI Small-Scale Water Ingression and Crust Strength tests (SSWICS) SSWICS-1 final data report, Rev. 1 February 10, 2003.; Report, Rev. 1

    International Nuclear Information System (INIS)

    Lomperski, S.; Farmer, M.T.; Kilsdonk, D.; Aeschlimann, B.

    2011-01-01

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure; and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼φ30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the first water ingression test, designated SSWICS-1. The test investigated the quench behavior of a 15 cm deep, fully

  9. Overview of in-vessel retention concept involving level of passivity: with application to evolutionary pressurized water reactor design

    International Nuclear Information System (INIS)

    Ghyym, Seong H.

    1998-01-01

    In this work, one strategy of severe accident management, the applicability of the in-vessel retention (IVR) concept, which has been incorporated in passive type reactor designs, to evolutionary type reactor designs, is examined with emphasis on the method of external reactor vessel cooling (ERVC) to realize the IVR concept in view of two aspects: for the regulatory aspect, it is addressed in the context of the resolution of the issue of corium coolability; for the technical one, the reliance on and the effectiveness of the IVR concept are mentioned. Additionally, for the ERVC method to be better applied to designs of the evolutionary type reactor, the conditions to be met are pointed out in view of the technical aspect. Concerning the issue of corium coolability/quenchability, based on results of the review, plausible alternative strategies are proposed. According to the decision maker's risk behavior, these would help materialize the conceptual design for evolutionary type reactors, especially Korea Next Generation Reactors (KNGRs), which have been developing at the Korea Electric Power Research Institute (KEPRI): (A1) Strategy 1A: strategy based on the global approach using the reliance on the wet cavity method; (A2) Strategy 1B: strategy based on the combined approach using both the reliance on the wet cavity method and the counter-measures for preserving containment integrity; (A3) Strategy 2A: strategy based on the global approach to the reliance on the ERVC method; (A4) Strategy 2B: strategy based on the balanced approach using both the reliance on the ERVC method and the countermeasures for preserving containment integrity. Finally, in application to an advanced pressurized water reactor (PWR) design, several recommendations are made in focusing on both monitoring the status of approaches and preparing countermeasures in regard to the regulatory and the technical aspects

  10. Severe accident research and management in Nordic Countries - A status report

    International Nuclear Information System (INIS)

    Frid, W.

    2002-01-01

    The report describes the status of severe accident research and accident management development in Finland, Sweden, Norway and Denmark. The emphasis is on severe accident phenomena and issues of special importance for the severe accident management strategies implemented in Sweden and in Finland. The main objective of the research has been to verify the protection provided by the accident mitigation measures and to reduce the uncertainties in risk dominant accident phenomena. Another objective has been to support validation and improvements of accident management strategies and procedures as well as to contribute to the development of level 2 PSA, computerised operator aids for accident management and certain aspects of emergency preparedness. Severe accident research addresses both the in-vessel and the ex-vessel accident progression phenomena and issues. Even though there are differences between Sweden and Finland as to the scope and content of the research programs, the focus of the research in both countries is on in-vessel coolability, integrity of the reactor vessel lower head and core melt behaviour in the containment, in particular the issues of core debris coolability and steam explosions. Notwithstanding that our understanding of these issues has significantly improved, and that experimental data base has been largely expanded, there are still important uncertainties which motivate continued research. Other important areas are thermal-hydraulic phenomena during reflooding of an overheated partially degraded core, fission product chemistry, in particular formation of organic iodine, and hydrogen transport and combustion phenomena. The development of severe accident management has embraced, among other things, improvements of accident mitigating procedures and strategies, further work at IFE Halden on Computerised Accident Management Support (CAMS) system, as well as plant modifications, including new instrumentation. Recent efforts in Sweden in this area

  11. In-vessel retention modeling capabilities in MAAP5

    International Nuclear Information System (INIS)

    Paik, Chan Y.; Lee, Sung Jin; Zhou, Quan; Luangdilok, W.; Reeves, R.W.; Henry, R.E.; Plys, M.; Scobel, J.H.

    2012-01-01

    Modular Accident Analysis Program (MAAP) is an integrated severe accident analysis code for both light water and heavy water reactors. New and improved models to address the complex phenomena associated with in-vessel retention (IVR) were incorporated into MAAP5.01. They include: -a) time-dependent volatile and non-volatile decay heat, -b) material properties at high temperatures, -c) finer vessel wall nodalization, -d) new correlations for natural convection heat transfer in the oxidic pool, -e) refined metal layer heat transfer to the reactor vessel wall and surroundings, -f) formation of a heavy metal layer, and -g) insulation cooling channel model and associated ex-vessel heat transfer and critical heat flux correlations. In this paper, the new and improved models in MAAP5.01 are described and sample calculation results are presented for the AP1000 passive plant. For the IVR evaluation, a transient calculation is useful because the timing of corium relocation, decaying heat load, and formation of separate layers in the lower plenum all affect integrity of the lower head. The key parameters affecting the IVR success are the metal layer emissivity and thickness of the top metal layer, which depends on the amount of steel in the oxidic pool and in the heavy metal layer. With the best estimate inputs for the debris mixing parameters in a conservative IVR scenario, the AP1000 plant results show that the maximum ex-vessel heat flux to CHF ratio is about 0.7, which occurs before 10.000 seconds when the decay heat is high. The AP1000 plant results demonstrate how MAAP5.01 can be used to evaluate IVR and to gain insight into responses of the lower head during a severe accident

  12. An assessment of Class-9 (core-melt) accidents for PWR dry-containment systems

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Saito, M.

    1981-01-01

    The phenomenology of core-melt accidents in dry containments was examined for the purpose of identifying the margins of safety in such Class-9 situations. The scale (geometry) effects appear to crucially limit the extent (severity) of steam explosions. This together with the established reduced explosivity of the corium-A/water system, and the inherently high capability of dry containments (redinforced concrete, and shields in some cases, seismic design etc.) lead to the conclusion that failure due to steam explosions may be considered essentially incredible. These premixture scaling considerations also impact ultimate debris disposition and coolability and need additional development. A water-flooded reactor cavity would have beneficial effects in limiting (but not necessarily eliminating) melt-concrete interactions. Independently of the initial degree of quenching and/or scale of fragmentation, mechanisms exist that drive the system towards ultimate stability (coolability). Additional studies, with intermediate-scale prototypic materials are recommended to better explore these mechanisms. Containment heat removal systems must provide the crucial capability of mitigating such accidents. Passive systems should be explored and assessed against currently available and/or improved active systems taking into account the rather loose time constraints required for activation. It appears that containment margins for accommodating the hydrogen problem are limited. This problem appears to stand out not only in terms of potential consequences but also in terms of lack of any readily available and clear cut solutions at this time. (orig.)

  13. Investigation of the Potential for In-Vessel Melt Retention in the Lower Head of a BWR by Cooling through the Control Rod Guide Tubes. APRl 4, Stage 2 Report

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Jasiulevicius, A.; Konovalikhin, M.

    2004-01-01

    recommended that further investigations, both experimental and model development, be conducted to (a) check reproducibility of data (b) employ different flow rates (c) employ different simulant materials and (d) develop a comprehensive model, in order to certify that the coolability that can be achieved with establishing a water flow in the CRGTs will be able to retain the melt in the lower head of a BWR. We believe it will be an extremely important accident management strategy for a Swedish BWR since it will obviate the consideration of the prime licensing issue of ex-vessel steam explosion induced containment failure associated with the present scheme of establishing a water pool in the lower drywell of all the Swedish BWRs

  14. Investigation of the structure of debris beds formed from fuel rods fragmentation

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Duc-Hanh; Fichot, Florian; Topin, Vincent, E-mail: vincent.topin@irsn.fr

    2017-03-15

    This paper is a study of debris beds that can form in the core of a nuclear power plant under severe accident conditions. Such beds are formed of fragments of pellets and cladding remnants, as observed in the TMI-2 core. Many important issues are related with the morphology of those debris beds: are they coolable in case of water injection and how does molten corium progress through them if they are not coolable? The answers to those questions depend on the structure of the debris bed: porosity, number and arrangement of particles. In order to obtain relevant information, a numerical simulation of the formation of the debris bed is proposed. It relies on a granular approach of the type called “Contact Dynamics” to simulate the collapse of debris and their accumulation. Two different schemes of fuel pellet fragmentation are considered and simulations for different degrees of fragmentation of the pellets are performed. The results show that the number of axial cracks on fuel pellets strongly influences the final porosity of the debris bed. Porosities vary between 31% (less coolable cases) and 45% (similar to TMI-2 observations), with a most probable configuration around 41%. The specific surface of the bed is also evaluated. In the last part, a simple model is used to estimate the impact of the variation in geometry of the numeric debris beds on their flow properties. We show that the permeability and passability can vary respectively with a range of 30% and 15% depending on the number of fragment per pellet. The other benefits of the approach are finally discussed. Among them, the possibility to print 3D samples from the calculated images of debris beds appears as a promising perspective to perform experiments with realistic debris beds.

  15. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Park, Hae Kyun; Kim, Sang Nyung

    2013-01-01

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

  16. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae Kyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

    2013-10-15

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was

  17. Evaluation of Ablation rate by the change of Sacrificial Material for PECS in EU-APR

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Do Hyun; Kim, Yong Soo; Lee, Keun Sung [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-05-15

    EU-APR, modified and improved from its original design of APR1400, has been developed to comply with European Utility Requirements (EUR) and nuclear design requirements of the European countries. In EU-APR, Severe Accident Mitigation Systems are dedicated to providing an independent defense line from that of Engineered Safety Feature (ESF) and Diverse Safety Feature (DSF). They consist of Emergency Reactor Depressurization System (ERDS), Passive Ex-vessel corium retaining and Cooling System (PECS), Severe Accident Containment Spray System (SACSS), Hydrogen Mitigation System (HMS) and Containment Filtered Vent System (CFVS). The PECS, so called core catcher, was introduced to prevent the Molten Core Concrete Interaction (MCCI) after Reactor Vessel (RV) failure. The PECS has experienced a lot of changes from its original design. Recently, the most significant change was that as a SM, limestone concrete is installed on PECS's body wall instead of previous sacrificial material rich in Fe{sub 2}O{sub 3}. The main reason of this design change is to overcome the issue that the sacrificial material is ablated rather too fast when reacting with corium that contains a large fraction of Zr metal. Other changes in the geometry of PECS's wall and downcomer design are considered as minor ones. In this paper, the comparison of ablation rates between previous SM and limestone concrete is carried out using MAAP5 code with respective MCCI model according to the material. In this paper, major improvements of MAAP5 model for PECS in EU-APR are presented and the evaluation of ablation rate for the previous SM model and the new LC model is carried out by means of ablation depths with LBLOCA sequence. Two models have respective unique ablation process. The ablation of LC model proceeds at a constant rate regardless of water while the ablation of SM model proceeds at a faster rate before the arrival of cooling water for corium and SM mixture. The change of sacrificial material

  18. Evaluation of Ablation rate by the change of Sacrificial Material for PECS in EU-APR

    International Nuclear Information System (INIS)

    Hwang, Do Hyun; Kim, Yong Soo; Lee, Keun Sung

    2015-01-01

    EU-APR, modified and improved from its original design of APR1400, has been developed to comply with European Utility Requirements (EUR) and nuclear design requirements of the European countries. In EU-APR, Severe Accident Mitigation Systems are dedicated to providing an independent defense line from that of Engineered Safety Feature (ESF) and Diverse Safety Feature (DSF). They consist of Emergency Reactor Depressurization System (ERDS), Passive Ex-vessel corium retaining and Cooling System (PECS), Severe Accident Containment Spray System (SACSS), Hydrogen Mitigation System (HMS) and Containment Filtered Vent System (CFVS). The PECS, so called core catcher, was introduced to prevent the Molten Core Concrete Interaction (MCCI) after Reactor Vessel (RV) failure. The PECS has experienced a lot of changes from its original design. Recently, the most significant change was that as a SM, limestone concrete is installed on PECS's body wall instead of previous sacrificial material rich in Fe 2 O 3 . The main reason of this design change is to overcome the issue that the sacrificial material is ablated rather too fast when reacting with corium that contains a large fraction of Zr metal. Other changes in the geometry of PECS's wall and downcomer design are considered as minor ones. In this paper, the comparison of ablation rates between previous SM and limestone concrete is carried out using MAAP5 code with respective MCCI model according to the material. In this paper, major improvements of MAAP5 model for PECS in EU-APR are presented and the evaluation of ablation rate for the previous SM model and the new LC model is carried out by means of ablation depths with LBLOCA sequence. Two models have respective unique ablation process. The ablation of LC model proceeds at a constant rate regardless of water while the ablation of SM model proceeds at a faster rate before the arrival of cooling water for corium and SM mixture. The change of sacrificial material also

  19. Contribution to the thermodynamic description of the corium - The U-Zr-O system

    Science.gov (United States)

    Quaini, A.; Guéneau, C.; Gossé, S.; Dupin, N.; Sundman, B.; Brackx, E.; Domenger, R.; Kurata, M.; Hodaj, F.

    2018-04-01

    In order to understand the stratification process that may occur in the late phase of the fuel degradation during a severe accident in a PWR, the thermodynamic knowledge of the U-Zr-O system is crucial. The presence of a miscibility gap in the U-Zr-O liquid phase may lead to a stratified configuration, which will impact the accidental scenario management. The aim of this work was to obtain new experimental data in the U-Zr-O liquid miscibility gap. New tie-line data were provided at 2567 ± 25 K. The related thermodynamic models were reassessed using present data and literature values. The reassessed model will be implemented in the TAF-ID international database. The composition and density of phases potentially formed during stratification will be predicted by coupling current thermodynamic model with thermal-hydraulics codes.

  20. Reactor Vessel External Cooling for Corium Retention SULTAN Experimental Program and Modelling with CATHARE Code

    International Nuclear Information System (INIS)

    Rouge, S.; Dor, I.; Geffraye, G.

    1999-01-01

    In case of severe accident, a molten pool may form at the bottom of the lower head, and some pessimistic scenarios estimate that heat fluxes up to 1.5 MW/m 2 should be transferred through the vessel wall. An efficient, though completely passive, removal of heat flux during a long time is necessary to prevent total wall ablation, and a possible solution is to flood the cavity with water and establish boiling in natural convection. High heat exchanges are expected, especially if the system design (deflector along the vessel, riser...) emphasize water natural circulation, but are unfortunately limited by the critical heat flux phenomena (CHF). CHF data are very scarce in the adequate range of hydraulic and geometric parameters and are clearly dependent of the system effect in natural convection. The system effect can both modify flow velocity and two phase flow regimes, counter-current phenomena and flow static or dynamic instabilities. The SULTAN experimental program purpose was of two kinds, increasing CHF data for realistic situations, and improving the modeling of large 3D two phase flow circuits in natural convection. The CATHARE thermal-hydraulic code is used for interpreting the data and for extrapolation to real geometry. As a first step, a one-dimensional model is used. It is shown that some closure laws have to be improved. Reasonable predictions may be obtained but, for some test conditions, multi-dimensional effects such as recirculation appear to be dominant. Therefore the 3-dimensional module of CATHARE is also used to investigate these effects. This model well predicts qualitatively the existence and the development of a 2-phase layer along the heated wall as well as the existence of a recirculation zone. But modelling problems still require further development as part of a long term program for a better prediction of multi-dimensional two-phase flows

  1. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.

    1988-04-01

    Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals, and gas bubble nucleation in molten metals are relevant problems that are addressed in this work

  2. Experimental investigations on the compatibility of a SNR-type corium with graphite

    International Nuclear Information System (INIS)

    Peehs, M.; Friedrich, H.J.; Heuvel, H.J.

    1976-01-01

    Tests on the compatibility between UO 2 -pellets, UO 2 /stainless steel mixtures (70 w/o UO 2 , 30 w/o SS) and graphite were carried out at 1600, 2000, and 2400 0 C. The test duration was 0.5 h. UO 2 -pellets are only completely converted to carbides at 2400 0 C, when liquid U-carbide phases occur. In the presence of SS the liquefaction temperature was lowered to 2000 0 C. The conversion rates in the case of UO 2 -liquefaction as indicated by the corresponding CO-generation are high. The interactions are completed within a quarter of an hour. Apart from the release of O in the form of CO, the pick-up of C and the evaporation losses at the highest temperatures no significant changes in the chemical composition could be found. Post melting analyses by microsections, microprobe and X-ray investigations detected UFeC 2 -phases and carbides from the M 7 C 3 -type together with austenite. A thermodynamic evaluation revealed that the C-activity of the melts after UO 2 -liquefaction is nearly 1

  3. Radiation heat transfer in a pressurized water reactor lower head filled with molten corium

    International Nuclear Information System (INIS)

    Šadek, Siniša; Grgić, Davor; Debrecin, Nenad

    2013-01-01

    Highlights: ► We develop radiation heat exchange model for a reactor pressure vessel lower head. ► Model is used during a late in-vessel phase of severe accidents. ► View factors are calculated automatically for a time-dependent enclosure. ► Model is included in the RELAP5/SCDAPSIM computer code. ► Inclusion of heat radiation causes faster heat-up rate of RPV lower head structures. - Abstract: Following a core melt, molten material may slump to the lower head of a reactor pressure vessel (RPV). In that case, some structures like lower parts of fuel elements and a core support plate would remain intact. Since the melt is at high temperature and there are no obstacles between the melt and the supporting plate, the plate is exposed to an intense radiation heating. The radiation heat exchange model of the lower head was developed and applied to a finite element code COUPLE which is a part of the detailed mechanistic code RELAP5/SCDAPSIM. The radiation enclosure consisted of three surfaces: the upper surface of the relocated material, the inner surface of the RPV wall above the relocated material and the lower surface of the core support plate. View factors were calculated for the enclosure geometry that is changing in time because of intermittent accumulation of molten material. The enclosure surfaces were covered by mesh of polygonal areas and view factors were calculated, for each pair of the element areas, by solving the definite integrals using the algorithms for adaptive integrations by means of Gaussian quadrature. Algebraic equations for radiosity and irradiation vectors were solved by LU decomposition and the radiation model was explicitly coupled with the heat conduction model. The results show that there is a possibility of the core support plate failure after being heated up due to radiation heat exchange with the melt.

  4. Development of gap measurement technique in-vessel corium retention using ultrasonic pulse echo method

    International Nuclear Information System (INIS)

    Koo, Kil Mo; Kim, Jong Hwan; Kang, Kyung Ho; Kim, Sang Baik; Sim, Cheul Muu

    1999-03-01

    A gap between a molten material and a lower vessel is formed in the LAVA experiment, a phase 1 study of Sonata-IV program. In this technical report, quantitative results of the gap measurement using an off-line ultrasonic pulse echo method are presented. This report aims at development of an appropriate ultrasonics test method, by analyzing the problems from the external environmental reason and the internal characteristic reason. The signal analyzing methods to improve the S/N ratio in these problems are divided into the time variant synthesized signal analyzing method and the time invariant synthesized signal analyzing method. In this report, the possibility of the application of these two methods to the gap signal and the noise is considered. In this test, the signal of the propagational direction and reflectional direction through solid-liquid-solid specimen was analyzed to understand the behavior of the reflectional signal in a multi-layered structure by filling the gap with water between the melt and the lower head vessel. The quantitative gap measurement using the off-line ultrasonic pulse echo method was available for a little of the scanned region. But furtherly using DSP technique and imaging technique, the better results will be obtained. Some of the measured signals are presented as 2-dimensional spherical mapping method using distance and amplitude. Other signals difficult in quantitative measurement are saved for a new signal processing method. (author). 11 refs., 4 tabs., 54 figs

  5. Assessment of thema code against spreading experiments

    International Nuclear Information System (INIS)

    Spindler, B.; Veteau, J.M.; Cecco, L. de; Montanelli, P.; Pineau, D.

    2000-01-01

    In the frame work of severe accident research, the spreading code THEMA, developed at CEA/DRN, aims at predicting the spreading extent of molten core after a vessel melt-through. The code solves fluid balance equations integrated over the fluid depth for oxidic and/or metallic phases under the shallow water assumption, using a finite difference scheme. Solidification is taken into account through crust formation on the substrate and at contact with the surroundings, as well as increase of fluid viscosity with solid fraction in the melt. A separate energy equation is solved for the solid substrate, including possible ablation. The assessment of THEMA code against the spreading experiments performed in the framework of the corium spreading and coolability project of the European Union is presented. These experiments use either simulating materials at medium (RIT), or at high temperature (KATS), or corium (VULCANO, FARO), conducted at different mass flow rates and with large or low solidification interval. THEMA appears to be able to simulate the whole set of the experiments investigated. Comparison between experimental and computed spreading lengths and substrate temperatures are quite satisfactory. The results show a rather large sensitivity at mass flow rate and inlet temperature, indicating that, generally, efforts should be made to improve the accuracy of the measurements of such parameters in the experiments. (orig.)

  6. Validation of Numerical Schemes in a Thermal-Hydraulic Analysis Code for a Natural Convection Heat Transfer of a Molten Pool

    International Nuclear Information System (INIS)

    Kim, Jong Tae; Ha, Kwang Soon; Kim, Hwan Yeol; Park, Rae Joon; Song, Jin Ho

    2010-01-01

    It is postulated that a fuel of a water-cooled nuclear reactor can be melted during a hypothetical severe accident. There are two strategies for cooling the molten corium, which are in-vessel corium cooling and exvessel corium cooling. They can be chosen depending on cooling characteristics of the reactor. The coolability of the molten pool is determined by comparing the thermal load from the pool and the maximum heat flux removable by cooling mechanism such as radiative or boiling heat transfer on the pool boundaries. In order to evaluate the molten pool coolability, it is important to correctly expect the thermal load by a natural convection heat transfer of the corium pool. Many correlations have been developed by conducting experiments for the natural convection of a pool. The main parameters of the heat transfer by the natural convection are Rayleigh (Ra) number, Prandtl (Pr) number and the geometry of the pool. Sometimes, the use of the correlations for the evaluation of the thermal load from the molten pool is limited by a high Ra number of the pool and its different shape from the existing correlations. Computational fluid dynamics (CFD) has been used for the analysis of the heat transfer by a natural convection. In principle, CFD is applicable to the corium pool analysis. But unfortunately, some difficulties are encountered during the analysis, which are from numerical and physical instabilities. The physical instability is from turbulence fluctuation and inverted thermal layer near the upper surface of the volumetric-heated molten pool with a high Ra number. In order to resolve turbulent natural convection, buoyancy-modified two-equation turbulence models such as a k-e or k-w model with time-averaged Navier- Stokes equations are commonly used. Because an unsteadiness of a natural convection becomes nontrivial in a high Ra number pool, it is very difficult to get accurate heat flux on the pool surface with the time averaged turbulence model. Recently

  7. OECD MCCI Small-Scale Water Ingression and Crust Strength Tests (SSWICS) SSWICS-3 test data report: thermal Hydraulic results, Rev. 0 February 19, 2003

    International Nuclear Information System (INIS)

    Lomperski, S.; Farmer, M.T.; Kilsdonk, D.; Aeschlimann, B.

    2011-01-01

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core/concrete interaction (MCCI) when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. However, due to the integral nature of these tests, several questions regarding the crust freezing behavior could not be adequately resolved. These questions include: (1) To what extent does water ingression into the crust increase the melt quench rate above the conduction-limited rate and how is this affected by melt composition and system pressure and (2) What is the fracture strength of the corium crust when subjected to a thermal-mechanical load and how does it depend upon the melt composition? A series of separate-effects experiments are being conducted to address these issues. The first employs an apparatus designed to measure the quench rate of a pool of corium (∼φ30 cm; up to 20 cm deep). The main parameter to be varied in these quench tests is the melt composition since it is thought to have a critical influence on the crust cracking behavior which, in turn, alters quench rate. The issue of crust strength will be addressed with a second apparatus designed to mechanically load the crust produced by the quench tests. This apparatus will measure the fracture strength of the crust while under a thermal load created by a heating element beneath the crust. The two apparatuses used to measure the melt quench rate and crust strength are jointly referred to as SSWICS (Small-Scale Water Ingression and Crust Strength). This report describes results of the third water ingression test, designated SSWICS