Core disruptive accident analysis in prototype fast breeder reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Kannan, S.E.; Singh, Om Pal; Chetal, S.C.; Bhoje, S.B.

2002-01-01

Liquid metal cooled fast breeder reactors, in particular, pool type have many inherent and engineered safety features and hence a core disruptive accident (CDA) involving melt down of the whole core is a very low probable event ( -6 /ry). The important mechanical consequences such as straining of the main vessel including top shield, structural integrity of safety grade decay heat exchangers (DHX) and intermediate heat exchangers (IHX) sodium release to reactor containment building (RCB) through the penetrations in the top shield, sodium fire and consequent temperature and pressure rise in RCB are theoretically analysed using computer codes. Through the analyses with these codes, it is demonstrated that an energetic CDA capability to the maximum 100 MJ mechanical energy in PFBR can be well contained in the primary containment. The sodium release to RCB is 350 kg and pressure rise in RCB is ∼10 kPa. In order to raise the confidence on the theoretical predictions, very systematic experimental program has been carried out. Totally 67 tests were conducted. This experimental study indicated that the primary containment is integral. The main vessel can withstand the energy release of ∼1200 MJ. The structural integrity of IHX and DHX is assured up to 200 MJ. The transient force transmitted to reactor vault is negligible. The average water leak measured under simulated tests for 122 MJ work potential is about 1.8 kg and the maximum leak is 2.41 kg. Extrapolation of the measured maximum leak based on simulation principles yields ∼ 233 kg of sodium leak in the reactor. Based on the above-mentioned theoretical and experimental investigations, the design pressure of 20 kPa is used for PFBR

Proposal for computer investigation of LMFBR core meltdown accidents

International Nuclear Information System (INIS)

Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

1974-01-01

The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

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

International Nuclear Information System (INIS)

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

1976-01-01

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

A risk-based evaluation of LMFBR containment response under core disruptive accident conditions

International Nuclear Information System (INIS)

Hartung, J.; Berk, S.

1978-01-01

Probabilistic risk methodology is utilized to evaluate the failure modes and effects of LMFBR containment systems under Core Disruptive Accident (CDA) conditions. First, the potential causes of LMFBR containment failure under CDA conditions are discussed and categorized. Then, a simple scoping-type risk assessment of a reference design is presented to help place these potential causes of failure in perspective. The highest risk containment failure modes are identified for the reference design, and several design and research and development options which appear capable of reducing these risks are discussed. The degree to which large LMFBR containment systems must mitigate the consequences of CDA's to achieve a level of risk (for LMFBR's) comparable to the already very low risk of contemporary LWR's is explored. Based on the results of this evaluation, several suggestions are offered concerning CDA-related design goals and research and development priorities for large LMFBR's. (author)

Hypothetical core disruptive accident analysis of a 2000 MWsub(e) liquid metal cooled fast breeder reactor

International Nuclear Information System (INIS)

Struwe, D.

1977-12-01

A structural phase diagram for hypothetical core disruptive accidents (HCDA) has been developed based on a variety of analyses for different LMFBR's. The intention was to identify the strategic phases of HCDA's important with regard to safety aspects of the plant. These phases are investigated in detail for a 2,000 MWsub(e) LMFBR (SNR-2,000). Characteristic data of SNR-2,000 are discussed concerning their influence on safety analysis. Reasons for the choice of model parameters for special phenomena as fuel coolant interaction, fuel pin failure mechanisms and sodium voiding are given. The results of calculations with CAPRI-2, HOPE and KADIS are analyzed for possibilities to enter energetic core disassembly with consequences, making power values below 2,000 MWsub(e) necessary. Investigation of these results shows that the expected consequences do not lead to design requirements, restricting the magnitude of the electrical power output of LMFBR's to values below 2,000 MWsub(e). Therefore, consequences of HCDA's are principal not expected to limit the feasibility of conventional core design of this order of magnitude. (orig.) [de

The role of fission product in whole core accidents - research in the USA

Energy Technology Data Exchange (ETDEWEB)

Dietrich, L W [Argonne National Laboratory, Division of Reactor Analysis and Safety, Argonne, IL (United States); Jackson, J F [Los Alamos Scientific Laboratory, Q Division - Energy, Los Alamos, NM (United States)

1977-07-01

Safety of nuclear reactors has been a central concern of the nuclear energy industry from the very beginning. This concern, and the resultant excellence of design, fabrication, and operation, aided by extensive engineered safety features, has given nuclear energy its superior record of protection of the environment and of the public health and safety. With respect to the fast reactor, it was recognized early in the programme that there exists a theoretical possibility of a core compaction leading to significant energy release. Early analysis of this problem employed a number of conservative assumptions in attempting to bound the energy release. As reactors have grown in size, the suitability of such bounding calculations has diminished, and research into hypothetical accident analysis has emphasized a more mechanistic approach. In the USA, much effort has been directed towards modeling and computer code development aimed at following the course of an accident from its initiation to its ultimate conclusion with a stable, permanently subcritical, coolable core geometry, along with considerations of post-accident heat removal and radiological consequence assessment. Throughout this effort, the potential role of fission products has been recognized and account taken of the effects of fission products in determining accident progression. It is important to recognize that reactor safety is a very diverse topic, requiring consideration of a number of factors. While the major questions of public risk appear to be related to the hypothetical core disruptive accident (HCDA), it is necessary that the probability of having such an accident be extremely low In order that acceptable public risk be demonstrated. Such a demonstration requires sound engineering design and Implementation, with high standards of reliability, inspectability, maintainability, and operation, along with the requisite quality control and assurance. Tile current approach, typified by that taken by the

Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

2003-01-01

The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

SIMMER-I: an S/sub n/, Implicit, Multifield, Multicomponent, Eulerian, Recriticality code for LMFBR disrupted core analysis

International Nuclear Information System (INIS)

Bell, C.R.; Bleiweis, P.B.; Boudreau, J.E.; Parker, F.R.; Smith, L.L.

1976-08-01

Physical models, numerical methods, and program description are presented for SIMMER-I, a computer program which predicts the neutronic and fluid dynamic behavior of an LMFBR during a hypothetical core disruptive accident

Simulation of a hypothetical core disruptive accident in the mars test-facility

International Nuclear Information System (INIS)

Robbe, M.F.; Lepareux, M.

2001-01-01

In France, a large experimental programme MARA/MARS was undertaken in the 80's to estimate the mechanical consequences of an HCDA (Hypothetical Core Disruptive Accident) and to validate the SIRIUS computer code used at that time for the numerical simulations. At the end of the 80's, it was preferred to add a HCDA sodium-bubble-argon tri-component constitutive law to the general ALE fast dynamics finite element CASTEM-PLEXUS code rather than going on developing and using the specialized SIRIUS code. The experimental results of the MARA programme were used in the 90's to validate and qualify the CASTEM-PLEXUS code. A first series of computations of the tests MARA 8, MARA 10 and MARS was realised. The simulations showed a rather good agreement between the experimental and computed results for the MARA 8 and MARA 10 tests - even if there were some discrepancies - but the prediction of the MARS structure displacements and strains was overestimated. This conservatism was supposed to come from the fact that several MARS non axisymmetric structures like core elements, pumps and heat exchangers were not represented in the CASTEM-PLEXUS model. These structures, acting as porous barriers, had a protective effect on the mock-up containment by absorbing energy and slowing down the fluid impacting the containment. For these reasons, we developed in CASTEM-PLEXUS a new HCDA constitutive law taking into account the presence of the internal structures (without meshing them) by means of an equivalent porosity method. In other respects, the process used for dealing with the fluid-structure coupling in CASTEM-PLEXUS was improved. Thus a second series of simulations of the tests MARA8 and MARA10 was realised. A simulation of the test MARS was carried out too with the same simplified representation of the peripheral structures as in order to estimate the improvement provided by the new fluid-structure coupling. This paper presents a third numerical simulation of the MARS test with the

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Analysis of an out-of-pile experiment for materials redistribution under core disruptive accident condition of fast breeder reactors

International Nuclear Information System (INIS)

Sawada, Tetsuo; Ninokata, Hisashi; Shimizu, Akinao

1995-01-01

Calculation of one of the SIMBATH experiments was performed using the SIMMER-II code. The experiments were intended to simulate the fuel pin disintegration, the molten materials relocation and following materials redistribution that could occur during core disruptive accidents assumed in fast breeder reactors. The calculation by SIMMER-II showed that the incorporated step-wise fuel pin disintegration model and the modified particle jamming model were capable of reproducing the course of materials relocation within the identified ranges of the parameters which governed the blockages formation, i.e. the characteristic radius of solid particles jamming and/or sieving out in the flow and the effective particle viscosity. In particular the final materials redistribution calculated by SIMMER-II very well reproduced the experiment. This fact made it possible to interpret theoretically the mechanisms of flow blockages formation and related materials redistribution. (author)

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

An assessment of fuel freezing and drainage phenomena in a reactor shield plug following a core disruptive accident

International Nuclear Information System (INIS)

El-Genk, M.; Cronenberg, A.W.

1978-01-01

An important problem related to the assessment of the recriticality potential for an LMFBR following a core disruptive accident is an understanding of the freezing phenomena of molten fuel on a cold structure which may prevent fuel dispersal and sunsequent shutdown. Transient analytical freezing and drainage calculations have been applied to molten UO 2 travel through the rather cold lower shield plug of the Clinch River Breeder Reactor (CRBR). The successive approximation technique is used to obtain a solution of the non-linear freezing problem, where such effects as heat generation, viscous heat dissipation, temperature dependent thermophysical properties and a convective boundary condition at the solidification front have been incorporated into the present analytical formulation. Results indicate that previous steady-state analysis overestimate the rate of frozen layer build-up by about a factor of two. However, of primary importance is the driving force for drainage and the diameter of the shield plug flow channel. (Auth.)

Analysis and research status of severe core damage accidents

International Nuclear Information System (INIS)

1984-03-01

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

Effects of recent modeling developments in prompt burst hypothetical core disruptive accident calculations

International Nuclear Information System (INIS)

Sienicki, J.J.; Abramson, P.B.

1978-01-01

The main objective of the development of multifield, multicomponent thermohydrodynamic computer codes is the detailed study of hypothetical core disruptive accidents (HCDAs) in liquid-metal fast breeder reactors. The main contributions such codes are expected to make are the inclusion of detailed modeling of the relative motion of liquid and vapor (slip), the inclusion of modeling of nonequilibrium/nonsaturation thermodynamics, and the use of more detailed neutronics methods. Scoping studies of the importance of including these phenomena performed with the parametric two-field, two-component coupled neutronic/thermodynamic/hydrodynamic code FX2-TWOPOOL indicate for the prompt burst portion of an HCDA that: (1) Vapor-liquid slip plays a relatively insignificant role in establishing energetics, implying that analyses that do not model vapor-liquid slip may be adequate. Furthermore, if conditions of saturation are assumed to be maintained, calculations that do not permit vapor-liquid slip appear to be conservative. (2) The modeling of conduction-limited fuel vaporization and condensation causes the energetics to be highly sensitive to variations in the droplet size (i.e., in the parametric values) for the sizes of interest in HCDA analysis. Care must therefore be exercised in the inclusion of this phenomenon in energetics calculations. (3) Insignificant differences are observed between the use of space-time kinetics (quasi-static diffusion theory) and point kinetics, indicating again that point kinetics is normally adequate for analysis of the prompt burst portion of an HCDA. (4) No significant differences were found to result from assuming that delayed neutron precursors remain stationary where they are created rather than assuming that they move together with fuel. (5) There is no need for implicit coupling between the neutronics and the hydrodynamics/thermodynamics routines, even outside the prompt burst portion

Tidal disruption of fuzzy dark matter subhalo cores

Science.gov (United States)

Du, Xiaolong; Schwabe, Bodo; Niemeyer, Jens C.; Bürger, David

2018-03-01

We study tidal stripping of fuzzy dark matter (FDM) subhalo cores using simulations of the Schrödinger-Poisson equations and analyze the dynamics of tidal disruption, highlighting the differences with standard cold dark matter. Mass loss outside of the tidal radius forces the core to relax into a less compact configuration, lowering the tidal radius. As the characteristic radius of a solitonic core scales inversely with its mass, tidal stripping results in a runaway effect and rapid tidal disruption of the core once its central density drops below 4.5 times the average density of the host within the orbital radius. Additionally, we find that the core is deformed into a tidally locked ellipsoid with increasing eccentricities until it is completely disrupted. Using the core mass loss rate, we compute the minimum mass of cores that can survive several orbits for different FDM particle masses and compare it with observed masses of satellite galaxies in the Milky Way.

Recriticality, a Key Phenomenon to Investigate in Core Disruptive Accident Scenarios of Current and Future Fast Reactor Designs

International Nuclear Information System (INIS)

Maschek, W.; Rineiski, A.; Flad, M.; Kriventsev, V.; Gabrielli, F.; Morita, K.

2012-01-01

Final comments and conclusions: • Modern plants, should have performed better under Fukushima type event. • In future fast reactor systems significantly higher active and passive safety features are installed, which should cope with events like Fukushima. • One important lesson: put a focus on rare initiators, accident routes and consequences that are neither expected nor have been observed, events that are categorized under ‘black swans’. • Importance of severe accident research demonstrated - both analytically and experimentally for assessing and interpreting accident scenarios and developments. Precondition for developing preventive & mitigative safety measures. Passive safety measures are in the focus of advanced design options and must work under conditions of multiple loads and aggravating events. • Fast reactor systems behavior as the SFR under severe accident conditions: – In fast spectrum systems as the SFR the core is not in its neutronically most reactive configuration and SFRs may be loaded with MAs for waste management; – Recriticalities have a high probability because of the higher enrichment levels; – Short time scales have to be envisioned for core melt-down; – Decay heat levels might be significantly higher, if MA bearing fuel is involved. • Improve design by measures for prevention and/or mitigation of recriticalities; – High reliability of simulations required for proof; • Assessment of fuel relocated on peripheral structures; • Preventive/mitigating measures should not replace containment measures

Specialists' meeting on role of fission products in whole core accidents

Energy Technology Data Exchange (ETDEWEB)

NONE

1977-07-01

Safety of nuclear reactors has been a central concern of the nuclear energy industry from the very beginning. This concern, and the resultant excellence of design, fabrication, and operation, aided by extensive engineered safety features, has given nuclear energy its superior record of protection of the environment and of the public health and safety. With respect to the fast reactor, it was recognized early in the program that there exists a theoretical possibility of a core compaction leading to significant energy release. The considerations of fission product effects are primarily on of the main concerns in evaluation of safety issues. Since fission products have the potential for dispersing fuel from the core region and thereby producing reactor shutdown, knowledge of their effects can contribute to demonstrating that there is a low probability producing whole-core involvement. Similarly, knowledge of fission product effects can contribute to demonstrating that there is a low probability of a whole-core disruptive accident leading to sufficient energy release to challenge the containment capability.

Specialists' meeting on role of fission products in whole core accidents

International Nuclear Information System (INIS)

1977-01-01

Safety of nuclear reactors has been a central concern of the nuclear energy industry from the very beginning. This concern, and the resultant excellence of design, fabrication, and operation, aided by extensive engineered safety features, has given nuclear energy its superior record of protection of the environment and of the public health and safety. With respect to the fast reactor, it was recognized early in the program that there exists a theoretical possibility of a core compaction leading to significant energy release. The considerations of fission product effects are primarily on of the main concerns in evaluation of safety issues. Since fission products have the potential for dispersing fuel from the core region and thereby producing reactor shutdown, knowledge of their effects can contribute to demonstrating that there is a low probability producing whole-core involvement. Similarly, knowledge of fission product effects can contribute to demonstrating that there is a low probability of a whole-core disruptive accident leading to sufficient energy release to challenge the containment capability

Assessment of radiological impact due to a hypothetical core disruptive accident for PFBR using an advanced atmospheric dispersion system

International Nuclear Information System (INIS)

Srinivas, C.V.; Venkatesan, R.; Natarajan, A.

2004-01-01

Radiological impact due to air borne effluent dispersion from a hypothetical Core Disruptive Accident (CDA) scenario for Prototype Fast Breeder Reactor (PFBR) at Kalpakkam coastal site is estimated using an advanced system consisting of a 3-d meso-scale atmospheric model and a random walk particle dispersion model. A simulation of dispersion for CDA carried out for a typical summer day on 24th May 2003 predicted development of land-sea breeze circulation and Thermal Internal Boundary Layer (TIBL) at Kalpakkam site, which have been confirmed by observations. Analysis of dose distribution corresponding to predicted atmospheric conditions shows maximum dose from stack releases beyond the site boundary at about 4 km during TIBL fumigation and stable conditions respectively. A multi mode spatial concentration distribution has been noticed with diurnal meandering of wind under land sea breeze circulation. Over a meso-scale range of 25 km, turning of plume under sea breeze and maximum concentration along plume centerline at distances of 3 to 10 km have been noticed. The study has enabled to simulate the more complex meteorological situation that is actually present at the site. (author)

Comparative analysis of unprotected loss-of-flow accidents for the 1.0 m EFR-LVC core using different computer codes

International Nuclear Information System (INIS)

Royl, P.; Frizonnet, J.M.; Moran, J.

1993-02-01

A comparative analysis of the unprotected loss of flow (ULOF) accident has been performed for the LVC core (Lower Void Core) of the European Fast Reactor EFR with the FRAX5B and FRAX5C codes from the AEA-T, the PHYSURAC code from CEA and the SAS4A REF92 code system developed jointly between KfK, CEA and PNC. The accident is triggered by the run down of the coolant pumps with failure to trip the reactor by the primary and/or secondary shutdown system. Only a limited amount of mitigating reactivity from the third shutdown line was considered so that the accident can progress into boiling and core disruption. This code outlines the important modelling differences and compares the different simulations. The discussion of the rather wide spectrum of calculated accident progressions identifies the generic differences, relates them to the applied models, and summarizes the key points that are responsible for the different progressions. A comparison of the consequence spectrum from all simulations indicates zero work energies for the majority of the calculations. All simulations show up the need for a continued accident analysis into the early and late transition phase

SIMMER-II: A computer program for LMFBR disrupted core analysis

Energy Technology Data Exchange (ETDEWEB)

Bohl, W.R.; Luck, L.B.

1990-06-01

SIMMER-2 (Version 12) is a computer program to predict the coupled neutronic and fluid-dynamics behavior of liquid-metal fast reactors during core-disruptive accident transients. The modeling philosophy is based on the use of general, but approximate, physics to represent interactions of accident phenomena and regimes rather than a detailed representation of specialized situations. Reactor neutronic behavior is predicted by solving space (r,z), energy, and time-dependent neutron conservation equations (discrete ordinates transport or diffusion). The neutronics and the fluid dynamics are coupled via temperature- and background-dependent cross sections and the reactor power distribution. The fluid-dynamics calculation solves multicomponent, multiphase, multifield equations for mass, momentum, and energy conservation in (r,z) or (x,y) geometry. A structure field with nine density and five energy components; a liquid field with eight density and six energy components; and a vapor field with six density and on energy component are coupled by exchange functions representing a modified-dispersed flow regime with a zero-dimensional intra-cell structure model.

SIMMER-II: A computer program for LMFBR disrupted core analysis

International Nuclear Information System (INIS)

Bohl, W.R.; Luck, L.B.

1990-06-01

SIMMER-2 (Version 12) is a computer program to predict the coupled neutronic and fluid-dynamics behavior of liquid-metal fast reactors during core-disruptive accident transients. The modeling philosophy is based on the use of general, but approximate, physics to represent interactions of accident phenomena and regimes rather than a detailed representation of specialized situations. Reactor neutronic behavior is predicted by solving space (r,z), energy, and time-dependent neutron conservation equations (discrete ordinates transport or diffusion). The neutronics and the fluid dynamics are coupled via temperature- and background-dependent cross sections and the reactor power distribution. The fluid-dynamics calculation solves multicomponent, multiphase, multifield equations for mass, momentum, and energy conservation in (r,z) or (x,y) geometry. A structure field with nine density and five energy components; a liquid field with eight density and six energy components; and a vapor field with six density and on energy component are coupled by exchange functions representing a modified-dispersed flow regime with a zero-dimensional intra-cell structure model

Core fusion accidents in nuclear power reactors. Knowledge review

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

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

An assessment of the effect of reactor size on hypothetical ore disruptive accidents

International Nuclear Information System (INIS)

Buttery, N.E.; Board, S.J.

1978-01-01

There is a general tendency towards larger plant sizes, in the interests primarily of economies of scale. In this paper the effect of core size on hypothetical core disruptive accidents (HCDA) is considered, and it is shown that the energy yield increases rapidly with size, primarily due to a tendency towards coherence of voiding in reactors with a large positive void coefficient. Small cores compare favourably in this respect with alternative large designs with low void coefficient cores, because the reduced mass more than compensates for the reduced doppler constant, and they also have a potential advantage in later stages of HCDA (transition phase and after). If energetic HCDA cannot be shown to be unrealistic and if containment of these events is provided as part of the general safety philosophy, then the costs (which may increase disproportionately with yield) of engineering an adequately reliable system needs to be accounted for. Containment costs are only one of many factors which need to be taken into account in optimising the design and so the energy release from a HCDA must take its proper place in the optimisation according to the safety principles and safety case agreed for LMFBRS. (author)

Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

Energy Technology Data Exchange (ETDEWEB)

Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

feedback (Doppler and displacement) VENUS-II has the following assumptions. The reactor materials behave like a homogeneous mixture with the property of an isotropic and nonviscous fluid. The reactivity change caused by a material displacement can be calculated with first-order perturbation theory. Further, the reactivity worth of spatial gradients remain constant and distort with the grid. The heat transfer from the fuel can be ignored. Although several heat transfer mechanisms can become significant, one of the greatest potential influence would appear to be a rapid molten-fuel-coolant interaction (MFCI). The non fuel core constituents are considered to be compressible, but inert, materials. The fuel vapor pressure and compression of the reactor materials are the only sources of internal pressure. Thus, such potential pressure sources such as fission gas and sodium vapor pressure are ignored. The time history of the power level can be described using point kinetics, and the spatial power-density distribution remains constant. In this work, the energy released from core disruptive accident (CDA) of sodium cooled fast reactor was investigated using CDA-ER and VENUS-II code for various reactivity insertion rates up to 100$/s, and their results were compared. The calculation results of two codes showed similar trends of energy, power and pressure from CDA. But most results of VENUS-II were found to be larger than those of CDA-ER. The released energy results calculated from VENUS-II were about 2 ∼ 3 times higher than those from CDA-ER.

Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

International Nuclear Information System (INIS)

Kang, S. H.; Ha, K. S.

2013-01-01

displacement) VENUS-II has the following assumptions. The reactor materials behave like a homogeneous mixture with the property of an isotropic and nonviscous fluid. The reactivity change caused by a material displacement can be calculated with first-order perturbation theory. Further, the reactivity worth of spatial gradients remain constant and distort with the grid. The heat transfer from the fuel can be ignored. Although several heat transfer mechanisms can become significant, one of the greatest potential influence would appear to be a rapid molten-fuel-coolant interaction (MFCI). The non fuel core constituents are considered to be compressible, but inert, materials. The fuel vapor pressure and compression of the reactor materials are the only sources of internal pressure. Thus, such potential pressure sources such as fission gas and sodium vapor pressure are ignored. The time history of the power level can be described using point kinetics, and the spatial power-density distribution remains constant. In this work, the energy released from core disruptive accident (CDA) of sodium cooled fast reactor was investigated using CDA-ER and VENUS-II code for various reactivity insertion rates up to 100$/s, and their results were compared. The calculation results of two codes showed similar trends of energy, power and pressure from CDA. But most results of VENUS-II were found to be larger than those of CDA-ER. The released energy results calculated from VENUS-II were about 2 ∼ 3 times higher than those from CDA-ER

Consequence analysis of core meltdown accidents in liquid metal fast reactor

International Nuclear Information System (INIS)

Suk, S.D.; Hahn, D.

2001-01-01

Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

International Nuclear Information System (INIS)

Albrecht, H.; Malinauskas, A.P.

1978-01-01

A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

2010-01-01

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

Nuclear power reactor core melt accidents. Current State of Knowledge

International Nuclear Information System (INIS)

Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

2015-11-01

For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt

Implications for accident management of adding water to a degrading reactor core

International Nuclear Information System (INIS)

Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J.

1994-02-01

This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents

Implications for accident management of adding water to a degrading reactor core

Energy Technology Data Exchange (ETDEWEB)

Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-02-01

This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

Structural and containment response to LMFBR accidents

International Nuclear Information System (INIS)

Marchaterre, J.F.; Fistedis, S.H.; Baker, L. Jr.; Stepnewski, D.D.; Peak, R.D.; Gluekler, E.L.

1978-01-01

The results of current developments in analysing the response of reactor structures and containment to LMFBR accidents are presented. The current status of analysis of the structural response of LMFBR's to core disruptive accidents, including head response, potential missile generation and the effects of internal structures are presented. The results of recent experiments to help clarify the thermal response of reactor structures to molten core debris are summarized, including the use of this data to calculate the response of the secondary containment. (author)

Core loss during a severe accident (COLOSS)

International Nuclear Information System (INIS)

Adroguer, B.; Bertrand, F.; Chatelard, P.; Cocuaud, N.; Van Dorsselaere, J.P.; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtilkova, V.; Belovsky, L.; Mueller, K.; Hering, W.; Homann, C.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrueck, M.; Stuckert, J.; Hozer, Z.; Bandini, G.; Birchley, J.; Berlepsch, T. von; Kleinhietpass, I.; Buck, M.; Benitez, J.A.F.; Virtanen, E.; Marguet, S.; Azarian, G.; Caillaux, A.; Plank, H.; Boldyrev, A.; Veshchunov, M.; Kobzar, V.; Zvonarev, Y.; Goryachev, A.

2005-01-01

The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy (b) simultaneous dissolution of UO 2 and ZrO 2 (c) oxidation of U-O-Zr mixtures (d) degradation-oxidation of B 4 C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4 C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Breakthroughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R and D activities are summarized in this paper

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium cooled fast reactors

International Nuclear Information System (INIS)

Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, Liancheng; Zhang, Bin

2014-01-01

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

International Nuclear Information System (INIS)

Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, LianCheng; Zhang, Bin

2016-01-01

The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

High enrichment to low enrichment core's conversion. Accidents analysis

International Nuclear Information System (INIS)

Abbate, P.; Rubio, R.; Doval, A.; Lovotti, O.

1990-01-01

This work analyzes the different accidents that may occur in the reactor's facility after the 20% high-enriched uranium core's conversion. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. This analysis includes: a) accidents by reactivity insertion; b) accidents by coolant loss; c) analysis by flow loss and d) fission products release. (Author) [es

Role of fission product in whole core accidents: research in the USA

International Nuclear Information System (INIS)

Jackson, J.F.; Deitrich, L.W.

1977-01-01

The techniques being developed in the United States for analyzing postulated whole-core accidents in LMFBRs are briefly reviewed. The key mechanistic analysis methods are discussed in detail. Important research projects in the area of fission product effects are examined. Some typical results on the role of fission products in whole-core accidents are presented

RBMK-1500 accident management for loss of long-term core cooling

International Nuclear Information System (INIS)

Uspuras, E.; Kaliatka, A.

2001-01-01

Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

analysis of reactivity accidents in MTR for various protection system parameters and core condition

International Nuclear Information System (INIS)

Mohamed, F.M.

2011-01-01

Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

How to arrest a core meltdown accident (doing nothing)

International Nuclear Information System (INIS)

Baron, Jorge H.

2000-01-01

In the eventual situation of a severe accident in a nuclear reactor, the molten core is able to relocate inside the pressure vessel. This may lead to the vessel failure, due to the thermal attack of the molten core (at approximation of 3000K) on the vessel steel wall. The vessel failure implies the failure of a very important barrier that contains the radioactive materials generated during the reactor operation, with a significant risk of producing high radiation doses both on operators and on the public. It is expected, for the new generation of nuclear reactors, that these will be required to withstand (by design) a core melt down accident, without the need for an immediate evacuation of the surrounding population. In this line, the use of a totally passive system is postulated, which fulfills the objective of containing the molten core inside the pressure vessel, at low temperature (approximation 1200K) precluding its failure. The conceptual design of a passive in-vessel core catcher is presented in this paper, built up of zinc, and designed for the CAREM-25 nuclear power plant. (author)

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

Large population center and core melt accident considerations in siting

International Nuclear Information System (INIS)

Camarinopoulos, L.; Yadigaroglu, G.

1983-01-01

The problem of providing suitable demographic siting criteria in the presence of a very large population center in an otherwise sparsely populated region is addressed. Simple calculations were performed making maximum use of pretabulated results of studies where core melt accidents are considered. These show that taking into consideration the air flow patterns in the region can lower the expected population doses from core melt accidents more effectively than distance alone. Expected doses are compared to the annual background radiation dose. A simple siting criterion combining geographical considerations with the probability of a release reaching the large population center is proposed

Fuel-disruption experiments under high-ramp-rate heating conditions

International Nuclear Information System (INIS)

Wright, S.A.; Worledge, D.H.; Cano, G.L.; Mast, P.K.; Briscoe, F.

1983-10-01

This topical report presents the preliminary results and analysis of the High Ramp Rate fuel-disruption experiment series. These experiments were performed in the Annular Core Research Reactor at Sandia National Laboratories to investigate the timing and mode of fuel disruption during the prompt-burst phase of a loss-of-flow accident. High-speed cinematography was used to observe the timing and mode of the fuel disruption in a stack of five fuel pellets. Of the four experiments discussed, one used fresh mixed-oxide fuel, and three used irradiated mixed-oxide fuel. Analysis of the experiments indicates that in all cases, the observed disruption occurred well before fuel-vapor pressure was high enough to cause the disruption. The disruption appeared as a rapid spray-like expansion and occurred near the onset of fuel melting in the irradiated-fuel experiments and near the time of complete fuel melting in the fresh-fuel experiment. This early occurrence of fuel disruption is significant because it can potentially lower the work-energy release resulting from a prompt-burst disassembly accident

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

Structural assessment of TAPS core shroud under accident loads

International Nuclear Information System (INIS)

Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1996-09-01

Over the last few years, the Core Shroud of Boiling Water Reactors (BWRs) operating in foreign countries, have developed cracks at weld locations. As a first step for assessment of structural safety of Tarapur Atomic Power Station (TAPS) core shroud, its detailed stress analysis was done for postulated accident loads. This report is concerned with structural assessment of core shroud, of BWR at TAPS, subjected to loads resulting from main steam line break (MSLB), recirculation line break (RLB) and safe shut down earthquake. The stress analysis was done for core shroud in healthy condition and without any crack since, visual examination conducted till now, do not indicate presence of any flaw. Dynamic structural analysis for MSLB and RLB events was done using dynamic load factor (DLF) method. The complete core shroud and its associated components were modelled and analysed using 3D plate/shell elements. Since, the components of core shroud are submerged in water, hence, hydrodynamic added mass was also considered for evaluation of natural frequencies. It was concluded that from structural point of view, adequate safety margin is available under all the accident loads. Nonlinear analysis was done to evaluate buckling/collapse load. The collapse/buckling load have sufficient margin against the allowable limits. The displacements are low hence, the insertion of control rod may not be affected. (author)

Analysis of space-time core dynamics on reactor accident at Chernobyl

International Nuclear Information System (INIS)

Takano, Makoto; Shindo, Ryuichi; Yamashita, Kiyonobu; Sawa, Kazuhiro

1987-05-01

Regarding reactor accident at Chernobyl in USSR, core dynamics has been analyzed by COMIC code which solves space-time dependent diffusion equation in three-dimension taking spatial thermohydraulic effect into account. The code was originally developed for high temperature gas-cooled reactors (HTGR), however, has been modified to include light water as coolant, instead of helium, for analysis of the accident. In the analysis, emphasis is placed on spatial effects on core dynamics. The analyses are performed for the cases of modeling the core fully and partially where 6 fuel channels surround one control rod channel. The result shows that the speed of applying void reactivity averaged over the core depends on the power and coolant flow distributions. Therefore, these distributions have potential to influence on the value and the time of peak power estimated by calculation. (author)

Core-melting accidents in Chernobyl and Harrisburg

International Nuclear Information System (INIS)

Loon, A.J. van; Vonderen, A.C.M. van

1987-01-01

This publication deals with the essences of the reactor accident in Chernobylsk and the conclusions to be drawn from these with regard to reactor safety. Therein the technical differences between the reactor types in the West and the East play an important role. Also attention is spent to the now generally accepted philosophy that by simplification and making use of proven technologies, a further deminishing of the risks can be achieved step by step. In ch.'s 2 and 4 the origin and course of the accidents in respectively Chernobylsk and Harrisburg are analyzed; in the analysis of the Chernobylsk accident also date have been used which were provided by the Sovjet-Union, supplied with results of studies of the U.S. Department of Energy (DOE). In ch. 3 this information is compared with the insights which have grown at KEMA about these on the base of reactor physical and thermohydraulic considerations and of computer calculations reproducing the course of the accident. An important question is if, and if so: to which extent, an accident such as the one in Chernobylsk also can take place in the West. In order to answer that question as accurate as possible the consequences of core meltings accidents and the risk for such an accident taking place are pursued. In ch. 6 the legal frameworks are indicated by which the risk may be limited and by which eventually yet occurring damage may be arranged. Ch. 7 finally deals with the lessons which the accidents in Chernobylsk and Harrisburg have learnt us and with the possible consequences of these for the further application of nuclear power in the Netherlands. (H.W.). 105 refs.; 42 figs.; 17 refs

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

Core failure accident pathways and ways to control it

International Nuclear Information System (INIS)

Mayinger, F.

1982-01-01

In the German Risk Study accidents are assumed to result in core meltdown whenever the criteria spelt out in the guidelines of the Advisory Committee on Reactor Safeguards are no longer met. This assumption must be seen in the light of an earlier state of the art in which no detailed information could be obtained about intermediate stages in emergency core cooling systems working according to permit up to the complete failure of all heat removal systems. However, experimental studies and theoretical analyses conducted over the past few years have advanced the state of the art such that it is now possible to predict with considerably more physical reality the behavior of a core in a loss-of-coolant accident. These findings are not only based on calculations, but also on the results of experiments in large facilities allowing direct comparisons to be made with conditions in nuclear power plants. Studies of the effects of systems failures both in major leakages and in the small leakages regarded to be much more dangerous show much more favorable conditions with respect to core coolability than had to be anticipated on the basis of earlier assumptions. This also implies that it would neither be necessary nor meaningful to reinforce emergency core cooling systems. Instead, it is much more important, besides having technically highly qualified and thoroughly trained operating crews, to inform those crews reliably of the hydrodynamic and thermodynamic state of the primary system, especially the core. (orig.) [de

Examination of offsite emergency protective measures for core melt accidents

International Nuclear Information System (INIS)

Aldrich, D.C.; McGrath, P.E.; Ericson, D.M. Jr.; Jones, R.B.; Rasmussen, N.C.

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

Development of a DNBR evaluation method for the CEA ejection accident in SMART core

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Yoo, Y. J.; In, W. K.; Chang, M. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

A methodology applicable to the analysis of the CEA ejection accident in SMART is developed for the evaluation of the fraction of fuel failure caused by DNB. The transient behavior of the core thermal-hydraulic conditions is calculated by the subchannel analysis code MATRA. The minimum DNBR during the accident is calculated by KRB-1 CHF correlation considering the 1/8 symmetry of hot assembly. The variation of hot assembly power during the accident is simulated by the LTC(Limiting transient Curve) which is determined from the analysis of power distribution data resulting from the three-dimensional core dynamics calculations. The initial condition of the accident is determined by considering LOC(Limiting Conditions for Operation) of SMART core. Two different methodologies for the evaluation of DNB failure rate are established; a deterministic method based on the DNB envelope, and a probabilistic method based on the DNB probability of each fuel rod. The methodology developed in this study is applied to the analysis of CEA ejection accident in the preliminary design core of SMART. As the result, the fractions of DNB fuel failure by the deterministic method and the probabilistic method are calculated as 38.7% and 7.8%, respectively. 16 refs., 16 figs., 5 tabs. (Author)

Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

Effect of engineered safety features on the risk of hypothetical LMFBR accidents

International Nuclear Information System (INIS)

Cybulskis, P.

1978-01-01

The risks of hypothetical core-disruptive accidents in liquid-metal-cooled fast breeder reactors which involve meltthrough of the reactor vessel are compared for two plant designs: one design without specific provisions to accommodate such an accident and the other design with an ex-vessel core catcher and a cvity hot liner. The approach to risk analysis used is that developed in the Reactor Safety Study (WASH-1400). Since the probability of occurrence of such an event has not been evaluated, however, insight into the potential risk is gained only on a relative basis. The principal conclusions of this study are: (1) adding a core catcher--hot liner reduces the probabilty of accidents having major consequences; (2) the degree to which hot liner--core catcher systems can reduce the risk of melt-through accidents is limited by the failure probability of these systems; (3) fractional radioactive releases to the environment in the liquid-metal-cooled fast breeder reactor accidents considered are comparable to those from the light-water reactors evaluated in WASH-1400; (4) since sodium--concrete reactions are a dominant driving force during the accident, the integrity of the cavity liner is as important as the function of the core catcher; (5) there may be other accidents or paths to radioactive releases that are not affected by the addition of a hot liner--core catcher

Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report. Vol. 1

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

This report gives the results of a study of the thermo-hydraulic aspects of severe accident sequences in CANDU reactors. The accident sequences considered are the loss of the moderator cooling system and the loss of the moderator heat sink, each following a large loss-of-coolant accident accompanied by loss of emergency coolant injection. Factors considered include expulsion and boil-off of the moderator, uncovery, overheating and disintegration of the fuel channels, quenching of channel debris, re-heating of channel debris following complete moderator expulsion, formation and possible boiling of a molten pool of core debris and the effectiveness of the cooling of the calandria wall by the shield tank water during the accident sequences. The effects of these accident sequences on the reactor containment are also considered. Results show that there would be no gross melting of fuel during moderator expulsion from the calandria, and for a considerable time thereafter, as quenched core debris re-heats. Core melting would not begin until about 135 minutes after accident initiation in a loss of the moderator cooling system and until about 30 minutes in a loss of the moderator heat sink. Eventually, a pool of molten material would form in the bottom of the calandria, which may or may not boil, depending on property values. In all cases, the molten core would be contained within the calandria, as long as the shield tank water cooling system remains operational. Finally, in the period from 8 to 50 hours after the initiation of the accident, the molten core would re-solidify within the calandria. There would be no consequent damage to containment resulting from these accident sequences, nor would there be a significant increase in fission product releases from containment above those that would otherwise occur in a dual failure LOCA plus LOECI

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Zircaloy-oxidation and hydrogen-generation rates in degraded-core accident situations

International Nuclear Information System (INIS)

Chung, H.M.; Thomas, G.R.

1983-02-01

Oxidation of Zircaloy cladding is the primary source of hydrogen generated during a degraded-core accident. In this paper, reported Zircaloy oxidation rates, either measured at 1500 to 1850 0 C or extrapolated from the low-temperature data obtained at 0 C, are critically reviewed with respect to their applicability to a degraded-core accident situation in which the high-temperature fuel cladding is likely to be exposed to and oxidized in mixtures of hydrogen and depleted steam, rather than in an unlimited flux of pure steam. New results of Zircaloy oxidation measurements in various mixtures of hydrogen and steam are reported for >1500 0 C. The results show significantly smaller oxidation and, hence, hydrogen-generation rates in the mixture, compared with those obtained in pure steam. It is also shown that a significant fraction of hydrogen, generated as a result of Zircaloy oxidation, is dissolved in the cladding material itself, which prevents that portion of hydrogen from reaching the containment building space. Implications of these findings are discussed in relation to a more realistic method of quantifying the hydrogen source term for a degraded-core accident analysis

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

International Nuclear Information System (INIS)

Martz, H.F.

1998-01-01

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

Behaviour of LWR core materials under accident conditions. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1996-12-01

At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs

Considerations on the influence of fission products in whole core accidents

International Nuclear Information System (INIS)

Meyer Heine, A.; Pattoret, A.; Schmitz, F.

1977-01-01

If the hypothetical Whole Core Accidents which are taken into account in reactor safety analysis can change from one country to another, there is nevertheless a general agreement over their description and main phases. Furthermore the important parameters have also been identified by every laboratory. During the development of such core accidents the role of the fission products in essential. It is not the purpose of this paper to give an exhaustive description of the phases which can be influenced by the fission products, we will try however to focus this study on the most important ones. In a second step we will discuss the equation of state of irradiated fuels; here again one principal preoccupation being to quantify the influence of fission products on reactor accidents. It is not our purpose to enter on the fundamental aspects of the equation of state. The studies and the experimental program launched at the CEA will then be described. Special attention will be directed towards the eventual role of fission products in molten fuel-coolant interactions (MFCls) or the events leading to the initiation of whole core accidents. This paper will be limited to oxide fuels. Whether the whole core accident is initiated by a reactivity defect or a coolant coast-down, one has to deal with four great categories of phenomena. Loss of flow: the power is around the nominal value, while the coolant flow has been reduced by a factor of 5 to 10. This induces boiling and clad weakening. Will the plenum pressure lead to a clad rupture? In case of a rupture, what will be the effect on the voiding of the channel? Transient over power: influence of gases from gaseous and volatile fission products on the fuel movements? MFCIs: Influence of the fission products in the mode of contact between fuel and coolant? Influence on the fuel characteristics. Sodium vapour bubble expansion: influence of the fission products on the heat transfer and eventual condensation of the bubble?

Fission-gas bubble modeling for LMFBR accidents

International Nuclear Information System (INIS)

Ostensen, R.W.

1977-01-01

The behavior of fission-gas bubbles in unrestructured oxide fuel can have a dominant effect on the course of a core disruptive accident in an LMFBR. The paper describes a simplified model of bubble behavior and presents results of that model in analyzing the relevant physical assumptions and predicting gas behavior in molten fuel

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

International Nuclear Information System (INIS)

Malinauskas, A.P.

1989-01-01

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

The Chernobyl accident: Causes and consequences

International Nuclear Information System (INIS)

Malinauskas, A.P.

1987-01-01

Two explosions, one immediately following the other, in Unit 4 of the Chernobyl nuclear power station in the Soviet Union signaled the worst disaster ever to befall the commercial nuclear power production industry. This accident, which occurred at 1:24 a.m. on April 26, 1986, resulted from an almost incredible series of operational errors associated, ironically, with an attempt to enhance the capability of the reactor to safely accommodate station blackout accidents (i.e., accidents arising from a loss of station electrical power). Disruption of the core, due to a prompt criticality excursion, resulted in the destruction of the core vault and reactor building and the sudden dispersal of about 3% of the fuel from the core region into the environment. Lesser but significant releases of radioactivity continued through May 6, 1986, before attempts to certain the radioactivity and cool the remnants of the core were successful. The amount and composition of material released in the course of the accident remain somewhat uncertain, and inconsistencies in the release estimates are evident. The Soviet estimates, in addition to the dispersal of about 3% of the fuel, include complete release of the noble gas core inventory, 20% of the fission product iodine inventory, 15% of the tellurium inventory, and 10 to 13% of the fission product cesium inventory. The iodine and cesium release estimates are not consistent with the noble gas values, and are as much as a factor of two less than some estimates made by experts outside the Soviet Union

Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

International Nuclear Information System (INIS)

Kroeger, P.G.; Colman, J.; Hsu, C.J.

1983-01-01

In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

Analysis of hypothetical LMFBR whole-core accidents in the USA

International Nuclear Information System (INIS)

Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

1978-01-01

The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current generation of fast reactors. An assessment of future trends in this area concludes the paper

The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

CERN Document Server

Kelly, G N

1983-01-01

The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

Qualification testing program plan for SIMMER. A computer code for LMFBR disrupted core analysis

International Nuclear Information System (INIS)

Basdekas, D.L.; Silberberg, M.; Curtis, R.T.; Kelber, C.N.

1978-07-01

The objective of SIMMER qualification testing program is to assure that the mathematical models and input parameters are derived from experimental data, which, on the basis of criteria still to be established, are representative of the phenomena and processes governing the progression of a CDA in an LMFBR. At the present time, the work to meet this objective can be classified into three general task areas as they relate to the use of SIMMER in CDA analysis: (1) The whole-core energetic disassembly accident, or the ''vessel problem'': The objective here is to predict the partition of the total energy release, by a postulated severe power excursion, between the primary containment and the energy absorbed through nondestructive dissipative processes. (2) Single subassembly accident: The objective here is to determine the pertinent phenomena and to develop the capability to assess the significance and likelihood that such accidents might propagate to involvement of larger fraction of the core. (3) The whole-core transition phase accident: The objective here is to advance the understanding of the phenomena and processes involved, so that reliable predictions can be made of the possible consequences of a CDA and the potential for further nuclear excursions through recriticality

Detonability of containment building atmospheres during core-meltdown accidents

International Nuclear Information System (INIS)

Jaung, R.; Berlad, L.; Pratt, W.

1983-01-01

During Core-Meltdown Accidents in Light Water Reactors, significant quantities of combustible gases could be released to the containment building. The highest possible peak pressure fields that may occur through combustion processes are associated with detonation phenomena. Accordingly, it is necessary to understand and identify the possible ways in which detonations may or may not occur. Although no comprehensive theory of detonation is currently available, there are useful guidelines, which can be derived from current theoretical concepts and the body of experimental data. This paper examines these guidelines and indicates how they may be used to evaluate the possible occurrence of detonation-related combustion processes. In particular, this study identifies three features that an initiation source must achieve if it is to ultimately result in a stable detonation. One of these features requires post-shock initial conditions that lead to very short ignition delays. This concept is used to examine the possibility of achieving quasi-steady detonation phenomena in nuclear reactor containment buildings during postulated core-melt accidents

The role of fission products in whole core accidents

Energy Technology Data Exchange (ETDEWEB)

Baker, A R [FRSD, UKAEA, RNPDE, Risley, Warrington (United Kingdom); Teague, H J [SRD, UKAEA, Culcheth, Warrington (United Kingdom)

1977-07-01

The review of the role of fission products in whole-core accidents falls into two parts. Firstly, there is a discussion of the hypothetical accidents usually considered in the UK and how they are dealt with. Secondly, there is a discussion of individual topics where fission products are known to be important or might be so. There is a brief discussion of the UK work on the establishment of an equation of state for unirradiated fuel and how this might be extended to incorporate fission product effects. The main issue is the contribution of fission products to the effective vapour pressure and the experimental programme on the pulsed reactor VIPER investigates this. Fission products may influence the probability of occurrence and the severity of MFCIs. Finally, the fission product effects in the pre-disassembly, disassembly and recriticality stages of an accident are discussed. (author)

Comparative study of heterogeneous and homogeneous LMFBR cores in some accident conditions

International Nuclear Information System (INIS)

Renard, A.; Evrard, G.

1978-01-01

An heterogeneous design and a homogeneous one of a LMFBR core with the same power and similar dimensions are compared from the safety point-of-view. The comparison is performed for several accident conditions, such as Loss-of-Flow and Transient Overpower, with the same failure criteria and model assumptions for both cores. Qualitative trends are deduced from the behaviour of the core designs in the investigated transient conditions. (author)

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Fuel relocation modeling in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

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

1985-01-01

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

Fuel relocation modeling in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

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

1986-01-01

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

Mitigation of Severe Accident Consequences Using Inherent Safety Principles

International Nuclear Information System (INIS)

Wigeland, R.A.; Cahalan, J.E.

2009-01-01

Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. For severe accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection system so that the probability of an ATWS event is reduced to less than 1 x 10-6 per reactor year, where larger accident consequences are allowed, meeting the U.S. NRC goal of relegating such accident consequences as core disruption to these extremely low probabilities. The main difficulty with this approach is to convincingly test and guarantee such increased reliability. Another approach is to increase the redundancy of the reactor scram system, which can also reduce the probability of an ATWS event to a frequency of less than 1 x 10-6 per reactor year or lower. The issues with this approach are more related to reactor core design, with the need for a greater number of control rod positions in the reactor core and the associated increase in complexity of the reactor protection system. A third approach is to use the inherent reactivity feedback that occurs in a fast reactor to

Accidents and transients analyses of a super fast reactor with single flow pass core

International Nuclear Information System (INIS)

Sutanto,; Oka, Yoshiaki

2014-01-01

Highlights: • Safety analysis of a Super FR with single flow pass core is conducted. • Loss of feed water flow leads to a direct effect on the loss of fuel channel flow. • The core pressure is sensitive to LOCA accidents due to the direct effect. • Small LOCA introduces a critical break. • The safety criteria for all selected events are satisfied. - Abstract: The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST

An assessment of core wide coherency effects in the multichannel modeling of the initiating phase of a severe accident in a sodium fast reactor

International Nuclear Information System (INIS)

Guyot, M.; Gubernatis, P.; Suteau, C.; Le Tellier, R.; Lecerf, J.

2014-01-01

To consolidate the safety assessment for liquid-metal fast breeder reactors (LMFBRs), hypothetical core disruptive accident (HCDA) sequences have been extensively studied over the past decades. Numerous analyses of the so called initiating phase (or primary phase) of a HCDA have been made with the safety analysis system code SAS4A. The SAS4A accident analysis code requires that subassemblies or groups of subassemblies be represented together as independent channels. For simulating a severe accident sequence, a subassembly-to-channel assignment procedure has to be implemented to produce the consistent SAS4A input decks. Generally, one uses imposed criteria over relevant reactor parameters to determine the subassembly to- channel arrangement. The multiple-assembly-per-channel approach introduces core wide coherency effects, which can affect the reactivity balance and therefore the overall accident development. In this paper, a subassembly-to channel assignment procedure based on the subassembly power-to-flow ratio is presented and implemented to generate the SAS4A input decks over a range of parameter values. The corresponding SAS4A calculations have been performed on a large LMFBR. The purpose of the present series of calculations is to investigate the magnitude of errors encountered in the analysis of the initiating phase related to the subassembly-to-channel arrangement selection, by comparison with a one-subassembly-per-channel reference solution. It appears that a refinement in the channel arrangement substantially reduces core wide coherency effects. Analysis of the calculations also suggests that an accurate representation of the scenario requires the number of channels to be on approximately the same order of magnitude as the total number of subassemblies. Numerical results are examined to provide the reader with quantitative measurements of bias related to subassembly to- channel arrangement. (authors)

MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

International Nuclear Information System (INIS)

Ball, S.J.

1991-10-01

The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR

Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt

International Nuclear Information System (INIS)

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

1978-06-01

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Dynamic response of cylindrical ACS support structures to core energy release

International Nuclear Information System (INIS)

Kennedy, J.M.; Belytschko, T.B.

1985-01-01

The code SAFE/RAS is applied to the analysis of a new design concept for the above-core structures when subjected to the loads of a core disruptive accident. The analysis involves the determination of the postbuckling response of a thin cylinder loaded both axially and vertically. The effects of variation of cylinder thickness and fluid-structure interaction are investigated

Licensing decisions and safety research related to LMFBR accidents

International Nuclear Information System (INIS)

Denise, R.P.; Speis, T.P.; Kelber, C.N.; Curtis, R.T.

1977-01-01

The licensing approach which ensures adequate protection of the public health and safety against serious accidents is described. This paper describes the role of core melt and core disruptive accidents in the design, safety research, and licensing processes, using the Clinch River Breeder Reactor (CRBR) as a focal point. Major design attention is placed on the prevention of these accidents so that the probability of core melt accidents is reduced to a sufficiently low level that they are not treated as design basis accidents. Additional requirements are placed upon the design to further reduce residual risk. This licensing process is supported by a confirmatory research program designed to provide an independent basis for licensing judgements. It has as a goal the resolution of generic safety issues prior to the establishment of a commercial LMFBR industry. The program includes accident analysis, experiments in materials interactions, aerosol transport and system integrity and planning for new safety test facilities. The problems are approached in a multi-disciplinary functional manner that identifies key safety issues and centralizes efforts to resolve them. The near term objectives of the program support the licensing of the Clinch River Breeder Reactor (CRBR) and the proposed Prototype Large Breeder Reactor (PLBR). The long term objectives of the program support the licensing of commercial LMFBRs during the late 1980's and beyond. This safety research is designed to provide an independent basis for the licensing judgements which must be made by the Nuclear Regulatory Commission

Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

International Nuclear Information System (INIS)

Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

1996-09-01

The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-11-01

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

Simulation of heat and mass transfer processes in molten core debris-concrete systems. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Felde, D K

1979-01-01

The heat and mass transport phenomena taking place in volumetrically-heated fluids have become of interest in recent years due to their significance in assessments of fast reactor safety and post-accident heat removal (PAHR). Following a hypothetical core disruptive accident (HCDA), the core and reactor internals may melt down. The core debis melting through the reactor vessel and guard vessel may eventually contact the concrete of the reactor cell floor. The interaction of the core debris with the concrete as well as the melting of the debris pool into the concrete will significantly affect efforts to prevent breaching of the containment and the resultant release of radioactive effluents to the environment.

Severe accident mitigation and core melt retention in the European pressurized reactor (EPR)

International Nuclear Information System (INIS)

Fischer, Manfred

2003-01-01

For the mitigation of severe accidents, the FPR has adopted and improved the defense-in-depth approaches of its predecessors, the French 'N4' and the German 'Konvoi' PWR's. Beyond these evolutionary changes, it includes a new, 4-th level of defense aimed at limiting the consequences of a postulated severe accident with core melting. This involves a strengthening of the confinement function and the avoidance of large early releases, by the prevention of scenarios and events with potentially high loads on the containment, incl. RPV failure at high pressure. The remaining low-pressure accidents are mitigated by dedicated design measures. The paper gives an overview and of the measures for H 2 -mitigation and steam explosion and focuses on a detailed description of the precautions and design measures for the stabilization and long-term cooling of the molten core. In the EPR the latter is achieved by melt spreading into a large outside-cooled crucible lateral to the pit, which is passively flooded and cooled with water from the IRWST. The separation of functions between pit and spreading room not only isolates the core catcher from the various loads during RPV failure, but also avoids any risks related to an unintended initiation of flooding during power operation. A stable state of the melt is reached after a few hours. Complete solidification is achieved within days. The core catcher can optionally be cooled actively by the CHRS, which avoids further steaming into the containment and establishes ambient pressure conditions in the long term. (author)

Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-03-01

JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-08-01

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

Valproic acid disrupts the oscillatory expression of core circadian rhythm transcription factors.

Science.gov (United States)

Griggs, Chanel A; Malm, Scott W; Jaime-Frias, Rosa; Smith, Catharine L

2018-01-15

Valproic acid (VPA) is a well-established therapeutic used in treatment of seizure and mood disorders as well as migraines and a known hepatotoxicant. About 50% of VPA users experience metabolic disruptions, including weight gain, hyperlipidemia, and hyperinsulinemia, among others. Several of these metabolic abnormalities are similar to the effects of circadian rhythm disruption. In the current study, we examine the effect of VPA exposure on the expression of core circadian transcription factors that drive the circadian clock via a transcription-translation feedback loop. In cells with an unsynchronized clock, VPA simultaneously upregulated the expression of genes encoding core circadian transcription factors that regulate the positive and negative limbs of the feedback loop. Using low dose glucocorticoid, we synchronized cultured fibroblast cells to a circadian oscillatory pattern. Whether VPA was added at the time of synchronization or 12h later at CT12, we found that VPA disrupted the oscillatory expression of multiple genes encoding essential transcription factors that regulate circadian rhythm. Therefore, we conclude that VPA has a potent effect on the circadian rhythm transcription-translation feedback loop that may be linked to negative VPA side effects in humans. Furthermore, our study suggests potential chronopharmacology implications of VPA usage. Copyright © 2017. Published by Elsevier Inc.

Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

Energy Technology Data Exchange (ETDEWEB)

Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

1995-09-01

The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

EPRTM engineered features for core melt mitigation in severe accidents

International Nuclear Information System (INIS)

Fischer, Manfred; Henning, Andreas

2009-01-01

For the prevention of accident conditions, the EPR TM relies on the proven 3-level safety concepts inherited from its predecessors, the French 'N4' and the German 'Konvoi' NPP. In addition, a new, fourth 'beyond safety' level is implemented for the mitigation of postulated severe accidents (SA) with core melting. It is aimed at preserving the integrity of the containment barrier and at significantly reducing the frequency and magnitude of activity releases into the environment under such extreme conditions. Loss of containment integrity is prevented by dedicated design measures that address short- and long-term challenges, like: the melt-through of the reactor pressure vessel under high internal pressure, energetic hydrogen/steam explosions, containment overpressure failure, and basemat melt-through. The EPR TM SA systems and components that address these issues are: - the dedicated SA valves for the depressurization the primary circuit, - the provisions for H 2 recombination, atmospheric mixing, steam dilution, - the core melt stabilization system, - the dedicated SA containment heat removal system. The core melt stabilization system (CMSS) of the EPR TM is based on a two-stage ex-vessel approach. After its release from the RPV the core debris is first accumulated and conditioned in the (dry) reactor pit by the addition of sacrificial concrete. Then the created molten pool is spread into a lateral core catcher to establish favorable conditions for the later flooding, quenching and cooling with water passively drained from the Internal Refueling Water Storage Tank. Long-term heat removal from the containment is achieved by sprays that are supplied with water by the containment heat removal system. Complementing earlier publications focused on the principle function, basic design, and validation background of the EPR TM CMSS, this paper describes the state achieved after detailed design, as well as the technical solutions chosen for its main components, including

Examination of off-site emergency protective measures for core melt accidents

International Nuclear Information System (INIS)

Aldrich, D.C.; Ericson, D.M. Jr.; Jones, R.B.

1978-01-01

Results from the Reactor Safety Study (RSS) have shown that to cause significant impacts off-site, i.e., sufficient quantities of biologically important radionuclides released, it is necessary to have a core melt accident. To mitigate the impact of such potential accidents, the design of appropriate emergency response actions requires information as to the relative merit of publicly available protective measures. In order to provide this information, a study using the consequence model developed for the RSS is being conducted to evaluate (in terms of reduced public health effects and dose exposure) potential off-site protective strategies. The paper describes the methods being used in the study as well as the results and conclusions obtained

Core disruptive accident margin seal

International Nuclear Information System (INIS)

Golden, M.P.

1979-01-01

An apparatus for sealing the annulus defined within a substantially cylindrical rotatable riser assembly and plug combination of a nuclear reactor closure head is described. The apparatus comprises an inflatable sealing mechanism disposed in one portion of the riser assembly near the annulus such that upon inflation the sealing mechanism is radially actuated against the other portion of the riser assembly thereby sealing the annulus. The apparatus further comprises a connecting mechanism which places one end of the sealing mechanism in fluid communication with the reactor cover gas so that overpressurization of the reactor cover gas will increase the radial actuation of the sealing mechanism thus enhancing sealing of the annulus

Core disruptive accident margin seal

International Nuclear Information System (INIS)

Garin, J.; Belsick, J.C.

1978-01-01

Disclosed is an apparatus for sealing the annulus defined between a substantially cylindrical rotatable first riser assembly and plug combination disposed in a substantially cylindrical second riser assembly and plug combination of a nuclear reactor system. The apparatus comprises a flexible member disposed between the first and second riser components and attached to a metal member which is attached to an actuating mechanism. When the actuating mechanism is not actuated, the flexible member does not contact the riser components thus allowing the free rotation of the riser components. When desired, the actuating mechanism causes the flexible member to contact the first and second riser components in a manner to block the annulus defined between the riser components, thereby sealing the annulus between the riser components

Analysis of the core reflooding of a PWR reactor under a loss-of-coolant postulated accident

International Nuclear Information System (INIS)

Austregesilo Filho, H.

1978-12-01

The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel element and containment integrity. The analytical model presented was applied to the simulation of Angra I core reflooding phase, after a double-ended break between pressure vessel and discharge of one of the main coolant pumps. For this accident, with a discharge coefficient of C sub(D) = 0.4, the highest peak cladding temperature is expected. (author) [pt

Evaluation of long-term post-accident core cooling of Three Mile Island Unit 2

Energy Technology Data Exchange (ETDEWEB)

None

1979-04-15

On the basis of current understanding of the accident scenario and available data, the staff reports here on its evaluation of the condition of the core and the core flow resistance as it might affect ability to cool the core by natural circulation. The natural circulation cooling capability of TMI-2 for the estimated core flow resistance and a variety of other conditions is evaluated and a comparison of the Base Case and off-nominal plant configurations is presented. The potential for and effects of natural convection core cooling are addressed, and the staff recommendations for reactor performance acceptance criteria upon initiation of natural convection are presented.

How did Fukushima-Dai-ichi core meltdown change the probability of nuclear accidents?

International Nuclear Information System (INIS)

Escobar Rangel, Lina; Leveque, Francois

2012-10-01

How to predict the probability of a nuclear accident using past observations? What increase in probability the Fukushima Dai-ichi event does entail? Many models and approaches can be used to answer these questions. Poisson regression as well as Bayesian updating are good candidates. However, they fail to address these issues properly because the independence assumption in which they are based on is violated. We propose a Poisson Exponentially Weighted Moving Average (PEWMA) based in a state-space time series approach to overcome this critical drawback. We find an increase in the risk of a core meltdown accident for the next year in the world by a factor of ten owing to the new major accident that took place in Japan in 2011. (authors)

Determination of the availability of core exit thermocouples during severe accident situations

International Nuclear Information System (INIS)

Edson, J.L.

1985-04-01

This report presents the findings and recommendations of the Nuclear Power Plant Instrumentation Evaluation (NPPIE) program concerning signal validation methods to determine the on-line availability of core exit thermocouples during accident situations. Methods of selecting appropriate signal validation techniques are discussed and sources of error identified. This report shows that through the use of these techniques the existence of high-temperature-caused errors may be detected as they occur. Specific recommendations for application of selected signal validation techniques to core exit thermocouples and other measurement systems are made. 23 refs., 22 figs., 3 tabs

Postulated accident conditions for air cleaning systems and radiological dose assessments for containment options

International Nuclear Information System (INIS)

Hilliard, R.K.; Postma, A.K.

1975-01-01

Ambient conditions and performance requirements for emergency air cleaning systems applicable to commercial LMFBR plants were studied. The focus of this study centered on aerosol removal under hypothetical core disruptive accident conditions. Effort completed includes a review of air cleaning systems related to LMFBR plants, selection of three reference containment system designs, postulation of the EACS design basis accident (EACS-DBA), analysis of thermal conditions resulting from the DBA, analysis of aerosol transport behavior following the DBA, and an estimate of bone dose at the site boundary for each of the reference plant designs. Reference plant concepts were a single containment system (e.g., FFTF), a double containment system (e.g., CRBRP with closed head compartment), and a containment-confinement design in which an inerted, sealed primary volume was located within a ventilated building whose exhaust was filtered. The reference design basis accident selected here involved release to the inner containment system of 1 percent of non-volatile solids and plutonium, 25 percent of core halogens, 25 percent of core volatile solids, 100 percent of core noble gases, 68 lbs of sodium vapor and 5000 lbs of liquid sodium. 13 references. (U.S.)

Isolated scaphotrapeziotrapezoid osteoarthritis: Prevalence, symptomatology and associated scapholunate ligament disruption in a population presenting to an accident and emergency department with acute wrist injuries

International Nuclear Information System (INIS)

Higginson, Antony P.; Braybrook, Jason; Williams, Stephen; Finlay, David

2001-01-01

AIM: To determine the prevalence of isolated scaphotrapeziotrapezoid osteoarthritis in a population presenting to an Accident and Emergency Department of Leicester Royal Infirmary with acute wrist injuries. Also to identify the presence of scapholunate ligament disruption in this patient group and quantify symptoms and loss of function in terms of the modified system of Green and O'Brien, a recognized clinical scoring system. MATERIALS AND METHODS: A total of 1711 radiographs of patients attending the Accident and Emergency Department were prospectively reviewed over a 5-month period. Those patients with isolated scaphotrapeziotrapezoid osteoarthritis were invited for clinical review. RESULTS: Sixteen patients were identified with isolated scaphotrapeziotrapezoid osteoarthritis. Two had a poor Green and O'Brien score and evidence of scapholunate ligament disruption (P < 0.05). CONCLUSION: Isolated scaphotrapeziotrapezoid osteoarthritis has a prevalence of 1% in a population presenting to an Accident and Emergency Department with acute wrist injuries over the age of 30 years. Isolated scaphotrapeziotrapezoid osteoarthritis may be asymptomatic even though the changes in the joint are severe. Scapholunate ligament disruption is associated with a poor Green and O'Brien score, but is not present in the majority of cases. Higginson, A.P. et al. (2001)

Research activities at JAERI on core material behaviour under severe accident conditions

International Nuclear Information System (INIS)

Uetsuka, H.; Katanashi, S.; Ishijima, K.

1996-01-01

At the Japan Atomic Energy Research Institute (JAERI), experimental studies on physical phenomena under the condition of a severe accident have been conducted. This paper presents the progress of the experimental studies on fuel and core materials behaviour such as the thermal shock fracture of fuel cladding due to quenching, the chemical interaction of core materials at high temperatures and the examination of TMI-2 debris. The mechanical behaviour of fuel rod with heavily embrittled cladding tube due to the thermal shock during delayed reflooding have been investigated at the Nuclear Safety Research Reactor (NSSR) of JAERI. A test fuel rod was heated in steam atmosphere by both electric and nuclear heating using the NSSR, then the rod was quenched by reflooding at the test section. Melting of core component materials having relatively low melting points and their eutectic reaction with other materials significantly influence on the degradation and melt down of fuel bundles during severe accidents. Therefore basic information on the reaction of core materials is necessary to understand and analyze the progress of core melting and relocation. Chemical interactions have been widely investigated at high temperatures for various binary systems of core component materials including absorber materials such as Zircaloy/Inconel, Zircaloy/stainless steel, Zircaloy/(Ag-In-Cd), stainless steel B 4 C and Zircaloy/B 4 C. It was found that the reaction generally obeyed a parabolic rate law and the reaction rate was determined for each reaction system. Many debris samples obtained from the degraded core of TMI-2 were transported to JAERI for numerous examinations and analyses. The microstructural examination revealed that the most part of debris was ceramic and it was not homogeneous in a microscopic sense. The thermal diffusivity data was also obtained for the temperature range up to about 1800K. The data from the large scale integral experiments were also obtained through the

Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

International Nuclear Information System (INIS)

Kroeger, P.G.; Colman, J.; Araj, K.

1983-01-01

The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

Safety characteristics of the US advanced liquid metal reactor core

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

1991-01-01

The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

Analysis of forces on core structures during a loss-of-coolant accident. Final report

International Nuclear Information System (INIS)

Griggs, D.P.; Vilim, R.B.; Wang, C.H.; Meyer, J.E.

1980-08-01

There are several design requirements related to the emergency core cooling which would follow a hypothetical loss-of-coolant accident (LOCA). One of these requirements is that the core must retain a coolable geometry throughout the accident. A possible cause of core damage leading to an uncoolable geometry is the action of forces on the core and associated support structures during the very early (blowdown) stage of the LOCA. An equally unsatisfactory design result would occur if calculated deformations and failures were so extensive that the geometry used for calculating the next stages of the LOCA (refill and reflood) could not be known reasonably well. Subsidiary questions involve damage preventing the operation of control assemblies and loss of integrity of other needed safety systems. A reliable method of calculating these forces is therefore an important part of LOCA analysis. These concerns provided the motivation for the study. The general objective of the study was to review the state-of-the-art in LOCA force determination. Specific objectives were: (1) determine state-of-the-art by reviewing current (and projected near future) techniques for LOCA force determination, and (2) consider each of the major assumptions involved in force determination and make a qualitative assessment of their validity

Fuel disruption mechanisms determined in-pile in the ACRR

International Nuclear Information System (INIS)

Wright, S.A.; Fischer, E.A.

1984-09-01

Over thirty in-pile experiments were performed to investigate fuel disruption behavior for LMFBR loss of flow (LOF) accidents. These experiments reproduced the heating transients for a variety of accidents ranging from slow LOF accidents to rapid LOF-driven-TOP accidents. In all experiments the timing and mode of the fuel disruption were observed with a high speed camera, enabling detailed comparisons with a fuel pin code, SANDPIN. This code transient intra- and inter-granular fission gas behavior to predict the macroscopic fuel behavior, such as fission gas induced swelling and frothing, cracking and breakup of solid fuel, and fuel vapor pressure driven dispersal. This report reviews the different modes of fuel disruption as seen in the experiments and then describes the mechanism responsible for the disruption. An analysis is presented that describes a set of conditions specifying the mode of fuel disruption and the heating conditions required to produce the disruption. The heating conditions are described in terms of heating rate (K/s), temperature gradient, and fuel temperature. A fuel disruption map is presented which plots heating rate as a function of fuel temperature to illustrate the different criteria for disruption. Although this approach to describing fuel disruption oversimplifies the fission gas processes modeled by SANDPIN, it does illustrate the criteria used to determine which fuel disruption mechanism is dominant and on what major fission gas parameters it depends

Control rod drop accident analysis for the mixed core project in Ling Ao NPS

International Nuclear Information System (INIS)

Zhang Shishun; Zhou Zhou; Xiao Min

2004-01-01

AFA-2G assemblies in Ling Ao NPS (LNPS) have been replaced gradually by AFA-3G assemblies from cycle 2 and subsequent cycles. the enrichment of the fuels will be increased from 3.2% to 3.7% from cycle 3 in Ling Ao. Therefore, the study of ling Ao mixed core and increased enrichment have been performed since 2001. Lots of accidents need to be re-analyzed in Ling Ao NPS in order to verify its safety requirements for the new fuel management. Control rod drop accident for LNPS was re-analyzed in 2001 in frame of FRAMATOME ANP analytical methodology. The analytical codes used in the accident analysis include SCIENCE, ESPADON, CINEMA, CANTAL and FLICA III. The control rod drop accident analysis is performed with respect to the 10 reference cycles of the generic fuel management design for Ling Ao mixed core and increased enrichment study. The pre-drop FδH for the first transition cycles and other cycles are 1.52 and 1.55, respectively. For detected dropped rod configurations, the negative flux rate protection system actuates a reactor trip. For the non-detected dropped rod configurations, the minimum DNBR values have been evaluated with conservative analysis methodology and assumptions and the DNBR fuel design limit is respected the analytical results shows that, for all the non-detected dropped rod configurations, the minimum DNB margin is about 2% which occurs in AFA-2G fuel assembly in the first transition cycle. (author)

Fuel and control rod failure behavior during degraded core accidents

International Nuclear Information System (INIS)

Chung, K.S.

1984-01-01

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

LMFBR fuel analysis. Task B. Post-accident heat removal. Final report, July 1, 1975--September 30, 1976

International Nuclear Information System (INIS)

Castle, J.; Catton, I.; Somerton, C.; Wu, R.

1976-11-01

The report deals with the behavior of molten core debris following a hypothetical core disruptive accident in the proposed Clinch River Breeder Reactor Plant. Heat dissipating characteristics of an ex-vessel sacrificial bed have been analyzed. A novel form of heat transfer, analogous to film boiling, has been proposed to describe heat transfer from a heat generating pool to surrounding steel walls. Bounding type heat transfer calculations are also made to quantify such hypothetical accident characteristics as debris bed remelting, debris bed dryout in sodium, and failure of the reactor cavity steel liner. Several documents that have been submitted to the NRC for its review of the CRBRP are discussed with attention being drawn to heat transfer related issues

Visual in-pile fuel disruption experiments

International Nuclear Information System (INIS)

Cano, G.L.; Ostensen, R.W.; Young, M.F.

1978-01-01

In a loss-of-flow (LOF) accident in an LMFBR, the mode of disruption of fuel may determine the probability of a subsequent energetic excursion. To investigate these phenomena, in-pile disruption of fission-heated irradiated fuel pellets was recorded by high speed cinematography. Instead of fuel frothing or dust-cloud breakup (as used in the SAS code) massive and very rapid fuel swelling, not predicted by analytical models, occurred. These tests support massive fuel swelling as the initial mode of fuel disruption in a LOF accident. (author)

Prevention and investigations of core degradation in case of beyond design accidents of the 2400 MWTH gas-cooled fast reactor

International Nuclear Information System (INIS)

Bertrand, F.; Gatin, V.; Bentivoglio, F.; Gueneau, C.

2011-01-01

The present paper deals with studies carried out to assess the ability of the core of the Gas Fast Reactor (GFR) to withstand beyond design accidents. The work presented here is aimed at simulating the behaviour of this core by using analytical models whose input parameters are calculated with the CATHARE2 code. Among possible severe accident initiators, the Unprotected Loss Of Coolant Accident (ULOCA of 3 Inches diameter) is investigated in detail in the paper with CATHARE2. Additionally, a simplified pessimistic assessment of the effect of a postulated power excursion that could result from the failure of prevention provisions is presented. (author)

Degraded core accidents: review of aerosol behaviour in the containment of a PWR

International Nuclear Information System (INIS)

Nichols, A.L.; Walker, B.C.

1981-09-01

Low probability-high consequence accidents have become an important issue in reactor safety studies. Such accidents would involve damage to the core and the subsequent release of radioactive fission products into the environment. Aerosols play a major role in the transport and removal of these fission products in the reactor building containment. The aerosol mechanisms, computer modelling codes and experimental studies used to predict aerosol behaviour in the containment of a PWR are reviewed. There are significant uncertainties in the aerosol source terms and specific recommendations have been made for further studies, particularly with respect to code development and high density aerosol-fission product transport within closed systems. (author)

Severe Accident Management Guidance: Lessons Still to be Learned after Fukushima

International Nuclear Information System (INIS)

Vayssier, G.

2016-01-01

After the accidents in Three Mile Island (TMI) and Chernobyl, many countries decided to develop and implement guidelines specifically directed to mitigate accidents with core damage, so-called severe accidents. The guidelines are usually named Severe Accident Management Guidelines (SAMG). In the USA, all operating plants had these guidelines in place at the end of 1998. Most other countries followed later, but today, it can be said that many nuclear power plants in the world have such guidelines in place. Typically, however, the guidelines were constructed under the assumption that many plant systems still will be available, i.e. there will be DC to feed the instruments, AC to feed equipment and water to restore cooling to the core. Typically, this was basically the situation at TMI: most equipment was functional, only the insight of what had happened had been lost and operators did not know how to respond. At Fukushima-Daiichi, a Site Disruptive Accident (SDA) occurred and it appeared that the situation was much more complex: much of the needed supportive equipment needed was unavailable, which greatly complicated the handling of the event. In this paper, the major shortcomings of the present existing SAMG are discussed, both from a technical, and an organisational viewpoint. It is concluded that, where proper regulation still is missing, the development of an industrial standard is recommended to define adequate tools and guidelines to mitigate severe accidents, including SDAs. (author).

Safety Strategy of JSFR establishing In-Vessel Retention of Core Disruptive Accident

International Nuclear Information System (INIS)

Tobita, Yoshiharu

2013-01-01

Coolability of debris bed was confirmed by debris bed temperature analysis coupled with the cooling system, according to the following material relocation scenario. → Case 1: Upward ejection in Transition Phase to cause shutdown. → Case 2: Early downward ejection of fuel through CRGT. → Case 3: Whole fuel accumulates on the core catcher (bounding). The flow reversal of a primary coolant loop of the two loop system of the JSFR which is caused by possible imbalance between two DHRS loops increase the flow in RV. Helpful for long-term cooling

Core dynamics of HTR under ATWS and accident conditions

International Nuclear Information System (INIS)

Nabbi, R.

1988-05-01

The systematic classification of the ATWS has been undertaken by analogy to the considerations made for LWR. The initiating events of ATWS and protection actions of safety systems resulting from monitoring of the system variables have been described. The main emphasis of this work is the analysis of the core dynamic consequences of scram failure during the anticipated transients. The investigation has shown that because of the temperature feedback mechanisms a temperature rise during the ATWS results in a self-shutdown of the reactor. Further inherent safety features of the HTR - conditioned by the high heat capacity of the core and by the compressibility of the coolant - do effectively counteract an undesirable increase of temperature and pressure in the primary circuit. In case of the long-term failure of the forced cooling and following core heatup, neutron physical phenomena appear which determine the reactivity behaviour of the HTR. They are, for instance, the decay of Xenon 135, release of the fission products and subsiding of the top reflector. The results of the computer simulations show that a recriticality has to be excluded during the first 2 days if the reactor is shutdown by the reflector rods at the beginning of the accident. (orig./HP) [de

Specific features of RBMK severe accidents progression and approach to the accident management

International Nuclear Information System (INIS)

Vasilevskij, V.P.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Cherkashov, Yu.M.

2001-01-01

Fundamental construction features of the LWGR facilities (absence of common external containment shell, disintegrated circulation circuit and multichannel reactor core, positive vapor reactivity coefficient, high mass of thermally capacious graphite moderator) predetermining development of assumed heavy non-projected accidents and handling them are treated. Rating the categories of the reactor core damages for non-projected accidents and accident types producing specific grope of damages is given. Passing standard non-projected accidents, possible methods of attack accident consequences, as well as methods of calculated analysis of non-projected accidents are demonstrated [ru

Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation

International Nuclear Information System (INIS)

Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

2010-01-01

An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

Energy Technology Data Exchange (ETDEWEB)

Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

2010-03-01

An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

Safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release

International Nuclear Information System (INIS)

Pointner, W.; Broecker, A.

2012-01-01

The report on safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release covers the following issues: assessment of the relevant status for PWR, evaluation of the national and international (USA, Canada, France) status, actualization of recommendations, transferability from PWR to BWR. Generic studies on the core cooling capability in case of insulation material release in BWR-type reactors were evaluated.

Methodological aspects of core meltdown accidents frequency estimates

International Nuclear Information System (INIS)

Matthis, P.

1984-01-01

A survey is given of the work of the ecological institute relating to models and methods used in the German Risk Study for the assessment of core meltdown accident frequency. A statistical model used by the ecological institute for the estimation of the outage behaviour of components is taken as a comparison, which leads to the conclusion that no appropriate methods for the assessment of component reliability are available to date. Furthermore, there are no secured methods for error propagation computation. The lower limits for the ranges of reliability of components are calculated by approximation. As a result of imperfect modelling and of a number of methodical inaccuracies and neglects, the German Risk Study underestimates the ranges of component reliability by a factor of 3 to 70 (depending on the type of component). (RF) [de

Cleanup of large areas contaminated as a result of a nuclear accident

International Nuclear Information System (INIS)

1989-01-01

The purposes of the report are to provide an overview of the methodology and technology available to clean up contaminated areas and to give preliminary guidance on matters related to the planning, implementation and management of such cleanups. This report provides an integrated overview of important aspects related to the cleanup of very large areas contaminated as a result of a serious nuclear accident, including information on methods and equipment available to: characterize the affected area and the radioactive fallout; stabilize or isolate the contamination; and clean up contaminated urban, rural and forested areas. The report also includes brief sections on planning and management considerations and the transport and disposal of the large volumes of wastes arising from such cleanups. For the purposes of this report, nuclear accidents which could result in the deposition of decontamination over large areas if the outer containment fails badly include: 1) An accident with a nuclear weapon involving detonation of the chemical high explosive but little, if any, nuclear fission. 2) A major loss of medium/high level liquid waste (HLLW) due to an explosion/fire at a storage site for such waste. 3) An accident at a nuclear power plant (NPP), for example a loss of coolant accident, which results in some core disruption and fuel melting. 4) An accident at an NPP involving an uncontrolled reactivity excursion resulting in the violent ejection of a reactor core material and rupture of the containment building. 117 refs, 32 figs, 12 tabs

A assessment of loss-of-heat-sink accident with scram in the LMFBR

International Nuclear Information System (INIS)

Bari, R.A.; Ludewig, H.; Pratt, W.T.; Sun, Y.H.

1978-01-01

A description of a slow core meltdown in a liquid metal fast breeder reactor is presented for conditions of loss-of-heat-sink following neutronic shutdown. Simple models are developed for the prediction of phase changes and/or relocation of the core materials including fuel, clad, ducts, control rod absorber material (B 4 C), and plenum gases. The sequence of events is accounted for and the accident progression is described up to the point of recriticality. The neutronic behavior of the disrupted core is analyzed in R-Z geometry with a static transport theory code. For most scenarios assessed, the reactor is expected to become recritical although large ramp rates are not anticipated. (author)

Assessment of the loss-of-heat-sink accident with scram in the LMFBR

International Nuclear Information System (INIS)

Bari, R.A.; Ludewig, H.; Pratt, W.T.; Sun, Y.H.

1978-01-01

A description of a slow core meltdown in a liquid metal fast breeder reactor is presented for the conditions of loss-of-heat-sink following neutronic shutdown. Simple models are developed for the prediction of phase changes and/or relocation of the core materials including fuel, clad, ducts, control rod absorber material (B 4 C), and plenum gases. The sequence of events is accounted for and the accident progression is described up to the point of recriticality. The neutronic behavior of the disrupted core is analyzed in R-Z geometry with a static transport theory code. For most scenarios assessed, the reactor is expected to become recritical although large ramp rates are not anticipated

Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

Energy Technology Data Exchange (ETDEWEB)

Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

2016-02-15

Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

Reactivity Accidents in CAREM-25 Core with and Without Safety Systems Actuation

International Nuclear Information System (INIS)

Gimenez, Marcelo; Vertullo, Alicia; Schlamp, Miguel

2000-01-01

A reactivity accident in CAREM core can be provoked by different initiating events, a cold water injection in pressure vessel, a secondary side steam line breakage and a failure in the absorbing rods drive system.The present work analyses inadverted control rod withdraws transients.Maximum worth control rod (2.5 $) at normal velocity (1 cm/s) is adopted for the simulations (Reactivity ramp of 0.018 $/s).Different scenarios considering actuation of first shutdown system (FSS), second shutdown system (SSS) and selflimiting conditions were modeled.Results of the accident with actuation of FSS show that safety margins are well above critical values (DNBR and CPR).In the cases with failure of the FSS and success of SSS or selflimited, safety margins are below critical values, however, the SSS provides a reduction of elapsed time under advised margins

Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

International Nuclear Information System (INIS)

Maschek, W.; Royl, P.

1988-09-01

Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

In-vessel core degradation in LWR severe accidents: a state of the art report to CSNI january 1991

International Nuclear Information System (INIS)

1991-11-01

This state of the art report on in-vessel core degradation has been produced at the request of CSNI Principal Working Group 2. The objective of the report is to present to CSNI member countries the status of research and related information on in-vessel degraded core behaviour in both Pressurised Water Reactors (PWR) and Boiling Water Reactors (BWR). Information on experiments, codes and comparisons of calculations with experiments up to january 1991 is summarised and reviewed. Integrated codes, which are wider in scope than just in-vessel degradation are covered as well as specialist, degraded core codes. Implications for PWR and BWR plant calculations are considered. Conclusions and recommendations for research, plant calculations and further CSNI activity in this area are the subject of the final chapter. A major conclusion of the report is that early phase core degradation is relatively well understood. However, codes need further development to bring them up to date with the experimental database, particularly to include low temperature liquefaction processes. These processes significantly affect early phase core degradation and their neglect could affect assessments of accident management actions (including recriticality in BWR severe accidents)

Analysis of the loss of coolant accident for LEU cores of Pakistan research reactor-1

International Nuclear Information System (INIS)

Khan, L.A.; Bokhari, I.H.; Raza, S.H.

1993-12-01

Response of LEU cores for PARR-1 to a Loss of Coolant Accident (LOCA) has been studied. It has been assumed that pool water drains out to double ended rupture of primary coolant pipe or complete shearing of an experimental beam tube. Results show that for an operating power level of 10 MW, both the first high power and equilibrium cores would enter into melting conditions if the pool drain time is less than 22 h and 11 h respectively. However, an Emergency Core Cooling System (ECCS) capable of spraying the core at flow rate of 8.3 m/sup 3/h, for the above mentioned duration, would keep the peak core temperature much below the critical value. Maximum operating power levels below which melting would not occur have been assessed to 3.4 MW and 4.8 MW, respectively, for the first high power and equilibrium cores. (author) 5 figs

Comparative analysis of a hypothetical 0.1 $/SEC transient overpower accident in an irradiated LMFBR core using different computer models

International Nuclear Information System (INIS)

Cacciabue, P.C.; Fremont, R. de; Renard, A.

1982-01-01

The Report gives the results of comparative calculations performed by the Whole Core Accident Codes Group which is a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee for a hypothetical transient overpower accident in an irradiated LMFBR core. Different computer codes from members of the European Community and the United States were used. The calculations are based on a Benchmark problem, using commonly agreed input data for the most important phenomena, such as the fuel pin failure threshold, FCl parameters, etc. Beside this, results with alternative assumptions for theoretical modelling are presented with the scope to show in a parametric way the influence of more advanced modelling capabilities and/or better (so-called best estimate) input data for the most important phenomena on the accident sequences

An investigation of core liquid level depression in small break loss-of-coolant accidents

International Nuclear Information System (INIS)

Schultz, R.R.; Watkins, J.C.; Motley, F.E.; Stumpf, H.; Chen, Y.S.

1991-08-01

Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant out of the core region and uncovers the rod upper elevations. The net result is core liquid level depression. Core liquid level depression and subsequent core heatups are investigated using subscale data from the ROSA-IV Program's 1/48-scale Large Scale Test Facility (LSTF) and the 1/1705-scale Semiscale facility. Both facilities are Westinghouse-type, four-loop, pressurized water reactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PF1/MOD1 (Version 12.7) thermal-hydraulic code, are also described and summarized. Finally, the response of a typical Westinghouse four-loop plant (RESAR-3S) was calculated to qualitatively study coal liquid level depression in a full-scale system. 31 refs., 37 figs., 6 tabs

Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report

International Nuclear Information System (INIS)

Rogers, J.T.

1984-06-01

This volume of appendices presents listings and sample runs of the computer codes used in the study of the thermalhydraulic behaviour of CANDU reactor cores during severe loss of coolant accidents. The codes, written in standard FORTRAN, are MODBOIL, to calculate moderator temperatures, pressures and water levels; DEBRIS, to calculate the transient temperature distribution in the debris of calandria and pressure tubes and fuel pellets; MOLTENPOOL, to calculate the temperature history in a pool of molten debris; CONFILM, to calculate the behaviour of a condensing film of vaporized core debris on the calandria wall, and BLDG, to calculate the pressurization of the containment during the expulsion of moderator through pressure relief ducts. In addition there are discussions of the average condensation heat transfer coefficient for vaporized core material on the calandria wall, and of vapor explosions

Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

International Nuclear Information System (INIS)

Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

1996-02-01

In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core

Thermal hydraulic features of the TMI accident

International Nuclear Information System (INIS)

Tolman, B.

1985-01-01

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

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

International Nuclear Information System (INIS)

Tukiran

2002-01-01

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

Evaluation of re-criticality potential in Fukushima Dai-ichi reactors following core damage accidents

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The re-criticality potential of the debris-bed, formed of the degraded core materials, cannot be ruled out during the cooling-down procedure of the Fukushima Dai-ichi NPPs. In this study the re-criticality potential has systematically investigated based on the core disruption phase analysis using a IMPACT-SAMPSON code prepared by The Institute of Applied Energy (IAE). The results obtained for the re-criticality potential, characterized by the eigen-values k-eff dependent on the debris composition formed at the core, RPV bottom, and PCV pedestal, are reflected to the arguments on the re-criticality prevention measures, such as timing and concentration of boron-compounds, during the cooling-down process of the Fukushima Dai-ichi NPPs. (author)

Containment loadings due to hydrogen burning in LWR core meltdown accidents

International Nuclear Information System (INIS)

Cybulskis, P.

1981-01-01

The potential pressure loadings due to hydrogen burning under conditions representative of meltdown accident conditions are examined for a variety of PWR and BWR containment designs. For the PWR, the large dry, ice condenser, as well as subatmospheric containments are considered. For the BWR, MARK I, II, and III pressure suppression containments are evaluated. The key factors considered are: free volume, design pressure, extend to hydrogen generation, and the flammability of the atmosphere under a range of accident conditions. The potential for and the possible implications of hydrogen detonation are also considered. The results of these analyses show that the accumulation and rapid burning of the quantities of hydrogen that would be generated during core meltdown accidents will lead to pressures above design levels in all of the containments considered. As would be expected, containments characterized by small volumes and/or low design pressures are the most vulnerable to damage due to hydrogen burning. Large volume, high pressure designs may also be threatened but offer significantly more potential for accomodating hydrogen burns. The attainment of detonable hydrogen mixtures is made easier by smaller containment volumes. Detonable mixtures are also possible in the larger volume containments, but imply the accumulation of hydrogen for long periods of time without prior ignition. Hydrogen detonations, if they occur, would probably challenge the integrity of any of the containments considered. (orig.)

The Accident Analysis Due to Reactivity Insertion of RSG GAS 3.55 g U/cc Silicide Core

International Nuclear Information System (INIS)

Endiah Puji-Hastuti; Surbakti, Tukiran

2004-01-01

The fuels of RSG-GAS reactor was changed from uranium oxide with 250 g U of loading or 2.96 g U/cc of fuel loading to uranium silicide with the same loading. The silicide fuels can be used in higher density, staying longer in the reactor core and hence having a longer cycle length. The silicide fuel in RSG-GAS core was made up in step-wise by using mixed up core Firstly, it was used silicide fuel with 250 g U of loading and then, silicide fuel with 300 g U of loading (3.55 g U/cc of fuel loading). In every step-wise of fuel loading, it must be analyzed its safety margin. In this occasion, the reactivity accident of RSG-GAS core with 300 g U of silicide fuel loading is analyzed. The calculation was done using EUREKA-2/RR code available at P2TRR. The calculation was done by reactivity insertion at start up and power rangers. The worst case accident is transient due to control rod with drawl failure at start up by means of lowest initial power (0.1 W), either in power range. From all cases which have been done, the results of analysis showed that there is no anomaly and safety margin break at RSG-GAS core with 300 g U silicide fuel loading. (author)

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

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali R.

2004-12-01

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

Stability Analysis of the EBR-I Mark-II Core Meltdown Accident

Energy Technology Data Exchange (ETDEWEB)

Lim, Jae-Yong; Kang, Chang Mu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The purpose of this paper is to analyze the stability of the EBR-I core meltdown accident using the NuSTAB code. The result of NuSTAB analysis is compared with previous stability analysis by Sandmeier using the root locus method. The Experimental Breeder Reactor I (EBR-1) at Argonne National Laboratory was designed to demonstrate fast reactor breeding and to prove the use of liquid-metal coolant for power production and reached criticality in August 1951. The EBR-I reactor was undergoing a series of physics experiments and the Mark-II core was melted accidentally on Nov. 29, 1955. The experiment was going to increase core temperature to 500C to see if the reactor loses reactivity, and scram when the power reached 1500 kW or doubling of fission rate per second. However the operator scrammed with a slow moving control and missed the shutdown by two seconds and caused the core meltdown. The NuSTAB code has an advantage of analyzing space-dependent fast reactors and predicting regional oscillations compared to the point kinetics. Also, NuSTAB can be useful when the coupled neutronic-thermal-hydraulic codes cannot be used for stability analysis. Future work includes analyses of the PGSFR for various operating conditions as well as further validation of the NuSTAB calculations against SFR stability experiments when such experiments become available.

Radioactive contamination of Danish territory after core-melt accidents at the Barsebaeck power plant

International Nuclear Information System (INIS)

Gjoerup, H.L.; Jensen, N.O.; Hedemann Jensen, P.; Kristensen, L.; Nielsen, O.J.; Petersen, E.L.; Petersen, T.; Roed, J.; Thykier-Nielsen, S.; Heikel Vinter, F.; Warming, L.; Aarkrog, A.

1982-03-01

An assessment is made of the radioactive contamination of Danish territory in the event of a core-melt accident at the Barsebaeck nuclear power plant in Sweden. Accidents including both core melt-down and containment failure are considered. Consequences are calculated for a BWR-3 release under common meteorological conditions and for a BWR-2 release under extreme meteorological conditions. Calculations are based on experiments and theoretical work relating to deposition velocities for different types of surface, shielding effect of structures, and weathering. The effects are described of different dose-reducing measures, e.g., decontamination, relocation, destruction of contaminated foodstuffs. The collective effective dose equivalent from external gamma radiation from deposited activity integrated over a time period of 30 years, is calculated to be 3.6 Megamanrem in the BWR-3 case without dose-reducing measures. For the BWR-2 case, the corresponding dose is approx. 41 Megamanrem. A combination of temporary relocation, hosing of roads etc. and digging of gardens is estimated to reduce these doses to approx. 2.5 Megamanrem and approx. 15 Megamanrem, respectively. The collective committed effective dose equivalent from the consumption of contaminated foodstuffs is calculated to 23 Megamanrem in the BWR-3 case without dose-reducing measures. This dose could be reduced to 0.2 Megamanrem if contaminated crops are destroyed during the first year after the accident and if changes are made in agricultural production in the contaminated area. The corresponding doses in the BWR-2 case would be 197 Megamanrem and 1.4 Megmanrem, respectively. (author)

PWR degraded core analysis

International Nuclear Information System (INIS)

Gittus, J.H.

1982-04-01

A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

Severe accident phenomena

International Nuclear Information System (INIS)

Jokiniemi, J.; Kilpi, K.; Lindholm, I.; Maekynen, J.; Pekkarinen, E.; Sairanen, R.; Silde, A.

1995-02-01

Severe accidents are nuclear reactor accidents in which the reactor core is substantially damaged. The report describes severe reactor accident phenomena and their significance for the safety of nuclear power plants. A comprehensive set of phenomena ranging from accident initiation to containment behaviour and containment integrity questions are covered. The report is based on expertise gained in the severe accident assessment projects conducted at the Technical Research Centre of Finland (VTT). (49 refs., 32 figs., 12 tabs.)

Use of PSA and severe accident assessment results for the accident management

International Nuclear Information System (INIS)

Jang, S. H.; Kim, H. G.; Jang, H. S.; Moon, S. K.; Park, J. U.

1993-12-01

The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management

Use of PSA and severe accident assessment results for the accident management

Energy Technology Data Exchange (ETDEWEB)

Jang, S H; Kim, H G; Jang, H S; Moon, S K; Park, J U [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

1993-12-15

The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management.

Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

2014-01-01

The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

How to arrest a core meltdown accident (doing nothing); Como detener un accidente con fusion de nucleo (sin hacer nada)

Energy Technology Data Exchange (ETDEWEB)

Baron, Jorge H [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

2000-07-01

In the eventual situation of a severe accident in a nuclear reactor, the molten core is able to relocate inside the pressure vessel. This may lead to the vessel failure, due to the thermal attack of the molten core (at approximation of 3000K) on the vessel steel wall. The vessel failure implies the failure of a very important barrier that contains the radioactive materials generated during the reactor operation, with a significant risk of producing high radiation doses both on operators and on the public. It is expected, for the new generation of nuclear reactors, that these will be required to withstand (by design) a core melt down accident, without the need for an immediate evacuation of the surrounding population. In this line, the use of a totally passive system is postulated, which fulfills the objective of containing the molten core inside the pressure vessel, at low temperature (approximation 1200K) precluding its failure. The conceptual design of a passive in-vessel core catcher is presented in this paper, built up of zinc, and designed for the CAREM-25 nuclear power plant. (author)

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

International Nuclear Information System (INIS)

Verfondern, K.; Nabielek, H.

2008-01-01

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

On high-temperature reactor accident topology

International Nuclear Information System (INIS)

Fassbender, J.; Kroeger, W.; Wolters, J.

1981-01-01

American and German risk studies for an HTGR and independent investigations of hypothetical accident sequences led to a fundamental understanding of the topology of HTGR accident sequences. The dominating importance of core heat-up accidents was confirmed and the initiating events were identified. Complications of core heat-up accidents by air or water ingress are of minor importance for the risk, whereas the long-term development of accidents during days and weeks plays an important role for the environmental impact. The risk caused by an HTGR at a German site cannot yet be determined exactly, because no modern German HTGR design has passed a licensing procedure. Cautious estimates show that risk will appear to be substantially smaller than the LWR risk. The main reasons are the considerably reduced release of fission procucts and the slow development of core heat-up accidents leaving much time for measures which reduce the risk. (orig.) [de

Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix I. Accident definition and use of event trees

International Nuclear Information System (INIS)

1975-10-01

Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the core

Intersubassembly incoherencies and grouping techniques in LMFBR hypothetical overpower accident

International Nuclear Information System (INIS)

Wilburn, N.P.

1977-10-01

A detailed analysis was made of the FTR core using the 100-channel MELT-IIIA code. Results were studied for the transient overpower accident (where 0.5$/sec and 1$/sec ramps) and in which the Damage Parameter and the Failure Potential criteria were used. Using the information obtained from these series of runs, a new method of grouping the subassemblies into channels has been developed. Also, it was demonstrated that a 7-channel representation of the FTR core using this method does an adequate job of representing the behavior during a hypothetical disruptive transient overpower core accident. It has been shown that this new 7-channel grouping method does a better job than an earlier 20-channel grouping. It has also been demonstrated that the incoherency effects between subassemblies as shown during the 76-channel representation of the reactor can be adequately modeled by 7-channels, provided the 7-channels are selected according to the criteria stated in the report. The overall results of power and net reactivity were shown to be only slightly different in the two cases of the 7-channel and the 76-channel runs. Therefore, it can be concluded that any intersubassembly incoherencies can be modeled adequately by a small number of channels, provided the subassemblies making up these channels are selected according to the criteria stated

Sensitivity analysis of thermal hydraulic response in containment at core meltdown accident

International Nuclear Information System (INIS)

Kobayashi, Kensuke; Ishigami, Tsutomu; Horii, Hideo; Chiba, Takemi.

1985-01-01

A sensitivity analysis of thermal hydraulic response in a containment during a 'station blackout' (the loss of all AC power) accident at Browns Ferry unit one plant was performed with the computer code MARCH 1.0. In the analysis, the plant station batteries were assumed to be available for 4h after the initiation of the accident. The thermal hydraulic response in the containment was calculated by varying several input data for MARCH 1.0 independently and the deviation among calculated results were investigated. The sensitivity analysis showed that (a) the containment would fail due to the overtemperature without any operator actions for plant recovery, which would be strongly dependent on the model of the debris-concrete interaction and the input parameters for specifying the containment failure modes in MARCH 1.0, (b) a core melting temperature and an amount of water left in a primary system at the end of the meltdown were identified as important parameters which influenced the time of the containment failure, and (c) experimental works regarding the parameters mentioned above could be recommended. (author)

Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario

Energy Technology Data Exchange (ETDEWEB)

Andrews, Nathan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Faucett, Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Haskin, Troy Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Luxat, Dave [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Geiger, Garrett [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Codella, Brittany [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-10-01

Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recovery of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.

The influence of chemistry on core melt accidents

International Nuclear Information System (INIS)

Liljenzin, J.O.

1990-01-01

Chemical reactions play an important role in assessing the safety of nuclear power plants. The main source of heat in the early stage of an accident is due to a chemical reaction between steam and the circonium encapsulating the nuclear fuel. The heating and melting of fuel leads to a release of fission products which rapidly condense to form particles suspended in the surrounding gas. These aerosols are the main carriers of radioactivity as they may transport active material from the reactor vessel into the reactor containment building where it is deposited. The content of fission products in the aerosol particles and their chemical form determine their interaction with water molecules. Chemical forces laed to an absorption of water in the particles which transforms them into droplets with increased mass. The particles become spherical and hence deposit more rapidly on surrounding surfaces. There is a rapid reaction between boron carbide and stainless steel in the control blades of boiling water reactors. There is only a small formation of boric acid. This leads to a smaller formation of volatile iodine compounds. But the alloying process is likely to cause melting of the control blades so the are removed from the reactor core, a process which may have negative secondary effects. It has been found that a series of materials that are present in the reactor containment are likely to participate in various chemical reactions during an accident. Among these are electric cables, motors, thermal insulation, surface coatings and sheet metal. Metallic surface coatings and sheet metal can be some of the main sources of hydrogen. Effects from chemical reactions can be more accurately predicted by the new SHMAPP code, developed within this project, combining thermal, hydraulic and chemical phenomena. (AB)

Severe accident tests and development of domestic severe accident system codes

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

Severe accident tests and development of domestic severe accident system codes

International Nuclear Information System (INIS)

2013-01-01

According to lessons learned from Fukushima-Daiichi NPS accidents, the safety evaluation will be started based on the NRA's New Safety Standards. In parallel with this movement, reinforcement of Severe Accident (SA) Measures and Accident Managements (AMs) has been undertaken and establishments of relevant regulations and standards are recognized as urgent subjects. Strengthening responses against nuclear plant hazards, as well as realistic protection measures and their standardization is also recognized as urgent subjects. Furthermore, decommissioning of Fukushima-Daiichi Unit1 through Unit4 is promoted diligently. Taking into account JNES's mission with regard to these SA Measures, AMs and decommissioning, movement of improving SA evaluation methodologies inside and outside Japan, and prioritization of subjects based on analyses of sequences of Fukushima-Daiichi NPS accidents, three viewpoints was extracted. These viewpoints were substantiated as the following three groups of R and D subjects: (1) Obtaining near term experimental subjects: Containment venting, Seawater injection, Iodine behaviors. (2) Obtaining mid and long experimental subjects: Fuel damage behavior at early phase of core degradation, Core melting and debris formation. (3) Development of a macroscopic level SA code for plant system behaviors and a mechanistic level code for core melting and debris formation. (author)

SEVERE ACCIDENT ISSUES RAISED BY THE FUKUSHIMA ACCIDENT AND IMPROVEMENTS SUGGESTED

OpenAIRE

SONG, JIN HO; KIM, TAE WOON

2014-01-01

This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accide...

An assessment of the radiological consequences of releases to groundwater following a core-melt accident at the Sizewell PWR

International Nuclear Information System (INIS)

Maul, P.R.

1984-03-01

In the extremely unlikely event of a degraded core accident at the proposed Sizewell PWR it is theoretically possible for the core to melt through the containment, after which activity could enter groundwater directly or as a result of subsequent leaching of the core in the ground. The radiological consequences of such an event are analysed and compared with the analysis undertaken by the NRPB for the corresponding releases to atmosphere. It is concluded that the risks associated with the groundwater route are much less important than those associated with the atmospheric route. The much longer transport times in the ground compared with those in the atmosphere enable countermeasures to be taken, if necessary, to restrict doses to members of the public to very low levels in the first few years following the accident. The entry of long-lived radionuclides into the sea over very long timescales results in the largest contribution to population doses, but these are delivered at extremely low dose rates which would be negligible compared with background exposure. (author)

Accident progression event tree analysis for postulated severe accidents at N Reactor

International Nuclear Information System (INIS)

Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M.; Medford, G.T.

1990-06-01

A Level II/III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The accident progression analysis documented in this report determines how core damage accidents identified in the Level I PRA progress from fuel damage to confinement response and potential releases the environment. The objectives of the study are to generate accident progression data for the Level II/III PRA source term model and to identify changes that could improve plant response under accident conditions. The scope of the analysis is comprehensive, excluding only sabotage and operator errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study. The accident progression model allows complex interactions and dependencies between systems to be explicitly considered. Latin Hypecube sampling was used to assess the phenomenological and systemic uncertainties associated with the primary and confinement system responses to the core damage accident. The results of the analysis show that the N Reactor confinement concept provides significant radiological protection for most of the accident progression pathways studied

SCDAP: a light water reactor computer code for severe core damage analysis

International Nuclear Information System (INIS)

Marino, G.P.; Allison, C.M.; Majumdar, D.

1982-01-01

Development of the first code version (MODO) of the Severe Core Damage Analysis Package (SCDAP) computer code is described, and calculations made with SCDAP/MODO are presented. The objective of this computer code development program is to develop a capability for analyzing severe disruption of a light water reactor core, including fuel and cladding liquefaction, flow, and freezing; fission product release; hydrogen generation; quenched-induced fragmentation; coolability of the resulting geometry; and ultimately vessel failure due to vessel-melt interaction. SCDAP will be used to identify the phenomena which control core behavior during a severe accident, to help quantify uncertainties in risk assessment analysis, and to support planning and evaluation of severe fuel damage experiments and data. SCDAP/MODO addresses the behavior of a single fuel bundle. Future versions will be developed with capabilities for core-wide and vessel-melt interaction analysis

Molten Core - Concrete interactions in nuclear accidents. Theory and design of an experimental facility

International Nuclear Information System (INIS)

Sevon, T.

2005-11-01

In a hypothetical severe accident in a nuclear power plant, the molten core of the reactor may flow onto the concrete floor of containment building. This would cause a molten core . concrete interaction (MCCI), in which the heat transfer from the hot melt to the concrete would cause melting of the concrete. In assessing the safety of nuclear reactors, it is important to know the consequences of such an interaction. As background to the subject, this publication includes a description of the core melt stabilization concept of the European Pressurized water Reactor (EPR), which is being built in Olkiluoto in Finland. The publication includes a description of the basic theory of the interaction and the process of spalling or cracking of concrete when it is heated rapidly. A literature survey and some calculations of the physical properties of concrete and corium. concrete mixtures at high temperatures have been conducted. In addition, an equation is derived for conservative calculation of the maximum possible concrete ablation depth. The publication also includes a literature survey of experimental research on the subject of the MCCI and discussion of the results and deficiencies of the experiments. The main result of this work is the general design of an experimental facility to examine the interaction of molten metals and concrete. The main objective of the experiments is to assess the probability of spalling, or cracking, of concrete under pouring of molten material. A program of five experiments has been designed, and pre-test calculations of the experiments have been conducted with MELCOR 1.8.5 accident analysis program and conservative analytic calculations. (orig.)

Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

2015-05-15

The conventional PSA techniques cannot adequately evaluate all events. The conventional PSA models usually focus on single internal events such as DBAs, the external hazards such as fire, seismic. However, the Fukushima accident of Japan in 2011 reveals that very rare event is necessary to be considered in the PSA model to prevent the radioactive release to environment caused by poor treatment based on lack of the information, and to improve the emergency operation procedure. Especially, the results from PSA can be used to decision making for regulators. Moreover, designers can consider the weakness of plant safety based on the quantified results and understand accident sequence based on human actions and system availability. This study is for PSA modeling of combined accidents including total loss of feedwater (TLOFW) accident. The TLOFW accident is a representative accident involving the failure of cooling through secondary side. If the amount of heat transfer is not enough due to the failure of secondary side, the heat will be accumulated to the primary side by continuous core decay heat. Transients with loss of feedwater include total loss of feedwater accident, loss of condenser vacuum accident, and closure of all MSIVs. When residual heat removal by the secondary side is terminated, the safety injection into the RCS with direct primary depressurization would provide alternative heat removal. This operation is called feed and bleed (F and B) operation. Combined accidents including TLOFW accident are very rare event and partially considered in conventional PSA model. Since the necessity of F and B operation is related to plant conditions, the PSA modeling for combined accidents including TLOFW accident is necessary to identify the design and operational vulnerabilities.The PSA is significant to assess the risk of NPPs, and to identify the design and operational vulnerabilities. Even though the combined accident is very rare event, the consequence of combined

Disruption of Core Planar Cell Polarity Signaling Regulates Renal Tubule Morphogenesis but Is Not Cystogenic.

Science.gov (United States)

Kunimoto, Koshi; Bayly, Roy D; Vladar, Eszter K; Vonderfecht, Tyson; Gallagher, Anna-Rachel; Axelrod, Jeffrey D

2017-10-23

Oriented cell division (OCD) and convergent extension (CE) shape developing renal tubules, and their disruption has been associated with polycystic kidney disease (PKD) genes, the majority of which encode proteins that localize to primary cilia. Core planar cell polarity (PCP) signaling controls OCD and CE in other contexts, leading to the hypothesis that disruption of PCP signaling interferes with CE and/or OCD to produce PKD. Nonetheless, the contribution of PCP to tubulogenesis and cystogenesis is uncertain, and two major questions remain unanswered. Specifically, the inference that mutation of PKD genes interferes with PCP signaling is untested, and the importance of PCP signaling for cystogenic PKD phenotypes has not been examined. We show that, during proliferative stages, PCP signaling polarizes renal tubules to control OCD. However, we find that, contrary to the prevailing model, PKD mutations do not disrupt PCP signaling but instead act independently and in parallel with PCP signaling to affect OCD. Indeed, PCP signaling that is normally downregulated once development is completed is retained in cystic adult kidneys. Disrupting PCP signaling results in inaccurate control of tubule diameter, a tightly regulated parameter with important physiological ramifications. However, we show that disruption of PCP signaling is not cystogenic. Our results suggest that regulating tubule diameter is a key function of PCP signaling but that loss of this control does not induce cysts. Copyright © 2017 Elsevier Ltd. All rights reserved.

Assessment of accident risks in the CRBRP. Volume 2. Appendices

Energy Technology Data Exchange (ETDEWEB)

None

1977-03-01

Appendices to Volume I include core-related accident-sequence definition, CRBRP risk-assessment sequence-probability determinations, failure-probability data, accident scenario evaluation, radioactive material release analysis, ex-core accident analysis, safety philosophy and design features, calculation of reactor accident consequences, sensitivity study, and risk from fires.

MELCOR Severe Accident Analysis on the SMART Reactor

International Nuclear Information System (INIS)

Kim, Tae Woon; Jin, Young Ho; Kim, Young In; Kim, Keung Koo; Wang, Ziao; Revankar, Shripad

2014-01-01

A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

International Nuclear Information System (INIS)

Zhang Longfei; Zhang Dafa; Wang Shaoming

2007-01-01

Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

On the sequence of core-melt accidents: Fission product release, source terms and Chernobyl release

Energy Technology Data Exchange (ETDEWEB)

Albrecht, H

1986-01-01

There is a sketch of our ideas on the course of a core melt-out accident in a PWR. There is then a survey of the most important results on fission product release, which were obtained by experiments on the SASCHA melt-out plant. The 3rd part considers questions which are important for determining source terms for the environment and the last part contains some considerations on radioactivity release from the Chernobyl reactor.

Core to surge-line energy transport in a severe accident scenario

International Nuclear Information System (INIS)

Marzo, M. di; Almenas, K.; Gopalnarayanan, S.

1994-01-01

The analysis of loss of coolant accidents in a nuclear power plant, which progress to the stage where the core is uncovered, poses important safety related questions. One of these concerns the rate of energy transport to metal components of the primary system. An experimental program has been conducted at the Univ. of Maryland test facility which quantifies the rate of energy transfer from an uncovered core in a B ampersand W (once-through type steam generators) plant. SF 6 is used to simulate the natural circulation driving force of the high pressure steam expected at prototypical conditions. A time-dependent scaling methodology is developed to transpose experimental data to prototypical conditions. To achieve this transformation, a nominal fluid temperature increase rate of 1.0 degrees C/s is inferred from available TMI-2 event data. To bracket the range of potential prototypical transient scenarios, temperature ramps of 0.8 degrees C/s and 1.2 degrees C/s are also considered. Repeated tests, covering a range of test facility conditions, lead to estimated failure times at the surge line nozzle of 1.5 to 2 hours after initiation of the natural circulation phase of the transient

Post-accident fuel relocation and heat removal in the LMFBR

International Nuclear Information System (INIS)

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

1976-08-01

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

Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

2009-01-01

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

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

International Nuclear Information System (INIS)

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

1997-12-01

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

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

International Nuclear Information System (INIS)

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

1998-11-01

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

Analysis of loss of coolant accident and emergency core cooling system

International Nuclear Information System (INIS)

Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

1977-01-01

In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

The effects of applying silicon carbide coating on core reactivity of pebble-bed HTR in water ingress accident

Energy Technology Data Exchange (ETDEWEB)

Zuhair, S.; Setiadipura, Topan [National Nuclear Energy Agency of Indonesia, Serpong Tagerang Selatan (Indonesia). Center for Nuclear Reactor Technology and Safety; Su' ud, Zaki [Bandung Institute of Technology (Indonesia). Dept. of Physics

2017-03-15

Graphite is used as the moderator, fuel barrier material, and core structure in High Temperature Reactors (HTRs). However, despite its good thermal and mechanical properties below the radiation and high temperatures, it cannot avoid corrosion as a consequence of an accident of water/air ingress. Degradation of graphite as a main HTR material and the formation of dangerous CO gas is a serious problem in HTR safety. One of the several steps that can be adopted to avoid or prevent the corrosion of graphite by the water/air ingress is the application of a thin layer of silicon carbide (SiC) on the surface of the fuel element. This study investigates the effect of applying SiC coating on the fuel surfaces of pebble-bed HTR in water ingress accident from the reactivity points of view. A series of reactivity calculations were done with the Monte Carlo transport code MCNPX and continuous energy nuclear data library ENDF/B-VII at temperature of 1200 K. Three options of UO{sub 2}, PuO{sub 2}, and ThO{sub 2}/UO{sub 2} fuel kernel were considered to obtain the inter comparison of the core reactivity of pebble-bed HTR in conditions of water/air ingress accident. The calculation results indicated that the UO{sub 2}-fueled pebble-bed HTR reactivity was slightly reduced and relatively more decreased when the thickness of the SiC coating increased. The reactivity characteristic of ThO{sub 2}/UO{sub 2}-fueled pebble-bed HTR showed a similar trend to that of UO{sub 2}, but did not show reactivity peak caused by water ingress. In contrast with UO{sub 2}- and ThO{sub 2}-fueled pebble-bed HTR, although the reactivity of PuO{sub 2}-fueled pebble-bed HTR was the lowest, its characteristics showed a very high reactivity peak (0.33 Δk/k) and this introduction of positive reactivity is difficult to control. SiC coating on the surface of the plutonium fuel pebble has no significant impact. From the comparison between reactivity characteristics of uranium, thorium and plutonium cores with 0

On disruption of reactor core of the Chernobylsk-4 reactor (retrospective analysis of experiments and facts)

International Nuclear Information System (INIS)

Platonov, P.A.

2007-01-01

Fragments of graphite blocks from the damaged Chernobyl NPP, unit 4 are studied, the results are analyzed. The temperature of the graphite blocks at the moment of accident release from the reactor is evaluated. Results of studying the fragments of fuel channel and fuel dispersion are considered. The fuel heat content at the moment of the explosion is evaluated and some conclusions are made about the character of the reactor core destruction [ru

Severe accidents at nuclear power plants. Their risk assessment and accident management

International Nuclear Information System (INIS)

Abe, Kiyoharu.

1995-05-01

This document is to explain the severe accident issues. Severe Accidents are defined as accidents which are far beyond the design basis and result in severe damage of the core. Accidents at Three Mild Island in USA and at Chernobyl in former Soviet Union are examples of severe accidents. The causes and progressions of the accidents as well as the actions taken are described. Probabilistic Safety Assessment (PSA) is a method to estimate the risk of severe accidents at nuclear reactors. The methodology for PSA is briefly described and current status on its application to safety related issues is introduced. The acceptability of the risks which inherently accompany every technology is then discussed. Finally, provision of accident management in Japan is introduced, including the description of accident management measures proposed for BWRs and PWRs. (author)

Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

International Nuclear Information System (INIS)

Raymond, P.; Caruge, D.; Paik, H.J.

1994-01-01

The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

Accident termination by element dropout in the GCFR

International Nuclear Information System (INIS)

Torri, A.; Tomkins, J.L.

1976-01-01

Severe loss-of-flow accidents are being investigated for the GCFR in order to assess the risk from those low-probability accidents which lead to a loss of coolable core geometry. Accident mitigating phenomena unique to the GCFR have been identified for the loss of decay heat removal accident. Circumferential assembly duct melting is calculated to occur at the core mid-plane before the fuel within the assembly melts. The GCFR core assemblies are top-mounted and there is clearance between assemblies to accommodate swelling and thermal distortions without interference. No lateral core clamping system is employed and there are no structures in the plenum below the core. Thus it is possible for the lower portion of the individual assemblies, including most of the fuel, to drop to the cavity floor unless interference or bonding between assemblies develops during the accident. Due to the delay in duct corner melting the melt front at the duct mid-flat progresses over about one-half of the core height. The possibility of inter-element bonding by molten duct steel dislocated into the gap between assemblies has been recognized and a test program to verify the duct melting sequence and to investigate the duct dropout is being planned at the Los Alamos Scientific Laboratory

Analyzing the BWR rod drop accident in high-burnup cores

International Nuclear Information System (INIS)

Diamond, D.J.; Neymotin, L.; Kohut, P.

1995-01-01

This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ''rim'' effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions

Severe Accident Research Program plan update

International Nuclear Information System (INIS)

1992-12-01

In August 1989, the staff published NUREG-1365, ''Revised Severe Accident Research Program Plan.'' Since 1989, significant progress has been made in severe accident research to warrant an update to NUREG-1365. The staff has prepared this SARP Plan Update to: (1) Identify those issues that have been closed or are near completion, (2) Describe the progress in our understanding of important severe accident phenomena, (3) Define the long-term research that is directed at improving our understanding of severe accident phenomena and developing improved methods for assessing core melt progression, direct containment heating, and fuel-coolant interactions, and (4) Reflect the growing emphasis in two additional areas--advanced light water reactors, and support for the assessment of criteria for containment performance during severe accidents. The report describes recent major accomplishments in understanding the underlying phenomena that can occur during a severe accident. These include Mark I liner failure, severe accident scaling methodology, source term issues, core-concrete interactions, hydrogen transport and combustion, TMI-2 Vessel Investigation Project, and direct containment heating. The report also describes the major planned activities under the SARP over the next several years. These activities will focus on two phenomenological issues (core melt progression, and fuel-coolant interactions and debris coolability) that have significant uncertainties that impact our understanding and ability to predict severe accident phenomena and their effect on containment performance SARP will also focus on severe accident code development, assessment and validation. As the staff completes the research on severe accident issues that relate to current generation reactors, continued research will focus on efforts to independently evaluate the capability of new advanced light water reactor designs to withstand severe accidents

Application of the severe accident code ATHLET-CD. Modelling and evaluation of accident management measures (Project WASA-BOSS)

Energy Technology Data Exchange (ETDEWEB)

Wilhelm, Polina; Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Schaefer, Frank [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety

2016-07-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. Numerical analyses are used to investigate the accident progression and the complex physical phenomena during the core degradation phase, as well as to evaluate the effectiveness of possible countermeasures in the preventive and mitigative domain [1, 2]. The presented analyses have been performed with the computer code ATHLET-CD developed by GRS [3, 4].

Safety characteristics of mid-sized MOX fueled liquid metal reactor core of high converter type in the initiating phase of unprotected loss of flow accident. Effect of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

International Nuclear Information System (INIS)

Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

2003-07-01

Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System. An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1 mm with 1.2 m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120)was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un

Nuclear reactor safety. Quarterly progress report, October 1--December 31, 1977

International Nuclear Information System (INIS)

Jackson, J.F.; Stevenson, M.G.

1978-02-01

Progress in reactor safety research is summarized. LWR studies include TRAC code development for thermal-hydraulic analysis of accidents, containment systems evaluation, and safety experiments. LMFBR studies include SIMMER code development and applications, modeling of core disruptive accidents, and safety test facilities studies. HTGR safety studies cover fission product release and transport, structural evaluation, phenomena modeling, systems analysis, and accident delineation. GCFR studies are focussed on core disruptive testing

CINETHICA - Core accident analysis code

International Nuclear Information System (INIS)

Nakata, H.

1989-10-01

A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

Deterministic analyses of severe accident issues

International Nuclear Information System (INIS)

Dua, S.S.; Moody, F.J.; Muralidharan, R.; Claassen, L.B.

2004-01-01

Severe accidents in light water reactors involve complex physical phenomena. In the past there has been a heavy reliance on simple assumptions regarding physical phenomena alongside of probability methods to evaluate risks associated with severe accidents. Recently GE has developed realistic methodologies that permit deterministic evaluations of severe accident progression and of some of the associated phenomena in the case of Boiling Water Reactors (BWRs). These deterministic analyses indicate that with appropriate system modifications, and operator actions, core damage can be prevented in most cases. Furthermore, in cases where core-melt is postulated, containment failure can either be prevented or significantly delayed to allow sufficient time for recovery actions to mitigate severe accidents

Information on the Chernobyl NPP accident and its consequencies prepared for IAEA

Energy Technology Data Exchange (ETDEWEB)

1986-11-01

The information on the accident at the 4th power unit of the Chernobyl NPP and its consequences prepared for IAEA on the basis of the conclusions made by the Government commission constituted for investigating the accident causes and implementing the necessary emergency and reconstruction measures is given. The accident with reactor core disruption and partial destruction of the building Lappened on 26.04.86 at 1 hour and 23 minutes. The accident occurred before reactor shut-down for planned repairs during the testing of one of turbogenerators. The design features of the RBMK-1000 reactor plant, its main physical characteristics and parameters of the NPP safety system are considered. The chronology of the accident development and the results of analysis carried out using a mathematical model are given. The causes of the accident are analyzed. The measures for preventing the accident development and lessening its consequences as well as those for the environment radioactive contamination control and sanitary provisions are described in detail. The conclusion is made that the original cause of the accident is highly improbable combination of disorder and errors in operational conditions made by the personnel of the power unit. It is emphasized that development of the world nuclear engineering, besides advantages in the field of power supply and natural resources conservation, incurs also damages of international character. Among these are transboundary radioactivity transport, in particular, during serious radiation accidents and the danger of international terrorism and specific radiation hazard of nuclear objects under war conditions. All this defines the key necessity of deep international cooperation in the field of nuclear power engineering and its safeguarding.

Approach to accident management in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.; Urbonavicius, E.; Uspuras, E.

2008-01-01

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK. Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed. This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account

Daylight Saving Time Transitions and Road Traffic Accidents

Directory of Open Access Journals (Sweden)

Tuuli Lahti

2010-01-01

Full Text Available Circadian rhythm disruptions may have harmful impacts on health. Circadian rhythm disruptions caused by jet lag compromise the quality and amount of sleep and may lead to a variety of symptoms such as fatigue, headache, and loss of attention and alertness. Even a minor change in time schedule may cause considerable stress for the body. Transitions into and out of daylight saving time alter the social and environmental timing twice a year. According to earlier studies, this change in time-schedule leads to sleep disruption and fragmentation of the circadian rhythm. Since sleep deprivation decreases motivation, attention, and alertness, transitions into and out of daylight saving time may increase the amount of accidents during the following days after the transition. We studied the amount of road traffic accidents one week before and one week after transitions into and out of daylight saving time during years from 1981 to 2006. Our results demonstrated that transitions into and out of daylight saving time did not increase the number of traffic road accidents.

CARNSORE: Hypothetical reactor accident study

International Nuclear Information System (INIS)

Walmod-Larsen, O.; Jensen, N.O.; Kristensen, L.; Meide, A.; Nedergaard, K.L.; Nielsen, F.; Lundtang Petersen, E.; Petersen, T.; Thykier-Nielsen, S.

1984-06-01

Two types of design-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are described and their consequences assessed. The PLUCON 2 model was used to calculate the consequences which are presented in terms of individual and collective doses, as well as early and late health consequences. The site proposed for the nucelar power station is Carnsore Point, County Wexford, south-east Ireland. The release fractions for the accidents described are those given in WASH-1400. The analyses are based on the resident population as given in the 1979 census and on 20 years of data from the meteorological stations at Rosslare Harbour, 8.5 km north of the site. The consequences of one of the hypothetical core-melt accidents are described in detail in a meteorological parametric study. Likewise the consequences of the worst conceivable combination of situations are described. Finally, the release fraction in one accident is varied and the consequences of a proposed, more probable ''Class 9 accident'' are presented. (author)

The role of systems availability and operator actions in accident management

International Nuclear Information System (INIS)

Lutz, R.J. Jr.; Scobel, J.H.

1988-01-01

Traditional analyses of severe accidents, such as those presented in Probabilistic Risk Assessment (PRA) studies of nuclear power stations, have generally been performed on the assumption that all means of cooling the reactor core are lost and that no operator actions to mitigate the consequences or progression of the severe accident are performed. The assumption to neglect the availability of safety systems and operator actions which do not prevent core melting can lead to erroneous conclusions regarding the plant severe accident profile. Recent work in severe accident management has identified the need to perform analyses which consider all systems availabilities and operator actions, irrespective of their contribution to the prevention of core melting. These new analyses have far reaching conclusions. The analysis results indicate an unacceptably high degree of simplicity in the present severe accident analyses for Probabilistic Risk Assessment studies; the simplicity is in the assumption that systems availabilities and operator actions which do not impact core melt frequency can be neglected in the severe accident analyses. This results in overly pessimistic predictions of the time of core melting and the subsequent potential for recovery of core cooling prior to core melting. This simplicity can have a considerable impact on severe accident decision making, particularly in the evaluation of alternate plant design features and the priorities for research studies

Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

International Nuclear Information System (INIS)

Ginsberg, T.; Chen, J.C.

1984-01-01

Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

Simulant - water experiments to characterize the debris bed formed in severe core melt accidents

International Nuclear Information System (INIS)

Mathai, Amala M.; Anandan, J.; Sharma, Anil Kumar; Murthy, S.S.; Malarvizhi, B.; Lydia, G.; Das, Sanjay Kumar; Nashine, B.K.; Selvaraj, P.

2015-01-01

Molten Fuel Coolant Interaction (WO) and debris bed configuration on the core catcher plate assumes importance in assessing the Post Accident Heat Removal (PARR) of a heat generating debris bed. The key factors affecting the coolability of the debris bed are the bed porosity, morphology of the fragmented particles, degree of spreading/heaping of the debris on the core catcher and the fraction of lump formed. Experiments are conducted to understand the fragmentation kinetics and subsequent debris bed formation of molten woods metal in water at interface temperatures near the spontaneous nucleation temperature of water. Morphology of the debris particles is investigated to understand the fragmentation mechanisms involved. The spreading behavior of the debris on the catcher plate and the particle size distribution are presented for 5 kg and 10 kg melt inventories. Porosity of the undisturbed bed on the catcher plate is evaluated using a LASER sensor technique. (author)

Analysis of hypothetical LMFBR whole-core accidents in the USA

International Nuclear Information System (INIS)

Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

1978-01-01

Methods used for analysis of material behaviour, accident phenomenology and integrated accident calculations are reviewed. Applications of these methods to hypothetical LOF and TOP accidents are discussed. Recent results obtained from applications to FFTF and CRBRP are presented. (author)

Use of decision trees for evaluating severe accident management strategies in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclerar Engineering; Lee, Yongjin; Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of). School of Energy Systems Engineering

2016-07-15

Accident management strategies are defined to innovative actions taken by plant operators to prevent core damage or to maintain the sound containment integrity. Such actions minimize the chance of offsite radioactive substance leaks that lead to and intensify core damage under power plant accident conditions. Accident management extends the concept of Defense in Depth against core meltdown accidents. In pressurized water reactors, emergency operating procedures are performed to extend the core cooling time. The effectiveness of Severe Accident Management Guidance (SAMG) became an important issue. Severe accident management strategies are evaluated with a methodology utilizing the decision tree technique.

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

NPP Krsko Severe Accident Management Guidelines Implementation

International Nuclear Information System (INIS)

Basic, I.; Krajnc, B.; Bilic-Zabric, T.; Spiler, J.

2002-01-01

Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. The USA NRC has indicated that the development of a licensee plant specific accident management program will be required in order to close out the severe accident regulatory issue (Ref. SECY-88-147). Generic Letter 88-20 ties the Accident management Program to IPE for each plant. The SECY-89-012 defines those actions taken during the course of an accident by the plant operating and technical staff to: 1) prevent core damage, 2) terminate the progress of core damage if it begins and retain the core within the reactor vessel, 3) maintain containment integrity as long as possible, and 4) minimize offsite releases. The subject of this paper is to document the severe accident management activities, which resulted in a plant specific Severe Accident Management Guidelines implementation. They have been developed based on the Krsko IPE (Individual Plant Examination) insights, Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidances) and plant specific documents developed within this effort. Among the required plant specific actions the following are the most important ones: Identification and documentation of those Krsko plant specific severe accident management features (which also resulted from the IPE investigations). The development of the Krsko plant specific background documents (Severe Accident Plant Specific Strategies and SAMG Setpoint Calculation). Also, paper discusses effort done in the areas of NPP Krsko SAMG review (internal and external ), validation on Krsko Full Scope Simulator (Severe Accident sequences are simulated by MAAP4 in real time) and world 1st IAEA Review of Accident Management Programmes (RAMP). (author)

Development of severe accident analysis code - A study on the molten core-concrete interaction under severe accidents

Energy Technology Data Exchange (ETDEWEB)

Jung, Chang Hyun; Lee, Byung Chul; Huh, Chang Wook; Kim, Doh Young; Kim, Ju Yeul [Seoul National University, Seoul (Korea, Republic of)

1996-07-01

The purpose of this study is to understand the phenomena of the molten core/concrete interaction during the hypothetical severe accident, and to develop the model for heat transfer and physical phenomena in MCCIs. The contents of this study are analysis of mechanism in MCCIs and assessment of heat transfer models, evaluation of model in CORCON code and verification in CORCON using SWISS and SURC Experiments, and 1000 MWe PWR reactor cavity coolability, and establishment a model for prediction of the crust formation and temperature of melt-pool. The properties and flow condition of melt pool covering with the conditions of severe accident are used to evaluate the heat transfer coefficients in each reviewed model. Also, the scope and limitation of each model for application is assessed. A phenomenological analysis is performed with MELCOR 1.8.2 and MELCOR 1.8.3 And its results is compared with corresponding experimental reports of SWISS and SURC experiments. And the calculation is performed to assess the 1000 MWe PWR reactor cavity coolability. To improve the heat transfer model between melt-pool and overlying coolant and analyze the phase change of melt-pool, 2 dimensional governing equations are established using the enthalpy method and computational program is accomplished in this study. The benchmarking calculation is performed and its results are compared to the experiment which has not considered effects of the coolant boiling and the gas injection. Ultimately, the model shall be developed for considering the gas injection effect and coolant boiling effect. 66 refs., 10 tabs., 29 refs. (author)

Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

International Nuclear Information System (INIS)

Chupin, A.; Hu, L. W.; Buongiorno, J.

2008-01-01

Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

Bubble behavior in LMFBR core disruptive accidents. Annual report, June 1, 1975--June 30, 1976

International Nuclear Information System (INIS)

Reynolds, A.B.; Erdman, C.A.; Garner, P.L.; Kennedy, M.F.; Rao, S.P.; Refling, J.G.

1976-08-01

The work reported here is part of the Aerosol Release and Transport program for LMFBR safety assessment for the Reactor Safety Research Division of the U.S. Nuclear Regulatory Commission. Six areas were at various stages of investigation during this reporting period. A study of nonequilibrium mass transfer during fuel expansion and a study of the dynamics of fuel expansion into the sodium pool were completed. Studies are underway on condensation on above-core structures and on generation of aerosols from condensation. Studies were initiated on small-particle generation from hydrodynamic fragmentation, on particle kinematics and on particle-surface interaction

Nuclear accidents. Three mile Island (United States)

International Nuclear Information System (INIS)

Duco, J.

2004-01-01

This paper describes the accident of Three Miles Island power plant which occurred the 28 march 1979 in the United States. The accident scenario, the consequences and the reactor core and vessel, after the accident, are analyzed. (A.L.B.)

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)

Calculation of individual and population doses on Danish territory resulting from hypothetical core-melt accidents at the Barsebaeck reactor

International Nuclear Information System (INIS)

1977-01-01

Individual and population doses within Danish territory are calculated from hypothetical, severe core-melt accidents at the Swedish nuclear plant at Barsebaeck. The fission product inventory of the Barsebaeck reactor is calculated. The release fractions for the accidents are taken from WASH-1400. Based on parametric studies, doses are calculated for very unfavourable, but not incredible weather conditions. The probability of such conditions in combination with wind direction towards Danish territory is estimated. Doses to bone marrow, lungs, GI-tract and thyroid are calculated based on dose models developed at Risoe. These doses are found to be consistent with doses calculated with the models used in WASH-1400. (author)

Conditions for oxygen-deficient combustion during accidents with severe core concrete thermal attack

International Nuclear Information System (INIS)

Luangdilok, W.; Elicson, G.T.; Berger, W.E. Jr.

1993-01-01

This paper addresses the interactions between MCCI (molten core-concrete interactions)-induced offgas releases, mostly the combustible gases, natural circulation between the cavity and the lower containment based on recent research developments in the area of mixed convection flow (Epstein, et al., 1989; Epstein, 1988; Epstein, 1992) between compartments, and their effects on combustion in PWR containments during prolonged severe accidents. Specifically, large dry PWR containments undergoing severe core-concrete attack during station blackouts where the containment atmosphere is expected to be inerted are objects of this analysis. The purpose of this paper, given the conditions that oxygen can be brought to the cavity, is to demonstrate that consumption of most oxygen present in the containment can be achieved in a reasonable time scale assuming that combustion is not subject to flammability limits due to the high cavity temperatures. The conditions for cavity combustion depend on several factors including good gas flowpaths between the cavity and other containment regions, and combustion processes within the cavity with the hot debris acting as the ignition source

The management of severe accidents in modern pressure tube reactors

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Blahnik, C.; Snell, V.G.; Duffey, R.B.

2007-01-01

Advanced new reactor designs resist severe accidents through a balance between prevention and mitigation. This balance is achieved by designing to ensure that such accidents are very rare; and by limiting core damage progression and releases from the plant in the event of such rare accidents. These design objectives are supported by a suitable combination of probabilistic safety analysis, engineering judgment and experimental and analytical study. This paper describes the approach used for the Advanced CANDU Reactor TM -1000 (ACR-1000) design, which includes provisions to both prevent and mitigate severe accidents. The paper describes the use of PSA as a 'design assist' tool; the analysis of core damage progression pathways; the definition of the core damage states; the capability of the mitigating systems to stop and control severe accident events; and the severe accident management opportunities for consequence reduction. (author)

TMI-2 core examination plan

International Nuclear Information System (INIS)

Owen, D.E.; MacDonald, P.E.; Hobbins, R.R.; Ploggr, S.A.

1982-01-01

The Three Mile Island (TMI-2) core examination is divided into four stages: (1) before removing the head; (2) before removing the plenum; (3) during defueling; and (4) offsite examinations. Core examinations recommended during the first three stages are primarily devoted to documenting the post-accident condition of the core. The detailed analysis of core damage structures will be performed during offsite examinations at government and commercial hot cell facilities. The primary objectives of these examinations are to enhance the understanding of the degraded core accident sequence, to develop the technical bases for reactor regulations, and to improve LWR design and operation

Thermal-hydraulic uncertainties affecting severe accident progression

International Nuclear Information System (INIS)

Haskin, F.E.; Behr, V.L.

1984-01-01

To provide the proper technical bases for decisions regarding severe accidents, the US Nuclear Regulatory Commission (NRC) is sponsoring the following activities: (a) a variety of severe accident research programs, combined under the Severe Accident Research Plan; (b) nationwide task forces on containment loading, containment response, and fission product source terms; (c) a review by the American Physical Society of state-of-the-art methods for calculating radiological source terms; and (d) technical exchange meetings with the Industry Degraded Core (IDCOR) program. One of the means for integrating this developing array of technical information is the Severe Accident Risk Reduction Program (SARRP). One of the current SARRP objectives is to utilize insights gained from the activities listed above to characterize the relative likelihoods of competing containment failure modes for core-melt accidents

Severe accident recriticality analyses (SARA)

DEFF Research Database (Denmark)

Frid, W.; Højerup, C.F.; Lindholm, I.

2001-01-01

with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality-both super-prompt power bursts and quasi steady-state power......Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

On fission product retention in the core of the low powered high temperature reactor under accident conditions

International Nuclear Information System (INIS)

Bastek, H.

1984-01-01

In the core of the high temperature reactor the fuel element and the coated particles contained herein provide the safest enclosure for fission products. The complex process of fission product transport out of the particle kernel, through the particle coating and within the fuel element graphite is described in a simplified form by the Fick's diffusion. The effective diffusion coefficient is used for calculation. Starting from the existing ideas of fission product transport five burn-up and temperature-dependent diffusion coefficients for Cesium in (Th,U)O 2 -kernels are derived in this study. The results have been gained from several fuel element radiation experiments in recent years, which showed extreme variation in regard to burn-up, temperature cycle, neutron flux and operation time. Cs-137 release measurements from single particle kernels were present from all the experiments. Furthermore, annealing tests of AVR-fuel elements were analyzed. Heat-temperatur and heating-time, the fuel element burn-up in the AVR-reactor, as well as the measured Cs-137 inventory of the fuel elements before and after annealing, are included in the investigation as essential parameters. With the aid of the derived diffusion coeffizients and already present data sets the Cs-137 release of fuel elements into a small reactor core is investigated under unrestricted core heat-up. While the released Cs-137 is derived mainly from defective particles at accident temperatures up to 1600 0 C, the main part diffuses through the particle coating at higher accident temperatures. (orig./HP) [de

ASTRID core: Design objectives, design approach, and R&D in support

International Nuclear Information System (INIS)

Mignot, G.; Devictor, N.

2012-01-01

ASTRID core design is mainly guided by safety objectives: 1. Prevention of the core meltdown accident: To prevent meltdown accidents: - by a natural behavior of the core and the reactor (no actuation of the two shutdown systems); - with adding passive complementary systems if natural behavior is not sufficient for some transient cases. 2. Mitigation of the fusion accident: To garantee that core fusion accidents don’t lead to significant mechanical energy release, whatever initiator event: - by a natural core behavior; - with adding specific mitigation dispositions in case of natural behavior is not suffficient

Mood Fluctuation and Psychobiological Instability: The Same Core Functions Are Disrupted by Novel Psychoactive Substances and Established Recreational Drugs

Directory of Open Access Journals (Sweden)

Andrew C. Parrott

2018-03-01

Full Text Available Many novel psychoactive substances (NPS have entered the recreational drug scene in recent years, yet the problems they cause are similar to those found with established drugs. This article will debate the psychobiological effects of these newer and more traditional substances. It will show how they disrupt the same core psychobiological functions, so damaging well-being in similar ways. Every psychoactive drug causes mood states to fluctuate. Users feel better on-drug, then feel worse off-drug. The strength of these mood fluctuations is closely related to their addiction potential. Cyclical changes can occur with many other core psychobiological functions, such as information processing and psychomotor speed. Hence the list of drug-related impairments can include: homeostatic imbalance, HPA axis disruption, increased stress, altered sleep patterns, neurohormonal changes, modified brain rhythms, neurocognitive impairments, and greater psychiatric vulnerability. Similar patterns of deficit are found with older drugs such as cocaine, nicotine and cannabis, and newer substances such as 3,4-methylenedioxymethamphetamine (MDMA, mephedrone and spice. All psychoactive drugs damage human well-being through similar basic neuropsychobiological mechanisms.

Fundamental study on flow characteristics of disrupted core pool at a low energy level (Joint research)

International Nuclear Information System (INIS)

Morita, Koji; Liu, Ping; Matsumoto, Tatsuya; Fukuda, Kenji; Tobita, Yoshiharu; Yamano, Hidemasa; Sato, Ikken

2009-09-01

Dynamic behaviors of solid-particle dominant multiphase flows were investigated to model the mobility of core materials in a low-energy disrupted core of a liquid metal fast reactor. Two series of experiments were performed, those were dam-break experiments and bubble visualization experiments. Verification of fluid-dynamics models used in the fast reactor safety analysis code SIMMER-III was also conducted based on the numerical simulations of these experiments. The experimental analyses show that SIMMER-III can represent effects of solid particle interaction on multiphase flow behaviors by adjusting model parameters of the particle jamming model if the particles are immersed in liquid phase. Further improvement of SIMMER-III with more generalized models is necessary to appropriately simulate interactions between solid particles in a wider range of flow conditions. (author)

Modelling of RPV lower head under core melt severe accident condition using OpenFOAM

International Nuclear Information System (INIS)

Madokoro, Hiroshi; Kretzschmar, Frank; Miassoedov, Alexei

2017-01-01

Although six years have been passed since the tragic severe accident at Fukushima Daiichi, still large uncertainties exist in modeling of core degradation and reactor pressure vessel (RPV) failure. It is extremely important to obtain a better understanding of complex phenomena in the lower head in order to improve accident management measures. The possible failure mode of reactor pressure vessel and its failure time are especially a matter of importance. Thermal behavior of the molten pool can be simulated by the Phase-change Effective Convectivity Model (PECM), which is a distributed-parameter model developed in the Royal Institute of Technology (KTH), Sweden. The model calculates convective currents not using a pure CFD approach but based on so called “characteristic velocities” that are determined by empirical correlations depending on the geometry and physical properties of the molten pool. At the Karlsruhe Institute of Technology (KIT), the PECM has been implemented in the open-source CFD software OpenFOAM in order to receive detailed predictions of a core melt behavior in the RPV lower head under severe accident conditions. An advantage of using OpenFOAM is that it is very flexible to add and modify models and physical properties. In the current work, the solver is extended to couple PECM with a structure analysis model of the vessel wall. The model considers thermal expansion, plasticity, creep and damage. The model and physical properties are based on those implemented in ANSYS. Although the previous implementation had restriction that the amount of and geometry of the melt cannot be changed, our coupled model allows flexibility of the melt amount and geometry. The extended solver was used to simulate the LIVE-L1 and -L7V experiments and has demonstrated good prediction of the temperature distribution in the molten pool and heat flux distribution through the vessel wall. Regarding the vessel failure the model was applied to one of the FOREVER tests

Comparative analysis of a hypothetical loss-of-flow accident in an irradiated LMFBR core using different computer models for a common benchmark problem

International Nuclear Information System (INIS)

Wider, H.U.; Devos, J.; Nguyen, H.; Goethem, G. Van.; Miles, K.J.; Tentner, A.M.; Pizzica, P.

1989-01-01

This report summarizes the results of an international exercise to compare whole-core accident calculations of the initiation phase of an unprotected LOF accident in a large irradiated LMFBR. The results for the accident phase before pin failure are in rather good agreement except for the fuel pin mechanics predictions. There are also some differences in the sodium boiling calculations but the voiding rates which are of key importance are very similar. The post - failure fuel motion and sodium voiding predictions show significant differences. However, the majority of these calculations agree that temporary fuel accumulations occur which increase the power beyond that caused by sodium voiding alone

Evaluation of final vapor pressures in the loss of flow accident in an irradiation device of a pool reactor core

International Nuclear Information System (INIS)

Verri, A.

1987-01-01

The reliability feature, are described for a device containing samples, at a temperatures of 300 grade centigrades, in a reactor core for a long time. After an examination of the maximum accident event, the maximum vapour pressure originated by the inlet of reactor cooling water into the experimental device, is evaluated

Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Generino, G.

1984-07-01

Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

Assessment of two BWR accident management strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Petek, M.

1991-01-01

A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

Fuel temperature analysis method for channel-blockage accident in HTTR

International Nuclear Information System (INIS)

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

1994-01-01

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

A safety design approach for sodium cooled fast reactor core toward commercialization in Japan

International Nuclear Information System (INIS)

Kubo, Shigenobu

2012-01-01

JAEA’s safety approach for SFR core design is based on defence‐in‐depth concept, which includes DBAs and DECs (prevention and mitigation): • The reactor core is designed to have inherent reactivity feedback characteristics with negative power coefficient. • Operation temperature range is set sufficiently below the coolant boiling temperature so as to avoid coolant boiling against anticipated operational occurrences and DBAs. • If the plant state deviates from operational states, the safe reactor shutdown is achieved by automatic insertion of control rods. 2 active reactor shutdown systems are provided. • Failure of active reactor shutdown is assumed in a design extension condition . Passive shutdown capability is provided by SASS under such condition. • As a design extension condition, core disruptive accident is assumed. In order to prevent severe mechanical energy release which might cause containment function failure, core sodium void worth is limited below 6 dollars and molten fuel discharge capability is utilized by FAIDUS. (author)

Severe accident training simulator APROS SA

International Nuclear Information System (INIS)

Raiko, Eerikki; Salminen, Kai; Lundstroem, Petra; Harti, Mika; Routamo, Tomi

2003-01-01

APROS SA is a severe accident training simulator based on the APROS simulation environment. APROS SA has been developed in Fortum Nuclear Services Ltd to serve as a training tool for the personnel of the Loviisa NPP. Training with APROS SA gives the personnel a deeper understanding of the severe accident phenomena and thus it is an important part of the implementation of the severe accident management strategy. APROS SA consists of two parts, a comprehensive Loviisa plant model and an external severe accident model. The external model is an extension to the Loviisa plant model, which allows the simulation to proceed into the severe accident phase. The severe accident model has three submodels: the core melting and relocation model, corium pool model and fission product model. In addition to these, a new thermal-hydraulic solver is introduced to the core region of the Loviisa plant model to replace the more limited APROS thermal-hydraulic solver. The full APROS SA training simulator has a graphical user interface with visualizations of both severe accident management panels at the operator room and the important physical phenomena during the accident. This paper describes the background of the APROS SA training simulator, the severe accident submodels and the graphical user interface. A short description how APROS SA will be used as a training tool at the Loviisa NPP is also given

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

International Nuclear Information System (INIS)

Alonso, A.

1980-01-01

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

Present status and needs of research on severe core damage

International Nuclear Information System (INIS)

1982-05-01

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

A commentary on the current status and the future role of the European accident code

International Nuclear Information System (INIS)

Butland, A.T.D.

1990-01-01

This paper describes the history of the project to produce the European Accident code (EAC), leading to the planned release of a version of EAC-2 at the end of 1989. The requirements of a computer code to model the initiation phase of Hypothetical Core Disruptive Accidents (HCDAs) are discussed, paying particular attention to the lessons learnt in the CABRI project. The current status and content of the EAC-2 code are examined in relation to these requirements, noting how the sophisticated modelling plans for EAC-2 make it a benchmark code. The validation status of EAC-2 and future plans are discussed, noting that currently it consists solely of stand-alone validation of the modules used in EAC-2, rather than validation of the combined code. The future role of EAC-2 is briefly discussed in relation to the fast reactor plans in the EEC countries. (author)

Evaluation of materials for retention of sodium and core debris in reactor systems. Annual progress report, September 1977-December 1978

International Nuclear Information System (INIS)

Swanson, D.G.; Zehms, E.H.; McClelland, J.D.; Meyer, R.A.; van Paassen, H.L.L.

1978-12-01

This report considers some of the consequences of a hypothetical core disruptive accident in a nuclear reactor. The interactions expected between molten core debris, liquid sodium, and materials that might be employed in an ex-vessel sacrificial-bed or in the reactor building are discussed. Experimental work performed for NRC by Sandia Laboratories and Hanford Engineering Development Laboratory on the interactions between liquid sodium and basalt concrete is reviewed. Studies of molten steel interactions with concrete at Sandia Laboratories and molten UO 2 interactions with concrete at The Aerospace Corporation are also discussed. The potential of MgO for use in core containment is discussed and refractory materials other than MgO are reviewed. Finally, results from earlier experiments with molten core debris and various materials performed at The Aerospace Corporation are presented

Uncertainties and severe-accident management

International Nuclear Information System (INIS)

Kastenberg, W.E.

1991-01-01

Severe-accident management can be defined as the use of existing and or alternative resources, systems, and actions to prevent or mitigate a core-melt accident. Together with risk management (e.g., changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-indepth safety philosophy for severe accidents. A significant number of probabilistic safety assessments have been completed, which yield the principal plant vulnerabilities, and can be categorized as (a) dominant sequences with respect to core-melt frequency, (b) dominant sequences with respect to various risk measures, (c) dominant threats that challenge safety functions, and (d) dominant threats with respect to failure of safety systems. Severe-accident management strategies can be generically classified as (a) use of alternative resources, (b) use of alternative equipment, and (c) use of alternative actions. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These include (a) uncertainty in key phenomena, (b) uncertainty in operator behavior, (c) uncertainty in system availability and behavior, and (d) uncertainty in information availability (i.e., instrumentation). This paper focuses on phenomenological uncertainties associated with severe-accident management strategies

Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR

International Nuclear Information System (INIS)

Rakhi; Sharma, Anil Kumar; Velusamy, K.

2017-01-01

Highlights: • Heat transfer enhancement using multi-tray core catcher for SFR is investigated. • The capability of a single core collector tray is estimated. • Double and triple collector trays with innovative designs is discussed. • Provision of openings in the trays contributed to enhanced natural circulation. - Abstract: To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the

Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

International Nuclear Information System (INIS)

Ginsberg, T.; Chen, J.C.

1984-01-01

Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench (removal of stored energy from initial temperature to saturation temperature) of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

γ radiation level simulation and analysis with MCNP in EPR containment during severe accident

International Nuclear Information System (INIS)

Zeng Jun; Liu Shuhuan; Wang Yang; Zhai Liang

2013-01-01

The γ dosimetry model based on the EPR core structure, material composition and the designed shielding system was established. The γ-ray dose rate distributions in EPR containment under different conditions including normal operation state, loss-of-coolant accident and core melt severe accident were simulated with MCNP5, and the calculation results under normal operation state and severe accident were compared and analyzed respectively with that of the designed limit. The study results may provide some relative data reference for EPR core accident prediction and reactor accident emergency decision making. (authors)

Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

International Nuclear Information System (INIS)

Gustavsson, Veine

2002-11-01

The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

1986-01-01

The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

Directory of Open Access Journals (Sweden)

Dalin Zhang

2012-06-01

Full Text Available Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium during a core disruptive accident (CDA may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within the CP-ESFR European project, focusing on the phenomena affecting the initiation and the transition phases of an unprotected loss of flow (ULOF accident. Key phenomena for the initiation phase are coolant boiling onset and further voiding of the core that lead to a reactivity increase in the case of a positive void reactivity effect. Therefore, the first level of optimization involves the reduction, by design, of the positive void effect in order to avoid entering a severe accident. If the core disruption cannot be avoided, the accident enters into the transition phase, characterized by the progression of core melting and recriticalities due to fuel compaction. Dedicated features that enhance and guarantee a sufficient and timely fuel discharge are considered for the optimization of this phase.

In-plant considerations for optimal offsite response to reactor accidents

International Nuclear Information System (INIS)

Burke, R.P.; Heising, C.D.; Aldrich, D.C.

1982-11-01

Offsite response decision-making methods based on in-plant conditions are developed for use during severe reactor-accident situations. Dose projections are used to eliminate all LWR plant systems except the reactor core and the spent-fuel storage pool from consideration for immediate offsite emergency response during accident situations. A simple plant information-management scheme is developed for use in offsite response decision-making. Detailed consequence calculations performed with the CRAC2 model are used to determine the appropriate timing of offsite-response implementation for a range of PWR accidents involving the reactor core. In-plant decision criteria for offsite-response implementation are defined. The definition of decision criteria is based on consideration of core-accident physical processes, in-plant accident monitoring information, and results of consequence calculations performed to determine the effectiveness of various public-protective measures. The benefits and negative aspects of the proposed response-implementation criteria are detailed

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Status of degraded core issues. Synthesis paper prepared by G. Bandini in collaboration with the NEA task group on degraded core cooling

International Nuclear Information System (INIS)

2001-02-01

The in-vessel evolution of a severe accident in a nuclear reactor is characterised, generally, by core uncover and heat-up, core material oxidation and melting, molten material relocation and debris behaviour in the lower plenum up to vessel failure. The in-vessel core melt progression involves a large number of physical and chemical phenomena that may depend on the severe accident sequence and the reactor type under consideration. Core melt progression has been studied in the last twenty years through many experimental works. Since then, computer codes are being developed and validated to analyse different reactor accident sequences. The experience gained from the TMI-2 accident also constitutes an important source of data. The understanding of core degradation process is necessary to evaluate initial conditions for subsequent phases of the accident (ex-vessel and within the containment), and define accident management strategies and mitigative actions for operating and advanced reactors. This synthesis paper, prepared within the Task Group on Degraded Core Cooling (TG-DCC) of PWG2, contains a brief summary of current views on the status of degraded core issues regarding light water reactors. The in-vessel fission product release and transport issue is not addressed in this paper. The areas with remaining uncertainties and the needs for further experimental investigation and model development have been identified. The early phase of core melt progression is reasonably well understood. Remaining uncertainties may be addressed on the basis of ongoing experimental activities, e.g. on core quenching, and research programs foreseen in the near future. The late phase of core melt progression is less understood. Ongoing research programs are providing additional valuable information on corium molten pool behaviour. Confirmatory research is still required. The pool crust behaviour and material relocation into the lower plenum are the areas where additional research should

In-core monitoring detectors

International Nuclear Information System (INIS)

Mitelman, M.G.

2001-01-01

The main task of in-core monitoring consists in securing observability of the reactor installation in all possible operation modes (normal, transient, accident and post-accident). Operation safety at acceptable cost can be achieved by optimized measurement errors. The range of sensors applied as in-core detectors for operative measurements in the industry is very limited in number. Among them might be cited self powered neutron detectors (SPND) and thermocouples. Sensors are incorporated in the in-core detectors assemblies (SVRD). The presentation makes an effort to touch upon the problems of assuring and increasing quality of in-core on-line measurements. So we do not consider systems using movable detectors, as the latter do not assure on-line measurements. (Authors)

NPP Krsko Severe Accident Management Guidelines Upgrade

International Nuclear Information System (INIS)

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

2014-01-01

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

Severe accident simulation at Olkiuoto

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Tirkkonen, H.; Saarenpaeae, T. [Teollisuuden Voima Oy (TVO), Olkiluoto (Finland); Cliff Po, L.C. [Micro-Simulation Technology, Montville, NJ (United States)

1995-09-01

A personal computer-based simulator was developed for the Olkiluoto nuclear plant in Finland for training in severe accident management. The generic software PCTRAN was expanded to model the plant-specific features of the ABB Atom designed BWR including its containment over-pressure protection and filtered vent systems. Scenarios including core heat-up, hydrogen generation, core melt and vessel penetration were developed in this work. Radiation leakage paths and dose rate distribution are presented graphically for operator use in diagnosis and mitigation of accidents. Operating on an graphically for operator use in diagnosis and mitigation of accidents. Operating on an 486 DX2-66, PCTRAN-TVO achieves a speed about 15 times faster than real-time. A convenient and user-friendly graphic interface allows full interactive control. In this paper a review of the component models and verification runs are presented.

Fundamental study on flow characteristics of disrupted core pool at a low energy level (Joint research)

International Nuclear Information System (INIS)

Morita, Koji; Liu, Ping; Matsumoto, Tatsuya; Fukuda, Kenji; Tobita, Yoshiharu; Sato, Ikken

2007-03-01

Dynamic behaviors of solid particle beds in a liquid pool against pressure transients were investigated to model the mobility of core materials in a low-energy disrupted core of a liquid metal fast reactor. A series of experiments was performed with a particle bed of different heights, comprising different monotype solid particles, where variable initial pressures of the originally pressurized nitrogen gas were adopted as the pressure source. Computational simulations of the experiments were performed using SIMMER-III, a fast reactor safety analysis code. Experimental analyses using the SIMMER-III code show that physical models and method used in the code can reasonably represent the transient behaviors of multiphase flows with rich solid phase as observed in the experiments. The validation of several key models of SIMMER-III was also discussed for treating transient behaviors of the solid-particle phase in multiphase flows. (author)

A study of entrainment at a break and in the core during small break loss-of-coolant accidents in PWRs

International Nuclear Information System (INIS)

Yonomoto, Taisuke

1996-05-01

Objectives of the present study are to obtain a better understanding of entrainment at a break and in the core during small break loss-of-coolant-accidents (SBLOCAs) in PWRs, and to develop a means for the best evaluation of the phenomena. For the study of entrainment at a break, a theoretical model was developed, which was assessed by comparisons with several experimental data bases. By modifying a LOCA analysis code using the present model, experimental results obtained from SBLOCA experiments at a PWR large-scale simulator were reproduced very well. For the study of entrainment in the core, reflooding experiments were conducted at high pressure, from which the onset conditions were obtained. It was confirmed that the cooling behavior for a dry-out core is very simple under typical high pressure reflooding conditions for PWRs, because liquid entrainment does not occur in the core. (author)

Accident management: What is it and how do you do it?

International Nuclear Information System (INIS)

Henry, Robert E.; Hammersley, Robert J.

2004-01-01

Accident management is the composite of those actions that would prevent, stop and/or mitigate a severe accident in a nuclear power plant. Since they act to prevent core damage, the Emergency Operating Procedures (EOPs) are an integral part of accident management. Each of the Owners Groups have developed EOPs that are well thought out for instructing the operator to respond to accident conditions which could threaten the core. However, for those very low probability events in which the core could be uncovered and damaged, accident management actions arise from a logical evaluation of possible actions (strategies) for recovering from the accident state and protecting the public health and safety. To understand the character of accident management it is first necessary to define: 1. What is threatened as a result of the accident? 2. Fundamentally, what needs to be protected? 3. What is known during an accident? 4. What have we learned from the TMI-2 accident? 5. What have we learned from the plant specific IPEs? Once these subjects are reviewed on a utility specific and plant specific basis, accident management actions become relatively straightforward and likely can be effectively addressed using the total capability available in a given design. This paper discusses these five questions in a global manner with the aim being to aid plant specific implementation. (author)

Statistical inference of the nuclear accidents occurrence number for the next decade

International Nuclear Information System (INIS)

Felizia, E.R.

1987-01-01

This paper aims to give a response using the classical statistical and bayesian inference techniques regarding the common characteristic in the Harrisburg and Chernobyl nuclear accidents: in both reactors, core fusion occurred. In relation to the last mentioned techniques, the most recent developments were applied, based on the decision theory of uncertainty; among others, the principle of maximum entropy. Besides, as a preliminar information on the accidents occurrence frequency with core fusion, the German risk analysis results were used. The estimations predicted for the next decade an average between one or two accidents with core fusion and low possibilities for the 'no accident' event in the same period. (Author)

Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

International Nuclear Information System (INIS)

Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

1995-01-01

Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

Analyses of severe accident scenarios in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.; Rimkevicius, S.; Uspuras, E.; Urbonavicius, E.

2006-01-01

Even though research of severe accidents in light water reactors is performed around the world for several decades many questions remain. Research is mostly performed for vessel-type reactors. RBMK is a channel type light water reactor, which differs from the vessel-type reactors in several aspects. These differences impose some specifics in the accident phenomena and processes that occur during severe accidents. Severe accident research for RBMK reactors is taking first steps and very little information is available in the open literature. The existing severe accident analysis codes are developed for vessel-type reactors and their application to the analysis of accidents in RBMK is not straightforward. This paper presents the results of an analysis of large loss-of-coolant accident scenarios with loss of coolant injection to the core of RBMK-1500. The analysis performed considers processes in the reactor core, in the reactor cooling system and in the confinement until the fuel melting started. This paper does not aim to answer all the questions regarding severe accidents in RBMK but rather to start a discussion, identify the expected timing of the key phenomena. (orig.)

Risk reduction of core-melt accidents in advaned CAPRA burner cores

International Nuclear Information System (INIS)

Maschek, W.; Struwe, D.; Eigemann, M.

1997-01-01

As part of the CAPRA Program (Consommation Accrue de Plutonium dans les RApides) the feasibility of fast reactors is investigated to burn plutonium and also to destruct minor actinides. The design of CAPRA cores shows significant differences compared to conventional cores. Especially the high Pu-enrichment has an important influence on the core melt-down behavior and the associated recriticality risk. To cope with this risk, inherent design features and special measures/devices are investigated for their potential of early fuel discharge to reduce the criticality of the reactor core. An assessment of such measures/devices is given and experimental needs are formulated. 11 refs., 5 figs

Severe Accident Recriticality Analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hoejerup, F. [Risoe National Lab. (Denmark); Lindholm, I.; Miettinen, J.; Puska, E.K. [VTT Energy, Helsinki (Finland); Nilsson, Lars [Studsvik Eco and Safety AB, Nykoeping (Sweden); Sjoevall, H. [Teoliisuuden Voima Oy (Finland)

1999-11-01

Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B{sub 4}C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

Severe Accident Recriticality Analyses (SARA)

International Nuclear Information System (INIS)

Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Puska, E.K.; Nilsson, Lars; Sjoevall, H.

1999-11-01

Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B 4 C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding

Analyses of systems availability and operator actions to support the development of severe accident procedures

International Nuclear Information System (INIS)

Lutz, R.J. Jr.; Scobel, J.H.

1989-01-01

This paper reports on traditional analyses of severe accidents, such as those presented in Probabilistic Risk Assessment (PRA) studies of nuclear power stations, that have generally been performed on the assumption that all means of cooling the reactor core are lost and that no operator actions to mitigate the consequences or progression of the severe accident are performed. The assumption to neglect the availability of safety systems and operator actions which do not prevent core melting can lead to erroneous conclusions regarding the plant severer accident profile. Recent work in severe accident management has identified the need to perform analyses which consider all systems availabilities and operator actions, irrespective of their contribution to the prevention of core melting. These new analyses indicate that the traditional analyses result in overfly pessimistic predictions of the time of core melting and the subsequent potential for recovery of core cooling prior to core melting. Additionally, since the traditional analyses do not model all of the operator actions which are prescribed, the impact of additional severe accident operator actions on the progression and consequences of the accident cannot be reliably identified. Further, the more detailed analysis can change the focus of the importance of various system to the prevention of core damage and the mitigation of severe accident consequences. Finally, the simplicity of the traditional analyses can have a considerable impact on severe accident decision making, particularly in the evaluation of alternate plant design features and the priorities for research studies

Severe accident management. Optimized guidelines and strategies

International Nuclear Information System (INIS)

Braun, Matthias; Löffler, Micha; Plank, Hermann; Asse, Dietmar; Dimmelmeier, Harald

2014-01-01

The highest priority for mitigating the consequences of a severe accident with core melt lies in securing containment integrity, as this represents the last barrier against fission product release to the environment. Containment integrity is endangered by several physical phenomena, especially highly transient phenomena following high-pressure reactor pressure vessel failure (like direct containment heating or steam explosions which can lead to early containment failure), hydrogen combustion, quasi-static over-pressure, temperature failure of penetrations, and basemat penetration by core melt. Each of these challenges can be counteracted by dedicated severe accident mitigation hardware, like dedicated primary circuit depressurization valves, hydrogen recombiners or igniters, filtered containment venting, containment cooling systems, and core melt stabilization systems (if available). However, besides their main safety function these systems often have also secondary effects that need to be considered. Filtered containment venting causes (though limited) fission product release into the environment, primary circuit depressurization leads to loss of coolant, and an ex-vessel core melt stabilization system as well as hydrogen igniters can generate high pressure and temperature loads on the containment. To ensure that during a severe accident any available systems are used to their full beneficial extent while minimizing their potential negative impact, AREVA has implemented a severe accident management for German nuclear power plants. This concept makes use of extensive numerical simulations of the entire plant, quantifying the impact of system activations (operational systems, safety systems, as well as dedicated severe accident systems) on the accident progression for various scenarios. Based on the knowledge gained, a handbook has been developed, allowing the plant operators to understand the current state of the plant (supported by computational aids), to predict

Evaluation of containment failure modes and fission product releases during core meltdown accidents in a BWR with a Mark III containment

International Nuclear Information System (INIS)

Ludewig, H.; Yu, W.S.; Jaung, R.; Pratt, W.T.

1985-01-01

An assessment is described of potential failure modes and fission product releases for a large number of postulated core meltdown accidents in a BWR with a Mark III containment. For this containment design, the most important failure mode was found to be due to hydrogen related phenomena. A one-dimensional lumped parameter computer code has been developed and used to determine the probability of various hydrogen phenomena for a range of postulated core meltdown sequences. Potential containment loads have been estimated and compared against the containment capacity to determine the probability of containment failure. The fission product release assessment began by using the MARCH/CORRAL system of codes with key input parameters varied over a reasonable range. The parameters relate to primary system retention, re-emission, pool scrubbing, and fission product release in-vessel vs ex-vessel. The final step used more mechanistic calculations based on the system of codes recently developed under sponsorship of the Accident Source Term Program Office, NRC, and compares these predictions with the range of releases calculated in the sensitivity study

Dominant accident sequences in Oconee-1 pressurized water reactor

International Nuclear Information System (INIS)

Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

1985-04-01

A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

Prevention and mitigation of severe accidents

International Nuclear Information System (INIS)

Weisshaeupl, H.

1996-01-01

For the European Pressurized water Reactor (EPR), jointly developed by French and German industry, great emphasis is laid to gain further improvement in prevention of severe accidents based on the accumulative experience and proven technology of the French and German PWR reactors. In this evolutionary development, a balanced and comprehensive approach in respect to implement new passive features has been chosen. Improvements in each step of the defense in depth concept lead to a further decrease in the probability of occurrence of a severe accident with partial or even gross melting of the core. The different phenomenons that occur during such an hypothetical accident must be taken into account during the conception of specific measurements necessary to mitigate accident consequences. To cope with the consequences of a severe accident with core melt down means to deal with different phenomena which may threaten the integrity of the containment or may lead to an enhanced fission product release into the environment: high pressure reactor pressure vessel failure; energetic molten fuel coolant interaction; direct containment heating, molten core concrete interaction; hydrogen combustion; long term pressure and temperature increase in the containment. The EPR approach follows the recommendations from the DFD (Deutsch-Franzosischer Direktionsausschuss), jointly prepared by the French and German safety authorities. The EPR concept consist to prevent or eliminate as far as possible scenarios which are connected with high loads (high pressure failure of the reactor pressure vessel, or global hydrogen detonation etc..) by dedicated design provisions, and to deal with the consequences of severe accident scenarios which are not ruled out by specific safety measures. The measures comprise: the primary system depressurization; the control of hydrogen; the stabilisation and cooling of the melted core; the containment heat removal. They are completed by specific characteristics

The management of severe accidents

International Nuclear Information System (INIS)

Pelce, J.; Brignon, P.

1987-01-01

In considering severe accidents in water power reactors, a major problem that arises is how to manage them in such a way that the situation can be controlled as well as possible, from the aspects both of preventing serious damage to the core of limiting the discharge of radioactivity. A number of countries have announced provisions in the field of accident management, some already set up, others planned, but these mainly apply to preventing damage to the core. Part of this report deals with this aspect, to show that there is a fairly wide consensus on how problems should be approached. Attitudes vary, on the other hand, in the approach to mitigate radioactive release. In fact, few countries have proposed concrete steps to manage severe accidents in the final stages when the core is seriously damaged. Since it is difficult to compare different approaches, only the French approach is described. This description is however very brief, because in the five or six years since it was defined, the approach has been presented many times. The stress is placed more on the comments which this type of approach suggests, to make the subsequent general discussion easier

Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code; Verbundprojekt WASA-BOSS: Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt B: Druckwasserreaktor-Stoerfallanalysen unter Verwendung des Severe-Accident-Codes ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

2017-02-15

Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

Licensing aspects in the verification of the SNR 300 design concept against hypothetical accidents

International Nuclear Information System (INIS)

Kugler, E.; Wiesner, S.

1976-01-01

The German prototype of a fast breeder reactor, the SNR 300, is being built near Kalkar on the Lower Rhine. It is a loop-type fast sodium-cooled reactor, designed and constructed by Interatom, Bensberg. Experiences gained from the first phase of construction are described. The report is restricted to the aspects of the SNR 300 design against a core disruptive accident (CDA) and its consequences and to the difficulties having arisen in the verification of the design concept so far. Some examples of the detailed design are described and discussed from the licensing authority's point of view showing that the difficulties have been typical for a prototype reactor subjected to a regular licensing procedure

Development of high performance core for large fast breeder reactors

International Nuclear Information System (INIS)

Inoue, Kotaro; Kawashima, Katsuyuki; Watari, Yoshio.

1982-01-01

Subsequently to the fast breeder prototype reactor ''Monju'', the construction of a demonstration reactor with 1000 MWe output is planned. This research aims at the establishment of the concept of a large core with excellent fuel breeding property and safety for a demonstration and commercial reactors. For the purpose, the optimum specification of fuel design as a large core was clarified, and the new construction of a core was examined, in which a disk-shaped blanket with thin peripheral edge is introduced at the center of a core. As the result, such prospect was obtained that the time for fuel doubling would be 1/2, and the energy generated in a core collapse accident would be about 1/5 as compared with a large core using the same fuel as ''Monju''. Generally, as a core is enlarged, the rate of breeding lowers. If a worst core collapse accident occurs, the scale of accident will be very large in the case of a ''Monju'' type large core. In an unhomogeneous core, an internal blanket is provided in the core for the purpose of improving the breeding property and safety. Hitachi Ltd. developed the concept of a large core unhomogeneous in axial direction and proposed it. The research on the fuel design for a large core, an unhomogeneous core and its core collapse accident is reported. (Kako, I.)

Environmental Impact Assessment following a Nuclear Accident to a Candu NPP

International Nuclear Information System (INIS)

Margeanu, C.A.; Margeanu, S.; Olteanu, Gh.

2009-01-01

The paper presents calculations of nuclear accident consequences to public and environment, for a Candu NPP using advanced fuel in two hypothetical accident scenarios: (1) large LOCA followed by partial core melting with early containment failure; (2) late core disassembly and containment bypass through ECCS. During both accidents a release occurs, radioactive contaminants being dispersed into atmosphere. As reference, estimations for Candu standard UO 2 fuel were used. The radioactive core inventory was obtained by using ORIGEN-S computer code included in ORNL,SCALE 5 programs package. Radiological consequences assessment to public and environment was performed by means of PC COSYMA computer code

Nuclear reactor core catcher

International Nuclear Information System (INIS)

1977-01-01

A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

Inherent safety features of the HTTR revealed in the accident condition

International Nuclear Information System (INIS)

Kunitomi, K.; Shinozaki, M.; Baba, O.; Saito, S.

1992-01-01

The High Temperature Engineering Test Reactor (HTTR) being constructed by JAERI (Japan Atomic Energy Research Institute) is a graphite-moderated and helium-cooled reactor with an outlet gas temperature of 950degC. The inherent safety characteristics in the HTTR prevent temperature increase of reactor fuels and fission product release from the reactor core in postulated accident conditions. The reactor core can be cooled by a Vessel Cooling System (VCS) indirectly, even in the case that no forced cooling is expected during the accident such as primary pipe break. The VCS consists of independent water cooling loop and cooling panel around the reactor pressure vessel. The cooling panel whose temperature of 60-90degC cools the reactor pressure vessel by radiation and removes the decay heat from the core indirectly. Furthermore, even if failure of VCS is assumed during this accident as a severe accident, the reactor core is remained safe despite the temperature increase of biological concrete shield around the reactor pressure vessel. This paper describes the inherent safety features of the HTTR specially focused on the accident condition without forced cooling. The detailed analytical results of such an accident are described together with clarifying the role of the VCS. (author)

Severe accidents: in nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

A ''severe'' nuclear accident refers to a reactor accident that could exceed reactor design specifications to such a degree as to prevent cooling of the reactor's core by normal means. This report summarizes the work of a NEA Senior Group of Experts who have studied the potential response of existing light-water reactors to severe accidents and have found that current designs of reactors are far more capable of coping with severe accidents than design specifications would suggest. The report emphasises the specific knowledge and means that can be used for diagnosing a severe accident and for managing its progression in order to prevent or mitigate its consequences

Methods to prevent the source term of methyl lodide during a core melt accident

Energy Technology Data Exchange (ETDEWEB)

Karhu, A. [VTT Energy (Finland)

1999-11-01

The purpose of this literature review is to gather available information of the methods to prevent a source term of methyl iodide during a core melt accident. The most widely studied methods for nuclear power plants include the impregnated carbon filters and alkaline additives and sprays. It is indicated that some deficiencies of these methods may emerge. More reactive impregnants and additives could make a great improvement. As a new method in the field of nuclear applications, the potential of transition metals to decompose methyl iodide, is introduced in this review. This area would require an additional research, which could elucidate the remaining questions of the reactions. The ionization of the gaseous methyl iodide by corona-discharge reactors is also shortly described. (au)

Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core

International Nuclear Information System (INIS)

Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

1985-01-01

The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO 2 and UO 2 /metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO 2 crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process

Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm2 in the cold leg of primary loop using RELAP5 code

International Nuclear Information System (INIS)

Borges, Eduardo M.; Sabundjian, Gaiane

2017-01-01

The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm 2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

In vessel core melt progression phenomena

International Nuclear Information System (INIS)

Courtaud, M.

1993-01-01

For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis

Melt Fragmentation Characteristics of Metal Fuel with Melt Injection Mass during Initiating Phase of SFR Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo; Lee, Min Ho; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of)

2016-05-15

The PGSFR has adopted the metal fuel for its inherent safety under severe accident conditions. However, this fuel type is not demonstrated clearly yet under the such severe accident conditions. Additional experiments for examining these issues should be performed to support its licensing activities. Under initiating phase of hypothetic core disruptive accident (HCDA) conditions, the molten metal could be better dispersed and fragmented into the coolant channel than in the case of using oxide fuel. This safety strategy provides negative reactivity driven by a good dispersion of melt. If the coolant channel does not sufficient coolability, the severe recriticality would occur within the core region. Thus, it is important to examine the extent of melt fragmentation. The fragmentation behaviors of melt are closely related to a formation of debris shape. Once the debris shape is formed through the fragmentation process, its coolability is determined by the porosity or thermal conductivity of the melt. There were very limited studies for transient irradiation experiments of the metal fuel. These studies were performed by Transient Reactor Test Facility (TREAT) M series tests in U.S. The TREAT M series tests provided basic information of metal fuel performance under transient conditions. The effect of melt injection mass was evaluated in terms of the fragmentation behaviors of melt. These behaviors seemed to be similar between single-pin and multi-pins failure condition. However, the more melt was agglomerated in case of multi-pins failure.

Hydrogen behavior in a large-dry pressurized water reactor containment building during a severe accident

International Nuclear Information System (INIS)

Hsu Wensheng; Chen Hungpei; Hung Zhenyu; Lin Huichen

2014-01-01

Following severe accidents in nuclear power plants, large quantities of hydrogen may be generated after core degradation. If the hydrogen is transported from the reactor vessel into the containment building, an explosion might occur, which might threaten the integrity of the building; this can ultimately cause the release of radioactive materials. During the Fukushima Daiichi nuclear accident in 2011, the primary containment structures remained intact but contaminated fragments broke off the secondary containment structures, which disrupted mitigation activities and triggered subsequent explosions. Therefore, the ability to predict the behavior of hydrogen after severe accidents may facilitate the development of effective nuclear reactor accident management procedures. The present study investigated the behavior of hydrogen in a large-dry pressurized water reactor (PWR). The amount of hydrogen produced was calculated using the Modular Accident Analysis Program. The hydrogen transport behavior and the effect of the explosion on the PWR containment building were simulated using the Flame Acceleration Simulator. The simulation results showed that the average hydrogen volume fraction is approximately 7% in the containment building and that the average temperature is 330 K. The maximum predicted pressure load after ignition is 2.55 bar, which does not endanger the structural integrity of the containment building. The results of this investigation indicate that the hydrogen mitigation system should be arranged on both the upper and lower parts of the containment building to reduce the impact of an explosion. (author)

Preliminary analysis of control rod accidents in the CRCN-R1 multipurpose reactor core of Recife in Brazil

International Nuclear Information System (INIS)

Souza dos Santos, Rubens; Rubens Maiorino, Jose

1999-01-01

The paper shows some results of the neutronic accident analyses arisen by uncontrolled control rod withdrawal, based on the Conceptual Project of the CRCN-R1 MultiPurpose Reactor of Recife. In that reactor, a project of the CNEN/Brazil, under the leadership of the IPEN/Sao Paulo, is verified the thermal hydraulic limits in the reactor core during transients that simulate startup and power operation accidents. It has utilized a computer program that solved the kinetic equations based on multigroup diffusion theory, in our case we have used 4 energy groups, Two-Dimensional X-Y in the space, and 6 groups of delayed neutrons. A simple model of feedback is admitted in the capture and scattering macroscopic cross sections, in the fuel regions, temperature and coolant densities dependents. Based on those models, the results demonstrated that the reactor exhibits good degree of safety. (author)

Managing water addition to a degraded core

International Nuclear Information System (INIS)

Kuan, P.; Hanson, D.J.; Odar, F.

1992-01-01

In this paper the authors present information that can be used in severe accident management by providing an improved understanding of the effects of water addition to a degraded core. This improved understanding is developed using a diagram showing a sequence of core damage states. Whenever possible, a temperature and a time after accident initiation are estimated for each damage state in the sequence diagram. This diagram can be used to anticipate the evolution of events during an accident. Possible responses of plant instruments are described to identify these damage states and the effects of water addition. The rate and amount of water addition needed (a) to remove energy from the core, (b) to stabilize the core or (c) to not adversely affect the damage progression, are estimated. Analysis of the capability to remove energy from large cohesive and particulate debris beds indicates that these beds may not be stabilized in the core region and they may partially relocate to the lower plenum of the reactor vessel

Severe accident considerations for modern KWU-PWR plants

International Nuclear Information System (INIS)

Eyink, J.

1987-01-01

In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

Modeling of reflood of severely damaged reactor core

International Nuclear Information System (INIS)

Bachrata, A.

2012-01-01

The TMI-2 accident and recently Fukushima accident demonstrated that the nuclear safety philosophy has to cover accident sequences involving massive core melt in order to develop reliable mitigation strategies for both, existing and advanced reactors. Although severe accidents are low likelihood and might be caused only by multiple failures, accident management is implemented for controlling their course and mitigating their consequences. In case of severe accident, the fuel rods may be severely damaged and oxidized. Finally, they collapse and form a debris bed on core support plate. Removal of decay heat from a damaged core is a challenging issue because of the difficulty for water to penetrate inside a porous medium. The reflooding (injection of water into core) may be applied only if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ from original rod bundle geometry and will resemble to the severe damaged core observed in TMI-2. The higher temperatures and smaller hydraulic diameters in a porous medium make the coolability more difficult than for intact fuel rods under typical loss of coolant accident conditions. The modeling of this kind of hydraulic and heat transfer is a one of key objectives of this. At IRSN, part of the studies is realized using an European thermo-hydraulic computer code for severe accident analysis ICARE-CATHARE. The objective of this thesis is to develop a 3D reflood model (implemented into ICARE-CATHARE) that is able to treat different configurations of degraded core in a case of severe accident. The proposed model is characterized by treating of non-equilibrium thermal between the solid, liquid and gas phase. It includes also two momentum balance equations. The model is based on a previously developed model but is improved in order to take into account intense boiling regimes (in particular

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.)

A review of the core catcher design in LMR

International Nuclear Information System (INIS)

Lee, Yong Bum; Hahn, Do Hee

2001-08-01

The overwhelming emphasis in reactor safety is on the prevention of core meltdown. Moreover, although there have been several accidents that have resulted in some fuel melting, to date there have been no accidents severe enough to cause the syndrome of core collapse, reactor vessel melt-through, containment penetration, and dispersal into the ground. Nevertheless, a number of proposals have been made for the design of core catcher systems to control or stop the motion of the molten core mass should such an accident take place. Core catchers may differ in both their location within the reactor system and in the mechanism that is used to cool and control the motion of the core debris. In this report the classification, configuration and main features of the core catcher are described. And also, The core catcher design technologies and processes are presented. Finally the core catcher provisions in constructed and planned LMRs (Liquid Metal Reactors) are summarized and the preliminary assessment on the core catcher installation in KALIMER is presented

Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

Energy Technology Data Exchange (ETDEWEB)

Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code

International Nuclear Information System (INIS)

Jobst, Matthias; Kliem, Soeren; Kozmenkov, Yaroslav; Wilhelm, Polina

2017-02-01

Within the framework of the project an ATHLET-CD input deck for a generic German PWR of type KONVOI has been created. This input deck was applied to the simulation of severe accidents from the accident categories station blackout (SBO) and small-break loss-of-coolant accidents (SBLOCA). The complete accident transient from initial event at full power until the damage of reactor pressure vessel (RPV) is covered and all relevant severe accident phenomena are modelled: start of core heat up, fission product release, melting of fuel and absorber material, oxidation and release of hydrogen, relocation of molten material inside the core, relocation to the lower plenum, damage and failure of the RPV. The model has been applied to the analysis of preventive and mitigative accident management measures for SBO and SBLOCA transients. Therefore, the measures primary side depressurization (PSD), injection to the primary circuit by mobile pumps and for SBLOCA the delayed injection by the cold leg hydro-accumulators have been investigated and the assumptions and start criteria of these measures have been varied. The time evolutions of the transients and time margins for the initiation of additional measures have been assessed. An uncertainty and sensitivity study has been performed for the early phase of one SBO scenario with PSD (until the start of core melt). In addition to that, a code -to-code comparison between ATHLET-CD and the severe accident code MELCOR has been carried out.

Screening and analysis of beyond design basis accident of 49-2 SPR

International Nuclear Information System (INIS)

Zhang Yadong; Guo Yue; Wu Yuanyuan; Zou Yao

2015-01-01

The beyond design basis accident was analyzed to ensure safe operation of 49-2 Swimming Pool Reactor (SPR) after design life. Because it's difficult to use PSA method, the unconditional assumed severe accidents were adopted to obtain a conservative result. The main conclusions were obtained by analyzing anticipated transients without scram in station blackout (SBO ATWS), horizontal channel rupture, core uncovering after shutdown and emergency response capacity. The results show that the core is safe in SBO ATWS, and the fuel elements will not melt as long as the core are not exposed in 2.5 h in loss of coolant accident caused by horizontal channel rupture and other factors. The passive siphon breaker function and various ways of emergency core makeup can ensure that the core is not exposed. (authors)

Evaluation of strategies for severe accident prevention and mitigation

International Nuclear Information System (INIS)

Tokarz, R.

1989-01-01

The NRC is planning to establish regulatory oversight on severe accident management capability in the US nuclear reactor industry. Accident management includes certain preparatory and recovery measures that can be taken by the plant operating and technical personnel to prevent or mitigate the consequences of a severe accident. Following an initiating event, accident management strategies include measures to (1) prevent core damage, (2) arrest the core damage if it begins and retain the core inside the vessel, (3) maintain containment integrity if the vessel is breached, and (4) minimize offsite releases. Objectives of the NRC Severe Accident Management Program are to assure that technically sound strategies are identified and guidance to implement these strategies is provided to utilities. This paper will describe work performed to date by Pacific Northwest Laboratory (PNL) and Battelle Memorial Institute (BMI) relative to severe accident strategy evaluation, as well as work to be performed and expected results. Working with Brookhaven National Laboratory, PNL evaluated a series of NRC suggested accident management strategies. The evaluation of these strategies was divided between PNL and Brookhaven National Laboratory and a similar paper will be presented by Brookhaven regarding their strategy evaluation. This paper will stress the overall safety issues related to the research and emphasize the strategies that are applicable to major safety issues. The relationship of these research activities to other projects is discussed, as well as planning for future changes in the direction of work to be undertaken

Study of diluting and absorber materials to control the reactivity during a postulated core meltdown accident in generation IV reactors

International Nuclear Information System (INIS)

Plevacova, Kamila

2010-01-01

In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic points of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author)

Study of diluting and absorber materials to control reactivity during a postulated core melt down accident in Generation IV reactors

International Nuclear Information System (INIS)

Plevacova, K.

2010-01-01

In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic point of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, a volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author) [fr

A condensed review of the core catcher in the LMR

International Nuclear Information System (INIS)

Lee, Yong Bum; Hahn, Do hee

2001-03-01

The overwhelming emphasis in reactor safety is on the prevention of core meltdown. Moreover, although there have been several accidents that have resulted in some fuel melting, to date there have been no accidents severe enough to cause the syndrome of core collapse, reactor vessel melt-through, containment penetration, and dispersal into the ground. Nevertheless, a number of proposals have been made for the design of core catcher systems to control or stop the motion of the molten core mass should such an accident take place. Core catchers may differ in both their location within the reactor system and in the mechanism that is used to cool and control the motion of the core debris. In this report the classification, configuration and main features of the core catcher are described. And also, the core catcher provisions in constructed and planned LMRs (Liquid Metal Reactors) are summarized

OSSA. A second generation of severe accident management

International Nuclear Information System (INIS)

Sauvage, E.C.; Musoyan, G.; Ducros, V.D.

2009-01-01

Nowadays the severe accident and their management are an integrated part of the new generation of power plants. The EPR, as the third generation of nuclear plants, includes both systems and instrumentation to mitigate a severe accident, but also a new generation of severe accident management guidelines: the OSSA. Severe accident management guidelines are highly dependent on human means available: emergency organization actors, training and knowledge shall be taken in consideration in an innovative way. Their impacts on ergonomy and content of the document lead to a new generation of guidelines with several innovative features. This second generation of severe accident management guidelines was developed in parallel with the PSA level 2, the human reliability analyses, the validation and verification process, the severe accident simulator progresses. By taking in consideration this variety of input the OSSA were developed in a user aspect orientation. For example in the OSSA a larger responsibility is given to the operational crew to better support the technical support group evaluation. Their existing knowledge of the plant and of the systems and instrumentation is used. This collaboration work implies a strong communication tool that has been developed to enhance the permanent communication within the emergency organization, but although to ensure the main up-to-date information for evaluation will be available where required. The entry condition is based on a strong and stand alone diagnostic for all plant states, that uses in particular a curve of core exit temperature as a function of primary pressure for a fixed core cladding temperature, or its equivalent in term of containment conditions. It ensures relatively consistent core conditions on entry. A first criterion for ultimate final primary depressurization is provided, ensuring all attempts to reflood the core with the available means have been ensured before the OSSA entry condition is reached. This

Severe accident recriticality analyses (SARA)

Energy Technology Data Exchange (ETDEWEB)

Frid, W. E-mail: wiktor.frid@ski.se; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H

2001-11-01

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s{sup -1} injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g{sup -1}, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s{sup -1}. In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated

Severe accident recriticality analyses (SARA)

International Nuclear Information System (INIS)

Frid, W.; Hoejerup, F.; Lindholm, I.; Miettinen, J.; Nilsson, L.; Puska, E.K.; Sjoevall, H.

2001-01-01

Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s -1 injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g -1 , was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s -1 . In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady

Generalization of Nuclear Safety and Course of Accident Events Research in the Ignalina NPP

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2001-01-01

The safety analysis shown that after implementation of SAR recommendations Ignalina NPP is adequately protected against accidents which required fast initiation of automatic protections. In case of accidents with long-term loss of core cooling additional operator actions are required. Accident management in case long-term core cooling are analyzed in this paper. (author)

Hepatitis B virus core protein allosteric modulators can distort and disrupt intact capsids.

Science.gov (United States)

Schlicksup, Christopher John; Wang, Joseph Che-Yen; Francis, Samson; Venkatakrishnan, Balasubramanian; Turner, William W; VanNieuwenhze, Michael; Zlotnick, Adam

2018-01-29

Defining mechanisms of direct-acting antivirals facilitates drug development and our understanding of virus function. Heteroaryldihydropyrimidines (HAPs) inappropriately activate assembly of hepatitis B virus (HBV) core protein (Cp), suppressing formation of virions. We examined a fluorophore-labeled HAP, HAP-TAMRA. HAP-TAMRA induced Cp assembly and also bound pre-assembled capsids. Kinetic and spectroscopic studies imply that HAP-binding sites are usually not available but are bound cooperatively. Using cryo-EM, we observed that HAP-TAMRA asymmetrically deformed capsids, creating a heterogeneous array of sharp angles, flat regions, and outright breaks. To achieve high resolution reconstruction (HAP-TAMRA caused quasi-sixfold vertices to become flatter and fivefold more angular. This transition led to asymmetric faceting. That a disordered crosslink could rescue symmetry implies that capsids have tensegrity properties. Capsid distortion and disruption is a new mechanism by which molecules like the HAPs can block HBV infection. © 2017, Schlicksup et al.

Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

1986-02-01

The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

Insights from Severe Accident Analyses for Verification of VVER SAMG

Energy Technology Data Exchange (ETDEWEB)

Gaikwad, A. J.; Rao, R. S.; Gupta, A.; Obaidurrahaman, K., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

2014-10-15

The severe accident analyses of simultaneous rupture of all four steam lines (case-a), simultaneous occurrence of LOCA with SBO (case-b) and Station blackout (case-c) were performed with the computer code ASTEC V2r2 for a typical VVER-1000. The results obtained will be used for verification of sever accident provisions and Severe Accident Management Guidelines (SAMG). Auxiliary feed water and emergency core cooling systems are modelled as boundary conditions. The ICARE module is used to simulate the reactor core, which is divided into five radial regions by grouping similarly powered fuel assemblies together. Initially, CESAR module computes thermal hydraulics in primary and secondary circuits. As soon as core uncovery begins, the ICARE module is actuated based on certain parameters, and after this, ICARE module computes the thermal hydraulics in the core, bypass, downcomer and the lower plenum. CESAR handles the remaining components in the primary and secondary loops. CPA module is used to simulate the containment and to predict the thermal-hydraulic and hydrogen behaviour in the containment. The accident sequences were selected in such a way that they cover low/high pressure and slow/fast core damage progression events. Events simulated included slow progression events with high pressure and fast accident progression with low primary pressure. Analysis was also carried out for the case of SBO with the opening of the PORVs when core exit temperature exceeds certain value as part of SAMG. Time step sensitivity study was carried out for LOCA with SBO. In general the trends and magnitude of the parameters are as expected. The key results of the above analyses are presented in this paper. (author)

EAC european accident code. A modular system of computer programs to simulate LMFBR hypothetical accidents

International Nuclear Information System (INIS)

Wider, H.; Cametti, J.; Clusaz, A.; Devos, J.; VanGoethem, G.; Nguyen, H.; Sola, A.

1985-01-01

One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat transfer together with coolant thermohydraulics in single- and two-phase flow. Temperature variations in fuel, coolant and neighbouring structures induce, in fact, thermal reactivity feedbacks which are added up and put in the neutronics calculation to predict the neutron flux and the subsequent heat generation in the reactor. At this point a whole-core analysis code is necessary to examine for any hypothetical transient whether the various feedbacks result effectively in a negative balance, which is the basis condition to ensure stability and safety. The European Accident Code (EAC), developed at the Joint Research Centre of the CEC at Ispra (Italy), fulfills this objective. It is a modular informatics structure (quasi 2-D multichannel approach) aimed at collecting stand-alone computer codes of neutronics, fuel pin mechanics and hydrodynamics, developed both in national laboratories and in the JRC itself. EAC makes these modules interact with each other and produces results for these hypothetical accidents in terms of core damage and total energy release. 10 refs

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

1992-01-01

This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

Identification and assessment of BWR in-vessel severe accident mitigation strategies

Energy Technology Data Exchange (ETDEWEB)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

Identification and assessment of BWR in-vessel severe accident mitigation strategies

International Nuclear Information System (INIS)

Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

1992-10-01

This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

Compendium of ECCS [Emergency Core Cooling Systems] research for realistic LOCA [loss-of-coolant accidents] analysis: Final report

International Nuclear Information System (INIS)

1988-12-01

In the United States, Emergency Core Cooling Systems (ECCS) are required for light water reactors (LWRs) to provide cooling of the reactor core in the event of a break or leak in the reactor piping or an inadvertent opening of a valve. These accidents are called loss-of-coolant accidents (LOCA), and they range from small leaks up to a postulated full break of the largest pipe in the reactor cooling system. Federal government regulations provide that LOCA analysis be performed to show that the ECCS will maintain fuel rod cladding temperatures, cladding oxidation, and hydrogen production within certain limits. The NRC and others have completed a large body of research which investigated fuel rod behavior and LOCA/ECCS performance. It is now possible to make a realistic estimate of the ECCS performance during a LOCA and to quantify the uncertainty of this calculation. The purpose of this report is to summarize this research and to serve as a general reference for the extensive research effort that has been performed. The report: (1) summarizes the understanding of LOCA phenomena in 1974; (2) reviews experimental and analytical programs developed to address the phenomena; (3) describes the best-estimate computer codes developed by the NRC; (4) discusses the salient technical aspects of the physical phenomena and our current understanding of them; (5) discusses probabilistic risk assessment results and perspectives, and (6) evaluates the impact of research results on the ECCS regulations. 736 refs., 412 figs., 66 tabs

Severe Accident R and D for Enhanced CANDU-6 Reactors

International Nuclear Information System (INIS)

Nitheanandan, Thambiayah

2012-01-01

CANDU reactors possess a number of inherent of inherent and designed safety features that make them resistant to core damage accidents. The unique feature is the low temperature moderator surrounding the fuel channels, which can serve as an alternate heat sink. The fuel is surrounded by three water systems: heavy water primary coolant, heavy water moderator, and light water calandria vault and shield water. In addition, the liquid inventory in the steam generators is a fourth indirect heat sink, able to cool the primary coolant. The water inventories in the emergency core cooling system and the reserve water tank at the dome of the containment can also provide fuel cooling and water makeup to prevent severe core damage or mitigate the consequences of a severe core damage accident. An assessment of the adequacy of the existing severe accident knowledge base, to confidently perform consequence analyses for the Enhanced CANDU-6 reactor in compliance with regulatory requirements, was recently completed. The assessment relied on systematic Phenomena Identification and Ranking Tables (PIRT) studies completed domestically and internationally. The assessment recommends cost-effective R and D to mitigate the consequences of severe accidents and associated risk vulnerabilities

Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm{sup 2} in the cold leg of primary loop using RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: borges.em@hotmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm{sup 2} of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

Addressing severe accidents in the CANDU 9 design

International Nuclear Information System (INIS)

Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

1998-01-01

CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

TMI-2 core examination

International Nuclear Information System (INIS)

Hobbins, R.R.; MacDonald, P.E.; Owen, D.E.

1983-01-01

The examination of the damaged core at the Three Mile Island Unit 2 (TMI-2) reactor is structured to address the following safety issues: fission product release, transport, and deposition; core coolability; containment integrity; and recriticality during severe accidents; as well as zircaloy cladding ballooning and oxidation during so-called design basis accidents. The numbers of TMI-2 components or samples to be examined, the priority of each examination, the safety issue addressed by each examination, the principal examination techniques to be employed, and the data to be obtained and the principal uses of the data are discussed in this paper

Power Excursion Accident Analysis of Research Water Reactor

International Nuclear Information System (INIS)

Khaled, S.M.; Doaa, G.M.

2009-01-01

A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the sub-channel thermal-hydraulic core analysis code SV based on the Single Mass Velocity Model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/SV for the accident analysis. The Training and Research Reactors at Budapest University of Technology and Economics (BME-Reactor) has been taken as a reference reactor. The cross-section generation procedure based on WIMS. The code uses an implicit difference approach for both the diffusion equations and thermal-hydraulics modules, with reactivity feedback effects due to coolant and fuel temperatures. The code system was applied to analyzing power excursion accidents initiated by ramp reactivity insertion of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt is expected even if the safety rods do not fall into the core

Severe accidents in nuclear reactors

International Nuclear Information System (INIS)

Ohai, Dumitru; Dumitrescu, Iulia; Tunaru, Mariana

2004-01-01

The likelihood of accidents leading to core meltdown in nuclear reactors is low. The consequences of such an event are but so severe that developing and implementing of adequate measures for preventing or diminishing the consequences of such events are of paramount importance. The analysis of major accidents requires sophisticated computation codes but necessary are also relevant experiments for checking the accuracy of the predictions and capability of these codes. In this paper an overview of the severe accidents worldwide with definitions, computation codes and relating experiments is presented. The experimental research activity of severe accidents was conducted in INR Pitesti since 2003, when the Institute jointed the SARNET Excellence Network. The INR activity within SARNET consists in studying scenarios of severe accidents by means of ASTEC and RELAP/SCDAP codes and conducting bench-scale experiments

Analysis of core and core barrel heat-up under conditions simulating severe reactor accidents

International Nuclear Information System (INIS)

Chellaiah, S.; Viskanta, R.; Ranganathan, P.; Anand, N.K.

1987-01-01

This paper reports on the development of a model for estimating the temperature distributions in the reactor core, core barrel, thermal shield and reactor pressure vessel of a PWR during an undercooling transient. A number of numerical calculations simulating the core uncovering of the TMI-2 reactor and the subsequent heat-up of the core have been performed. The results of the calculations show that the exothermic heat release due to Zircaloy oxidation contributes to the sharp heat-up of the core. However, the core barrel temperature rise which is driven by the temperature increase of the edge of the core (e.g., the core baffle) is very modest. The maximum temperature of the core barrel never exceeded 610 K (at a system pressure of 68 bar) after a 75 minute simulation following the start of core uncovering

Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

2009-06-15

Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

A description of nuclear reactor accidents and their consequences

International Nuclear Information System (INIS)

Murray, A.

1989-01-01

Nuclear reactor accidents which have caused core damage, released a significant amount of radioactivity, or caused death or serious injury are described. The reactor accidents discussed in detail include Chernobyl, Three Mile Island, SL-1 and Windscale, although information on other less consequential accidents is also provided. The consequences of these accidents are examined in terms of the amounts of radioactivity released, the radiation doses received, and remedial actions and interventions taken following the accident. 10 refs., 1 fig., 2 tabs

Material problems in accident analysis of prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Bazant, Z.P.

1977-01-01

Due to their very high energy absorption capability, as well as their inherent safety advantages, prestressed concrete reactor vessels are presently being keenly studied as the basic barrier to contain hypothetical core disruptive accidents in a fast breeder reactor. One problem investigated is the nonlinear constitutive behavior and failure criteria for concrete. Previously, a comprehensive theory, called endochronic theory, has been shown to satisfy all basic currently known features of test data. Nevertheless uncertainty still exists with regard to non-proportional loading paths, for which good test data are lacking at present. An extension of the endochronic theory which correlates best with general experimental evidence and includes fracturing terms is given, and a comparison with vertex-type hardening in plasticity is made. A second problem which must be analysed in accident situations is the high temperature shock on the concrete walls (due to liquid sodium, up to 850 0 C). Refining a previous crude formulation, a rational model for calculating moisture and heat transfer and pore pressures in concrete subjected to thermal shock is presented. In conclusion, a new design concept, in which the concrete vessel is completely dehydrated and kept hot throughout its service life in order to substantially improve its response to thermal shock as well as liquid sodium contact, is described. (Auth.)

Perspectives on phenomenology and simulation of severe accident in light water reactors

International Nuclear Information System (INIS)

Sugimoto, Jun

2014-01-01

Severe accident phenomena in light water reactors (LWRs) are generally characterized by their physically and chemically complex processes involved with high temperature core melt, multi-component and multi-phase flows, transport of radioactive materials and sometimes highly non-equilibrium state. Severe accident phenomenology is usually categorized into four phases; (1) fuel degradation, (2) in-vessel phenomena, (3) ex-vessel phenomena and (4) fission product release and transport. Among these, ex-vessel phenomena consist of five subcategories; 1) direct containment heating, 2) fuel coolant interaction (steam explosion), 3) molten core concrete interaction, 4) hydrogen behaviour and control and 5) containment failure/leakage. In the field of simulation of severe accident, severe accident analytical codes have been developed in the United States, EU and Japan, such as MAAP, MELCOR, ASTEC, THALES and SAMPSON. Many different kinds of analytical codes for the specific severe accident phenomena have also been developed worldwide. After the accident at Fukushima Daiichi Nuclear Power Station, review of severe accident research issues has been conducted and several issues are reconsidered, such as effects of BWR core degradation behaviors, sea water injection, pool scrubbing under rapid depressurization, containment failure/leakage and re-criticality. Some new experimental and analytical efforts have been started after the Fukushima accident. The present paper describes the perspectives on phenomenology and simulation of severe accident in LWRs, with the emphasis of insights obtained in the review of Fukushima accident. (author)

Influence of boron reduction strategies on PWR accident management flexibility

International Nuclear Information System (INIS)

Papukchiev, Angel Aleksandrov; Liu, Yubo; Schaefer, Anselm

2007-01-01

In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR. (author)

Consideration of severe accident issues for the general electric BWR standard plant a status report

International Nuclear Information System (INIS)

Holtzclaw, K.W.

1983-01-01

In early 1982 the U.S. NRC proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. This paper, GE's submission, discusses the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at TMI, is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

Hepatitis B virus core protein allosteric modulators can distort and disrupt intact capsids

Science.gov (United States)

Schlicksup, Christopher John; Wang, Joseph Che-Yen; Francis, Samson; Venkatakrishnan, Balasubramanian; Turner, William W; VanNieuwenhze, Michael

2018-01-01

Defining mechanisms of direct-acting antivirals facilitates drug development and our understanding of virus function. Heteroaryldihydropyrimidines (HAPs) inappropriately activate assembly of hepatitis B virus (HBV) core protein (Cp), suppressing formation of virions. We examined a fluorophore-labeled HAP, HAP-TAMRA. HAP-TAMRA induced Cp assembly and also bound pre-assembled capsids. Kinetic and spectroscopic studies imply that HAP-binding sites are usually not available but are bound cooperatively. Using cryo-EM, we observed that HAP-TAMRA asymmetrically deformed capsids, creating a heterogeneous array of sharp angles, flat regions, and outright breaks. To achieve high resolution reconstruction (particle symmetry. We deduced that HAP-TAMRA caused quasi-sixfold vertices to become flatter and fivefold more angular. This transition led to asymmetric faceting. That a disordered crosslink could rescue symmetry implies that capsids have tensegrity properties. Capsid distortion and disruption is a new mechanism by which molecules like the HAPs can block HBV infection. PMID:29377794

Revisiting Ulchin 4 SGTR Accident - Analysis for EOP Improvement

Energy Technology Data Exchange (ETDEWEB)

Lee, Eun-Hye; Lee, Wook-Jo; Jerng, Dong-Wook [Chung-Ang University, Seoul (Korea, Republic of)

2016-10-15

The Steam Generator Tube Ruputure (SGTR) is an accident that U-tube inside the SG is defected so that the reactor coolant releases through broken U-tube and this is one of design basis accidents. Operating the Nuclear Power Plants (NPP), maintaing the integrity of core and preventing radiation release are most important things. Because of risks, many researchers have studied scenarios, impacts and the ways to mitigate SGTR accidents. The study to provide an experimental database of aerosol particle retention and to develop models to support accident management interventions during SGTR was performed. The scaled-down models of NPP were used for experiments, also, MELCOR and SCDAP/RELAP5 were used to simulate a design basis SGTR accident. This study had a major role to resolve uncertainties of various physical models for aerosol mechanical resuspension. The other study which analyzed SGTR accident for System-integrated Modular Advanced Reactor (SMART) was performed. In this analysis, the amount of break flow was focused and TASS/SMRS code was used. It assumed that maximum leak was generated, and found that high RCS pressure, low core inlet coolant temperature, and low break location of the SG cassette contributed to leakage. Although the leakage was large, there was no direct release to atmosphere because the pressure of secondary loop was maintained below the safety relief valve set point. In this analysis, comparison of mitigating procedure when SGTR occurs between shutdown condition and full power condition was performed. In shutdown condition, the core uncovery would not take place in 16 hours whether the cooling procedures are performed or not. Therefore, the integrated amount of break flow should be considered only. In this point of view, cooling through intact SG only, case 3, is the best way to minimize the amount of break flow. In full power condition, the core water level is changed due to high reactor power. The important thing to protect NPP is to keep

APRI-6. Accident Phenomena of Risk Importance

International Nuclear Information System (INIS)

Garis, Ninos; Ljung, J

2009-06-01

Since the early 1980s, nuclear power utilities in Sweden and the Swedish Radiation Safety Authority (SSM) collaborate on the research in severe reactor accidents. In the beginning focus was mostly on strengthening protection against environmental impacts after a severe reactor accident, for example by develop systems for the filtered relief of the reactor containment. Since the early 90s, this focus has shifted to the phenomenological issues of risk-dominant significance. During the years 2006-2008, the partnership continued in the research project APRI-6. The aim was to show whether the solutions adopted in the Swedish strategy for incident management provides adequate protection for the environment. This is done by studying important phenomena in the core melt estimating the amount of radioactivity that can be released to the atmosphere in a severe accident. To achieve these objectives the research has included monitoring of international research on severe accidents and evaluation of results and continued support for research of severe accidents at the Royal Inst. of Technology (KTH) and Chalmers University. The follow-up of international research has promoted the exchange of knowledge and experience and has given access to a wealth of information on various phenomena relevant to events in severe accidents. The continued support to KTH has provided increased knowledge about the possibility of cooling the molten core in the reactor tank and the processes associated with coolability in the confinement and about steam explosions. Support for Chalmers has increased knowledge of the accident chemistry, mainly the behavior of iodine and ruthenium in the containment after an accident

APRI-6. Accident Phenomena of Risk Importance

Energy Technology Data Exchange (ETDEWEB)

Garis, Ninos; Ljung, J [eds.; Swedish Radiation Safety Authority, Stockholm (Sweden); Agrenius, Lennart [ed.; Agrenius Ingenjoersbyraa AB, Stockholm (Sweden)

2009-06-15

Since the early 1980s, nuclear power utilities in Sweden and the Swedish Radiation Safety Authority (SSM) collaborate on the research in severe reactor accidents. In the beginning focus was mostly on strengthening protection against environmental impacts after a severe reactor accident, for example by develop systems for the filtered relief of the reactor containment. Since the early 90s, this focus has shifted to the phenomenological issues of risk-dominant significance. During the years 2006-2008, the partnership continued in the research project APRI-6. The aim was to show whether the solutions adopted in the Swedish strategy for incident management provides adequate protection for the environment. This is done by studying important phenomena in the core melt estimating the amount of radioactivity that can be released to the atmosphere in a severe accident. To achieve these objectives the research has included monitoring of international research on severe accidents and evaluation of results and continued support for research of severe accidents at the Royal Inst. of Technology (KTH) and Chalmers University. The follow-up of international research has promoted the exchange of knowledge and experience and has given access to a wealth of information on various phenomena relevant to events in severe accidents. The continued support to KTH has provided increased knowledge about the possibility of cooling the molten core in the reactor tank and the processes associated with coolability in the confinement and about steam explosions. Support for Chalmers has increased knowledge of the accident chemistry, mainly the behavior of iodine and ruthenium in the containment after an accident.

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

Analysis of some accident conditions in confirmation of the HTGR safety

Energy Technology Data Exchange (ETDEWEB)

Grebennik, V. N.; Grishanin, E. I.; Kukharkin, N. E.; Mikhailov, P. V.; Pinchuk, V. V.; Ponomarev-Stepnoy, N. N.; Fedin, G. I.; Shilov, V. N.; Yanushevich, I. V. [Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii

1981-01-15

This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved.

Analysis of some accident conditions in confirmation of the HTGR safety

International Nuclear Information System (INIS)

Grebennik, V.N.; Grishanin, E.I.; Kukharkin, N.E.; Mikhailov, P.V.; Pinchuk, V.V.; Ponomarev-Stepnoy, N.N.; Fedin, G.I.; Shilov, V.N.; Yanushevich, I.V.

1981-01-01

This report concerns some accident conditions for the HTGR-50 demonstrational reactor which along with the safety features common to the typical HTGR differs in design. The analyses carried out on the accident situations showed that due to the high heat capacity of the graphite core and negative temperature effect of the reactivity the HTGR-50 reactor is effectively selfcontrolled at different perturbations of the reactivity and has low sensitivity to the failure of the core cooling. The primary circuit depressurization accident should be thoroughly studied because of the dangerous consequences i.e. the core overheating and the reactivity release into the environment. As a whole, the studies now in progress show that the problem of the HTGR safety can be successfully solved

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

Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report; Fortschrittliche Rechenmethoden zum Kernverhalten bei Reaktivitaetsstoerfaellen. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Pautz, A.; Perin, Y.; Pasichnyk, I.; Velkov, K.; Zwermann, W.; Seubert, A.; Klein, M.; Gallner, L.; Krzycacz-Hausmann, B.

2012-05-15

The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

Emergency core cooling system

International Nuclear Information System (INIS)

Ando, Masaki.

1987-01-01

Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

ON THE SURVIVABILITY AND METAMORPHISM OF TIDALLY DISRUPTED GIANT PLANETS: THE ROLE OF DENSE CORES

Energy Technology Data Exchange (ETDEWEB)

Liu, Shang-Fei; Lin, Douglas N. C. [Kavli Institute for Astronomy and Astrophysics and Department of Astronomy, Peking University, Beijing 100871 (China); Guillochon, James; Ramirez-Ruiz, Enrico, E-mail: liushangfei@pku.edu.cn [Department of Astronomy and Astrophysics, University of California, Santa Cruz, CA 95064 (United States)

2013-01-01

A large population of planetary candidates in short-period orbits have been found recently through transit searches, mostly with the Kepler mission. Radial velocity surveys have also revealed several Jupiter-mass planets with highly eccentric orbits. Measurements of the Rossiter-McLaughlin effect indicate that the orbital angular momentum vector of some planets is inclined relative to the spin axis of their host stars. This diversity could be induced by post-formation dynamical processes such as planet-planet scattering, the Kozai effect, or secular chaos which brings planets to the vicinity of their host stars. In this work, we propose a novel mechanism to form close-in super-Earths and Neptune-like planets through the tidal disruption of gas giant planets as a consequence of these dynamical processes. We model the core-envelope structure of gas giant planets with composite polytropes which characterize the distinct chemical composition of the core and envelope. Using three-dimensional hydrodynamical simulations of close encounters between Jupiter-like planets and their host stars, we find that the presence of a core with a mass more than 10 times that of the Earth can significantly increase the fraction of envelope which remains bound to it. After the encounter, planets with cores are more likely to be retained by their host stars in contrast with previous studies which suggested that coreless planets are often ejected. As a substantial fraction of their gaseous envelopes is preferentially lost while the dense incompressible cores retain most of their original mass, the resulting metallicity of the surviving planets is increased. Our results suggest that some gas giant planets can be effectively transformed into either super-Earths or Neptune-like planets after multiple close stellar passages. Finally, we analyze the orbits and structure of known planets and Kepler candidates and find that our model is capable of producing some of the shortest-period objects.

The CIEMAT’s forensic analyses of Fukushima accident: Contribution to the BSAF project

Energy Technology Data Exchange (ETDEWEB)

Herranz, L.E.; López, C.; Fontanet, J.; Fernández, E.

2015-07-01

The Fukushima accident is being both a unique opportunity and a huge challenge for severe accident analysis. Through the simulation of the accidents in Units 1 through 3 with MELCOR 2.1, three scenarios have been postulated which outcomes look consistent with data. These analyses indicate that a massive core damage should have happened in Unit 1, with most core molten and located in the containment, whereas Units 2 and 3 core damage is anticipated to be much less; however, there might be differences among these “twin” units. Anyway, in all the units the amount of H2 produced is over 500 kg. This work has been carried out in the frame of the international project for the understanding of the severe accidents occurred at Fukushima, the OECD-BSAF project. (Author)

Computational modeling for hexcan failure under core distruptive accidental conditions

Energy Technology Data Exchange (ETDEWEB)

Sawada, T.; Ninokata, H.; Shimizu, A. [Tokyo Institute of Technology (Japan)

1995-09-01

This paper describes the development of computational modeling for hexcan wall failures under core disruptive accident conditions of fast breeder reactors. A series of out-of-pile experiments named SIMBATH has been analyzed by using the SIMMER-II code. The SIMBATH experiments were performed at KfK in Germany. The experiments used a thermite mixture to simulate fuel. The test geometry of SIMBATH ranged from single pin to 37-pin bundles. In this study, phenomena of hexcan wall failure found in a SIMBATH test were analyzed by SIMMER-II. Although the original model of SIMMER-II did not calculate any hexcan failure, several simple modifications made it possible to reproduce the hexcan wall melt-through observed in the experiment. In this paper the modifications and their significance are discussed for further modeling improvements.

SEVERE ACCIDENT MANAGEMENT STATUS AT Loviisa

International Nuclear Information System (INIS)

Kymalainen, O.; Tuomisto, H.

1997-01-01

Some of the specific design features of IVO's Loviisa Plant, most notably the ice-condenser containment, strongly affect the plant response in a hypothetical core melt accident. They have together with the relatively stringent Finnish regulatory requirements forced IVO to develop a tailor made severe accident management strategy for Loviisa. The low design pressure of the ice-condenser containment complicates the design of the hydrogen management system. On the other hand, the ice-condensers and the water available from them are facilitating factors regarding in-vessel retention of corium by external cooling of reactor pressure vessel. This paper summarizes the Finnish severe accident requirements, IVO's approach to severe accidents, and its application to the Loviisa Plant

Theoretical analysis and numerical modelling of heat transfer and fuel migration in underlying soils and constructive elements of nuclear plants during an accident release from the core

International Nuclear Information System (INIS)

Arutunjan, R.V.; Bolshov, L.A.; Vitukov, V.V.; Goloviznin, V.M.; Dykhne, A.M.; Kiselev, V.P.; Klementova, S.V.; Krayushkin, I.E.; Moskovchenko, A.V.; Pismennii, V.D.; Popkov, A.G.; Chernov, S.Y.; Chudanov, V.V.; Khoruzhii, O.V.; Yudin, A.I.

1990-01-01

Migration of fuel fragments and core fission products during severe accidents on nuclear plants is studied analytically and numerically. The problems of heat transfer and migration of volume heat sources in construction materials and underlying soils are considered

131I release from a HTGR during the LOFC accident

International Nuclear Information System (INIS)

Foley, J.E.

1975-03-01

The time-dependent release of 131 I from both the core and the containment building of a high temperature gas-cooled (HTGR) reactor during the loss of forced coolant (LOFC) accident is studied. A simplified core release model is combined with a containment building release model so that the total amount of the isotope released to the environment can be calculated. The time-dependent release of 131 I from the core during the LOFC accident is primarily a function of the time-dependent core temperatures and the failed fuel release constants. The most important factor in calculating the amount of the isotope released to the environment is the total amount released into the containment building. (U.S.)

Analysis of Fukushima unit 2 accident considering the operating conditions of RCIC system

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Il, E-mail: sikim@kaeri.re.kr; Park, Jong Hwa; Ha, Kwang Soon; Cho, Song-Won; Song, JinHo

2016-03-15

Highlights: • Fukushima unit 2 accident was analyzed using MELCOR 1.8.6. • RCIC operating conditions were assumed and best case was selected. • Effect of RCIC operating condition on accident scenario was found. - Abstract: A severe accident in Fukushima occurred on March 11, 2011 and units 1, 2 and 3 were damaged severely. A tsunami following an earthquake made the supply of electricity power stop, and the safety systems, which use AC or DC power in plants could not operate properly. It is supposed that the degree of core degradation of unit 2 is less serious than in the other plants, and it was estimated that the operation of reactor core isolation cooling (RCIC) system at the initial stage of the accident minimized the core damage through decay heat removal. Although the operating conditions of the RCIC system are not known clearly, it can be important to analyze the accident scenario of unit 2. In this study, best case of the Fukushima unit 2 accident was presented considering the operating conditions of the RCIC system. The effects of operating condition on core degradation and fission product release rate to environment were also examined. In addition, importance of torus room flooding level in the accident analysis was discussed. MELCOR 1.8.6 was used in this research, and the geometries of plant and operating conditions of safety system were obtained from TEPCO through OECD/NEA BSAF Project.

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

Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

Directory of Open Access Journals (Sweden)

Gilberto Espinosa-Paredes

2012-01-01

Full Text Available The loss-of-coolant accident (LOCA simulation in the boiling water reactor (BWR of Laguna Verde Nuclear Power Plant (LVNPP at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

Severe accident management: a summary of the VAHTI and ROIMA projects

International Nuclear Information System (INIS)

Sairanen, R.

1998-01-01

Two severe accident research projects: 'Severe Accident Management' (VAHTI), 1994-96 and 'Reactor Accidents' Phenomena and Simulation (ROIMA) 1997-98. have been conducted at VTT Energy within the RETU research programme. The main objective was to assist the severe accident management programmes of the Finnish nuclear power plants. The projects had several subtopics. These included thermal hydraulic validation of the APROS code, studies of failure mode of the BWR pressure vessel, investigation of core melt progression within a BWR pressure vessel, containment phenomena, development of a computerised severe accident training tool, and aerosol behaviour experiments. The last topic is summarised by another paper in the seminar. The projects have met the objectives set at the project commencement. Calculation tools have been developed and validated suitable for analyses of questions specific for the Finnish plants. Experimental fission product data have been produced that can be used to validate containment aerosol codes. The tools and results have been utilised in plant assessments. One of the main achievements has been the computer code PASULA for analysis of interactions between core melt and pressure vessel. The code has been applied to pressure vessel penetration analysis. The results have shown the importance of the nozzle construction. Modelling possibilities have recently improved by addition of a creep and porous debris models. Cooling of a degraded BWR core has been systematically studied as joint Nordic projects with a set of severe accident codes. Estimates for coolable conditions have been provided. Recriticality due to reflooding of a damaged core has been evaluated. (orig.)

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

Energy Technology Data Exchange (ETDEWEB)

Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

2017-06-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

International Nuclear Information System (INIS)

Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

2017-01-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Consideration of severe accident issues for the General Electric BWR standard plant: Chapter 10

International Nuclear Information System (INIS)

Holtzclaw, K.W.

1983-01-01

In early 1982, the U.S. Nuclear Regulatory Commission (NRC) proposed a policy to address severe accident rulemaking on future plants by utilizing standard plant licensing documentation. GE provided appendices to the licensing documentation of its standard plant design, GESSAR II, which address severe accidents for the GE BWR/6 Mark III 238 nuclear island design. The GE submittals discuss the features of the design that prevent severe accidents from leading to core damage or that mitigate the effects of severe accidents should core damage occur. The quantification of the accident prevention and mitigation features, including those incorporated in the design since the accident at Three Mile Island (TMI), is provided by means of a comprehensive probabilistic risk assessment, which provides an analysis of the probability and consequences of postulated severe accidents

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Tchernobyl accident

International Nuclear Information System (INIS)

1986-06-01

First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given [fr

Learning from nuclear accident experience

International Nuclear Information System (INIS)

Vaurio, J.K.

1984-01-01

Statistical procedures are developed to estimate accident occurrence rates from historical event records, to predict future rates and trends, and to estimate the accuracy of the rate estimates and predictions. Maximum likelihood estimation is applied to several learning models, and results are compared to earlier graphical and analytical estimates. The models are based on (1) the cumulative number of operating years, (2) the cumulative number of plants built, and (3) accidents (explicitly), with the accident rate distinctly different before and after an accident. The statistical accuracies of the parameters estimated are obtained in analytical form using the Fisher information matrix. Using data on core damage accidents in electricity producing plants, it is estimated that the probability for a plant to have a serious flaw has decreased from 0.1 to 0.01 during the developmental phase of the nuclear industry. At the same time the equivalent frequency of accidents has decreased from 0.04 per reactor year to 0.0004 per reactor year, partly due to the increasing population of plants. 10 references, 7 figures, 2 tables

Accident at Harrisburg

International Nuclear Information System (INIS)

1979-05-01

The course of events during the accident on 28 March 1979 at Three Mile Island-2 Reactor at Harrisburg, Pennsylvania, is described in detail. The effects (in the environment and within the safety containment) are described. The following points are then discussed: the possibility of a comparable accident occurring in the nuclear power stations in the German Federal Republic; the possibility of any point having been overlooked in the design of nuclear power stations in the Federal Republic; whether previous risk analyses are still valid; and how near the Three Mile Island reactor was to a core meltdown. Some conclusions are drawn. (U.K.)

Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

International Nuclear Information System (INIS)

Zhang Yanzhao; Zhang Fan; Zhao Xinwen; Zheng Yingfeng

2013-01-01

Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

Accident sequence quantification with KIRAP

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP`s cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs.

Accident sequence quantification with KIRAP

International Nuclear Information System (INIS)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP's cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs

Post-accident heat removal research: A state of the art review

International Nuclear Information System (INIS)

Mueller, U.; Schulenberg, T.

1983-11-01

For a realistic assessment of the consequence of extremely unlikely reactor accidents resulting in core degradation or core meltdown key questions are how to remove the decay heat from the reactor system and how to retain the radioactive core debris within the containment. Usually, this complex of questions is referred to as Post-Accident Heat Removal (PAHR). In this article the research work on PAHR performed by various institutions during the last decade has been reviewed. The main results have been summarized under the chapter headings ''Accident Scenarios,'' - ''Core Debris Accommodation Concepts,'' and ''PAHR Topics.'' Particular emphasis has been placed on the presentation of the following problems: characteristics and coolability of solid core debris in the vector vessel, heat removal from molten pools of core material, and core-melt interaction with structural materials. Some unresolved or insufficiently answered questions relating to special ''PAHR Topics'' have been mentioned or discussed at the end of the particular Chapter. Problem areas of major uncertainty have been identified and listed at the end of the review article. They include the following subjects: formation of debris beds and bed characteristics, post dryout behaviour of particle beds, long-term availability and proper location of heat sinks, creep rupture of structures under high thermal loads. (orig.) [de

Thermal-hydraulic studies on molten core-concrete interactions

International Nuclear Information System (INIS)

Greene, G.A.

1986-10-01

This report discusses studies carried out in connection with light water power reactor accidents. Recent assessments have indicated that the consequences of molten-core concrete interactions dominate the considerations of severe accidents. The two areas of interest that have been investigated are interlayer heat and mass transfer and liquid-liquid boiling. Interlayer heat and mass transfer refers to processes that occur within a core melt between the stratified, immiscible phases of core oxides and metals. Liquid-liquid boiling refers to processes that occur at the melt-concrete on melt-coolant interface

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

International Nuclear Information System (INIS)

Longworth, J.P.; Tobias, A.

1986-07-01

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

Chemical phenomena under severe accident conditions

International Nuclear Information System (INIS)

Powers, D.A.

1988-01-01

A severe nuclear reactor accident is expected to involve a vast number of chemical processes. The chemical processes of major safety significance begin with the production of hydrogen during steam oxidation of fuel cladding. Physico-chemical changes in the fuel and the vaporization of radionuclides during reactor accidents have captured much of the attention of the safety community in recent years. Protracted chemical interactions of core debris with structural concrete mark the conclusion of dynamic events in a severe accident. An overview of the current understanding of chemical processes in severe reactor accident is provided in this paper. It is shown that most of this understanding has come from application of findings from other fields though a few areas have in the past been subject to in-depth study of a fundamental nature. Challenges in the study of severe accident chemistry are delineated

Outline of Fukushima nuclear accident and future action. Lessons learned from accident and countermeasure plan

International Nuclear Information System (INIS)

Fukuda, Toshihiko

2012-01-01

Fukushima nuclear accident was caused by loss of all AC power sources (SBO) and loss of ultimate heat sink (LUHS) at Fukushima Daiichi Nuclear Power Plants (NPPs) hit by the Great East Japan Earthquake. This article reviewed outline of Fukushima nuclear accident progression when on year had passed since and referred to lessons learned from accident and countermeasure plan to prevent severe accident in SBO and LUHS events by earthquake and tsunami as future action. This countermeasure would be taken to (1) prevent serious flooding in case a tsunami overwhelms the breakwater, with improving water tightness of rooms for emergency diesel generator, batteries and power centers, (2) enhance emergency power supply and cooling function with mobile electricity generator, high pressure fire pump car and alternate water supply source, (3) mitigate environmental effects caused by core damage with installing containment filtered venting, and (4) enforce emergency preparedness in case of severe accident. Definite countermeasure plan for Kashiwazaki-Kariwa NPPs was enumerated. (T. Tanaka)

Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

2017-03-15

Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

A thermohydraulic analysis for LOCA accident of a CANDU 600 reactor core charged with SEU 43 fuel by means of FIREBIRD code

International Nuclear Information System (INIS)

Serbanel, M.; Catana, A.

2001-01-01

This report presents a comparative analysis of the behaviour of primary circuit during a LOCA 20% RIH accident for two types of reactor core, namely, normally charged, i.e., with clusters of 37 rods and charged with clusters of 43 rods, respectively. This type of accident was chosen since Canadian analyses showed that the associated transient regime stress the fuel elements. The void reactivity as a function of coolant average density was calibrated for a reference regime (LOCA 20% RIH) so that the results of the model be able to reproduce the average distribution in the reference transient regime. The computation makes use of CERBERUS and FIREBIRD codes externally coupled by files. The void reactivity of the hot pencil was obtained this way. An extremely conservative hypothesis was used, namely that the momentary power of the cluster hosting the pencil is the maximal power over the cluster for the corresponding half reactor core. To carry out this work the following steps were covered: 1. The scenario for the LOCA 20% RIH accident was worked out and the input data corresponding to the thermohydraulic and neutronic modules, for the complex model and the 37 rod clusters, were checked; 2. The input data corresponding to the thermohydraulic module for the complex model and the 43 rod cluster were checked; 3. The kinetic parameters corresponding to the 37 rod cluster were computed; 4. The kinetic parameters corresponding to the 43 rod cluster were computed and the file for the input data in the neutronic module was built; 5. A sub-routine for writing files with the thermohydraulic and neutronic quantities, in a format adequate to the other programs, was implemented; 6. The two transient regimes considered were implemented and the archives containing the quantities were built ;7. The results obtained were analyzed. The conclusion of this work is that in case of LOCA 20% RIH accident the 43 bar clusters have a better behaviour than the 37 bar clusters

ALWR severe accident issue resolution in support of updated emergency planning

International Nuclear Information System (INIS)

Additon, Stephen L.; Leaver, David E.; Sorrell, Steven W.; Theofanous, Theo G.

2004-01-01

The Advanced Light Water Reactor (ALWR) Program in the U.S. is a cooperative, cost-sharing undertaking between the U.S. government, industry, and a number of international participants, with the objective of developing the next generation of nuclear power plants. The ALWR designs emphasize improvements in safety and operational reliability through simplification, improved safety margins, innovative passive safety systems, enhanced man-machine interfaces, and incorporation of the lessons learned from the operation of existing LWR plants. An important component of the improved safety characteristics of ALWRs is the consideration of severe accidents in the plant design. The U.S. Department of Energy (DOE) initiated the Advanced Reactor Severe Accident Program (ARSAP) to assist in the transfer of severe accident technology from the U.S. national laboratories to the industry to implement this approach. The basic design requirements for this new generation of nuclear power plants were developed, under the management of the Electric Power Research Institute (EPRI) by the utilities and documented in the Utility Requirements Document (URD). The URD safety policy is based on the traditional 'defense-in-depth' approach, which emphasizes prevention through safety systems which prevent accidents from progressing to core damage, and mitigation to ensure that accidents are mitigated and contained. In a major departure from previous practice, severe accidents, including postulated core melt events, are specifically included in the defense-in-depth design considerations for ALWRs. As a result of this approach, the emergency planning assumptions and criteria warrant a review and reevaluation for ALWR designs. ALWRs present a risk profile that is significantly different than that which served as the basis for the emergency planning requirements for operating plants. The determination of this profile necessarily requires the characterization of the severe accident response of ALWRs

Experimental study on air ingress during a primary pipe rupture accident with a graphite reactor core simulator

International Nuclear Information System (INIS)

Takeda, Tetsuaki; Hishida, Makoto; Baba, Shinichi

1991-11-01

When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary cooling system reduces the pressure. After the pressures are balanced between the reactor and the container, air is expected to enter into the reactor core from the breach. It seems to be probable that the graphite structures is oxidized by air. Hence, it is necessary to investigate the air ingress process and the behavior of the generating gases by the oxidation reactions. The previous experimental study is performed on the molecular diffusion and natural convection of the two component gas mixtures using a test model simulating simply the reactor. Objective of the study was to investigate the air ingress process during the early stage of the primary pipe rupture accident. However, since the model did not have any kind of graphite components, the reaction between graphite and oxygen was not simulated. The present model includes the reactor core and the high temperature plenum simulators made of graphite. The major results obtained in the present study are summarized in the followings: (1) The air ingress process with graphite oxidation reaction is similar to that without the reaction qualitatively. (2) When the reactor core simulator is maintained at low temperatures (lower than 450degC), the initiation time of the natural circulation of air is almost equal to that of the natural circulation of nitrogen. On the other hand, when the temperature of the reactor core simulator is high (more than 500degC), the initiation time of the natural circulation of air is earlier than that of nitrogen. (3) When the temperature of the reactor core simulator is higher than 600degC, oxygen is almost dissipated by the graphite structures. When the temperature of the reactor core simulator is below 700degC, carbon dioxide mainly is generated by the oxidation reactions. (author)

Small break LOCA [loss of coolant accident] mitigation for Bellefonte

International Nuclear Information System (INIS)

Bayless, P.D.; Dobbe, C.A.

1986-01-01

Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

Energy Technology Data Exchange (ETDEWEB)

Kirkland, Karen Vierow [Texas A & M Univ., College Station, TX (United States); Ross, Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Beeny, Bradley [Texas A & M Univ., College Station, TX (United States); Luthman, Nicholas [Texas A& M Engineering Experiment Station, College Station, TX (United States); Strater, Zachary [Texas A & M Univ., College Station, TX (United States)

2017-12-23

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that the system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.

Development of a parametric containment event tree model for a severe BWR accident

Energy Technology Data Exchange (ETDEWEB)

Okkonen, T [OTO-Consulting Ay, Helsinki (Finland)

1995-04-01

A containment event tree (CET) is built for analysis of severe accidents at the TVO boiling water reactor (BWR) units. Parametric models of severe accident progression and fission product behaviour are developed and integrated in order to construct a compact and self-contained Level 2 PSA model. The model can be easily updated to correspond to new research results. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting, and are focused mainly on the use and effects of the dedicated severe accident management (SAM) systems. Severe accident progression from eight plant damage states (PDS), involving different pre-core-damage accident evolution, is examined, but the inclusion of their relative or absolute probabilities, by integration with Level 1, is deferred to integral safety assessments. (33 refs., 5 figs., 7 tabs.).

Perspectives on the economic risks of LWR accidents

International Nuclear Information System (INIS)

Ritchie, L.T.; Burke, R.P.

1986-01-01

Models which can be used for the analysis of the economic risks from events which may occur during LWR operation have been developed. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The economic consequence models have been applied in studies of the economic risks from the operation of US LWR plants. The results of the analyses provide some important perspectives regarding the economic risks of LWR accidents. The analyses indicate that economic risks, in contrast to public health risks, are dominated by the onsite costs of relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for a typical US plant

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal fires during mid-loop operations. Volume 3, Part 1, Main report

International Nuclear Information System (INIS)

Musicki, Z.; Chu, T.L.; Yang, J.; Ho, V.; Hou, Y.M.; Lin, J.; Siu, N.

1994-07-01

During l989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than fun power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in ' the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few. procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful

MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

International Nuclear Information System (INIS)

Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

2011-01-01

In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

SIMMER as a safety analysis tool

International Nuclear Information System (INIS)

Smith, L.L.; Bell, C.R.; Bohl, W.R.; Bott, T.F.; Dearing, J.F.; Luck, L.B.

1982-01-01

SIMMER has been used for numerous applications in fast reactor safety, encompassing both accident and experiment analysis. Recent analyses of transition-phase behavior in potential core disruptive accidents have integrated SIMMER testing with the accident analysis. Results of both the accident analysis and the verification effort are presented as a comprehensive safety analysis program

Transport of nuclides during a core meltdown accident, with consideration of filtered venting

International Nuclear Information System (INIS)

Haeggblom, H.

1981-01-01

A BWR core meltdown accident has been studied with respect to the transport of radioactive and nonactive gases and aerosols. A system consisting of a containment with an outer stone condenser in three parts was considered. Calculations of the aerosol behaviour have been made with the computer programme NAUA and HAARM-3, assuming one single compartment. Results from these calculations have been used for multicompartment calculations with CORRAL II. The code was modified so that particles of different sizes could be considered in the different compartments, and the time dependence of the particles can be arbitrary. In addition to the aerosol transport and deposition, the corresponding quantities for elemental iodine were calculated. It was concluded, that if the total volume of the condenser system is of the order of 10 5 m 3 , practically all elemental iodine and particles can be retained in the system. The only leakage to the environment will be caused by inefficient sealing during the first five hours. The pressure can never damage the condenser. (author)

A severe accident analysis for the system-integrated modular advanced reactor

International Nuclear Information System (INIS)

Jung, Gunhyo; Jae, Moosung

2015-01-01

The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

CERN Document Server

Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

1982-01-01

The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

Review of the SCDAP/RELAP5/MOD3.1 code structure and core T/H model before core damage

International Nuclear Information System (INIS)

Kim, See Darl; Kim, Dong Ha

1998-04-01

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code is being developed at the INEL under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. NRC. As The current time, the SCDAP/RELAP5/MOD3.1 code is the result of merging the RELAP5/MOD3 and SCDAP models. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. Major purpose of the report is to provide information about the characteristics of SCDAP/RELAP5/MOD3.1 core T/H models for an integrated severe accident computer code being developed under the mid/long-term project. This report analyzes the overall code structure which consists of the input processor, transient controller, and plot file handler. The basic governing equations to simulate the thermohydraulics of the primary system are also described. As the focus is currently concentrated in the core, core nodalization parameters of the intact geometry and the phenomenological subroutines for the damaged core are summarized for the future usage. In addition, the numerical approach for the heat conduction model is investigated along with heat convection model. These studies could provide a foundation for input preparation and model improvement. (author). 6 refs., 3 tabs., 4 figs

Annual meeting on nuclear technology 1982. Technical meeting: Possibilities and effects of serious reactor accidents

International Nuclear Information System (INIS)

1982-01-01

A critical examination of the forecast of a design basis accident, the view of the Sandia National Laboratory on the probability of a steam explosion after a core meltdown accident is comparison with WASH-1400, the possibilities of interactions with the containment structure and fission product release, as well as the influences for the assessment of risk in Germany taken from the analysis of core meltdown accidents are dealt with in these papers. (DG) [de

The internal core catcher in Super Phenix 1

International Nuclear Information System (INIS)

Le Rigoleur, C.; Kayser, G.; Maurin, G.; Magnon, B.

1982-07-01

The internal core catcher in SUPER PHENIX 1 is described here in some detail. The fuel retention capabilities are presented for situations of increasing severity. The first situation corresponds to the core catcher design. It relates to a hypothetical subassembly accident that would cause a limited quantity of fuel, corresponding to the mass of seven subassemblies, to be deposited on the core catcher. For this situation and at all levels of the analysis, the most conservative assumptions are made in order to prove the integrity of the core catcher. The second situation corresponds to a hypothetical larger core melt accident. In this case, for some of the parameters, assumptions are made that correspond to the most likely situations based on engineering considerations. Then the maximum retention capabilities are presented

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-30

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

Development of Severe Accident Containment Analysis Package

Energy Technology Data Exchange (ETDEWEB)

Park, Chang-Hwan; Kim, Dong-Min; Seo, Jea-Uk; Lee, Dea-Young; Park, Soon-Ho; Lee, Jae-Gwon; Lee, Jin-Yong; Lee, Byung-Chul [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

In safety viewpoint, the pressure and temperature of the containment is the important parameters, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In addition, there are possibilities of occurrence of other relevant phenomena following the reactor core melting such as DCH(direct containment heating) due to HPME(high pressure melt ejection), steam explosion due to fuel-coolant interaction in the reactor cavity and molten core concrete interaction at the late stage. It is important to predict the containment responses during a severe accident by a reasonable accuracy for establishing of effective mitigation strategies and preparation of the safety features required. In this paper, the overview of the SACAP development status is presented. SACAP is developed so as to be able to analyze, so called, Ex-Vessel severe accident phenomena including thermal-hydraulics, combustible gas burn, direct containment heating, steam explosion and molten core-concrete interaction. At the parallel time, SACAP and In-Vessel analysis module named CSPACE are processed for integration through MPI communication coupling. Development of the integrated severe accident analysis code system will be completed in following one year to make the code revision zero to be released.

Severe accident mitigation through containment design

International Nuclear Information System (INIS)

Bergeron, K.D.

1990-01-01

Recent U.S. Department of Energy plans to construct a Heavy Water Reactor for the production of defense nuclear materials have created a unique opportunity to explore ways to mitigate severe accident concerns in the design stage. Drawing on an extensive background in US-NRC-sponsored severe accident work, Sandia National Laboratories has been exploring a number of Heavy Water New Production Reactor (HW-NPR) containment design strategies that might mitigate the consequences of a core-melt accident without greatly impacting construction cost or reactor operations. Severe accident specialists have undertaken these assessments with the intent of providing the plant designers with some of the phenomenological advantages and disadvantages of various mitigation strategies. This paper will highlight some of the more interesting concepts and summarize the results obtained. (author). 9 refs., 2 tabs

Severe accident mitigation through containment design

International Nuclear Information System (INIS)

Bergeron, K.D.

1990-01-01

Recent US Department of Energy plans to construct a Heavy Water Reactor for the production of defense nuclear materials have created a unique opportunity to explore ways to mitigate severe accident concerns in the design stage. Drawing on an extensive background in USNRC-sponsored severe accident work, Sandia National Laboratories has been exploring a number of Heavy Water New Production Reactor (HW-NPR) containment design strategies that might mitigate the consequences of a core-melt accident without greatly impacting construction cost or reactor operations. Severe accident specialists have undertaken these assessments with the intent of providing the plant designers with some of the phenomenological advantages and disadvantages of various mitigation strategies. This paper will highlight some of the more interesting concepts and summarize the results obtained. 9 refs., 2 tabs

Accident sequence precursor analysis level 2/3 model development

International Nuclear Information System (INIS)

Lui, C.H.; Galyean, W.J.; Brownson, D.A.

1997-01-01

The US Nuclear Regulatory Commission's Accident Sequence Precursor (ASP) program currently uses simple Level 1 models to assess the conditional core damage probability for operational events occurring in commercial nuclear power plants (NPP). Since not all accident sequences leading to core damage will result in the same radiological consequences, it is necessary to develop simple Level 2/3 models that can be used to analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude of the resulting radioactive releases to the environment, and calculate the consequences associated with these releases. The simple Level 2/3 model development work was initiated in 1995, and several prototype models have been completed. Once developed, these simple Level 2/3 models are linked to the simple Level 1 models to provide risk perspectives for operational events. This paper describes the methods implemented for the development of these simple Level 2/3 ASP models, and the linkage process to the existing Level 1 models

CFD investigating the air ingress accident for a HTGR simulation of graphite corrosion oxidation

International Nuclear Information System (INIS)

Ferng, Y.M.; Chi, C.W.

2012-01-01

Highlights: ► A CFD model is proposed to investigate graphite oxidation corrosion in the HTR-10. ► A postulated air ingress accident is assumed in this paper. ► Air ingress flowrate is the predicted result, instead of the preset one. ► O 2 would react with graphite on pebble surface, causing the graphite corrosion. ► No fuel exposure is predicted to be occurred under the air ingress accident. - Abstract: Through a compressible multi-component CFD model, this paper investigates the characteristics of graphite oxidation corrosion in the HTR-10 core under the postulated accident of gas duct rupture. In this accident, air in the steam generator cavity would enter into the core after pressure equilibrium is achieved between the core and the cavity, which is also called as the air ingress accident. Oxygen in the air would react with graphite on pebble surface, subsequently resulting in oxidation corrosion and challenging fuel integrity. In this paper, characteristics of graphite oxidation corrosion during the air ingress accident can be reasonably captured, including distributions of graphite corrosion amount on the different cross-sections, time histories of local corrosion amount at the monitoring points and overall corrosion amount in the core, respectively. Based on the transient simulation results, the corrosion pattern and its corrosion rate would approach to the steady-state conditions as the accident continuously progresses. The total amount of graphite corrosion during a 3-day accident time is predicted to be about 31 kg with the predicted asymptotic corrosion rate. This predicted value is less than that from the previous work of Gao and Shi.

Identification and evaluation of PWR in-vessel severe accident management strategies

International Nuclear Information System (INIS)

Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

1992-03-01

This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

A physical tool for severe accident mitigation studies

Energy Technology Data Exchange (ETDEWEB)

Marie, N., E-mail: nathalie.marie@cea.fr [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Bachrata, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France); Barjot, F. [EDF R& D, SINETICS, F-93141 Clamart (France); Marrel, A. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France); Gossé, S. [CEA, DEN, DPC, F-91191 Gif Sur Yvette (France); Bertrand, F. [CEA, DEN, DER, F-13108 Saint Paul Lez Durance (France)

2016-12-01

Highlights: • Physical tool for mitigation studies devoted to SFR safety. • Physical models to describe the material discharge from core. • Comparison to SIMMER III results. • Studies for ASTRID safety assessment and support to core design. - Abstract: Within the framework of the Generation IV Sodium-cooled Fast Reactors (SFR) R&D program of CEA, the core behavior in case of severe accidents is being assessed. Such transients are usually simulated with mechanistic codes (such as SIMMER-III). As a complement to this code, which gives reference accidental transient, a physico-statistical approach is currently followed; its final objective being to derive the variability of the main results of interest for the safety. This approach involves a fast-running simulation of extended accident sequences coupling low-dimensional physical models to advanced statistical analysis techniques. In this context, this paper presents such a low-dimensional physical tool (models and simulation results) dedicated to molten core materials discharge. This 0D tool handles heat transfers from molten (possibly boiling) pools, fuel crust evolution, phase separation/mixing of fuel/steel pools, radial thermal erosion of mitigation tubes, discharge of core materials and associated axial thermal erosion of mitigation tubes. All modules are coupled with a global neutronic evolution model of the degraded core. This physical tool is used to study and to define mitigation features (function of tubes devoted to mitigation inside the core, impact of absorbers falling into the degraded core…) to avoid energetic core recriticality during a secondary phase of a potential severe accident. In the future, this physical tool, associated to statistical treatments of the effect of uncertainties would enable sensitivity analysis studies. This physical tool is described before presenting its comparison against SIMMER-III code results, including a space-and energy-dependent neutron transport kinetic

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)

Proceedings of the specialist meeting on selected containment severe accident management strategies

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-15

Twenty papers were presented at the first specialist meeting on Selected Containment Severe Accident management Strategies, held in Stockholm, Sweden, in 1994, half of them dealing with accident management strategies implementation status, half of them with research aspects. The four sessions were: general aspects of containment accident management strategies, hydrogen management techniques, other containment accident management strategies (spray cooling, core catcher...), surveillance and protection of containment function

Proceedings of the specialist meeting on selected containment severe accident management strategies

International Nuclear Information System (INIS)

1995-07-01

Twenty papers were presented at the first specialist meeting on Selected Containment Severe Accident management Strategies, held in Stockholm, Sweden, in 1994, half of them dealing with accident management strategies implementation status, half of them with research aspects. The four sessions were: general aspects of containment accident management strategies, hydrogen management techniques, other containment accident management strategies (spray cooling, core catcher...), surveillance and protection of containment function

Re criticality assessment following reactor core damage in Fukushima unit 2

International Nuclear Information System (INIS)

Jeong, Hae Sun; Song, Jin Ho; Park, Chang Je; Ha, Kwang Soon; Song, Yong Mann; Ryu, Eun Hyun

2012-01-01

Following the severe core damage accident at the Fukushima nuclear power plants (NPPs), many researchers have studied a possible Re criticality caused by core melting or corium. However, no one can accurately examine the internal conditions of the reactor vessel, and thus there have been different opinions from some organizations depending on their assumption and analysis methods. If there is a potential Re criticality in the reactor vessel, some counter plans for the accident management should be established to prevent and mitigate re criticality, and to return the plant to a safe and stable state. In this study, the criticality level following a severe core damage accident was first analyzed using the MCNPX 2.6.0 code. Based on this result, practical strategies in terms of accident management were obtained by charging soluble boron (H 3B O 3) into re flooded water

Accident analysis of Fukushima Daiichi Nuclear Power Station unit 1

International Nuclear Information System (INIS)

Kobayashi, Masahide; Narabayashi, Tadashi; Tsuji, Masashi; Chiba, Go; Nagata, Yasunori; Shimoe, Tomohiro

2015-01-01

As a result of the Great East Japan Earthquake that occurred on 11 March 2011, all AC and DC power at the Fukushima Daiichi NPP units 1 to 3 were lost soon after the tsunami. The core cooling function was lost, and the cores of units 1 to 3 were damaged. The purpose of this work is to clarify the progress of the accident in unit 1, which was damaged the earliest among the 3 units. Therefore, an original severe accident analysis code was developed, and the progress of the accident was evaluated from the analysis results and the actual data. As a result, the leakage path from a pressure vessel was clarified, and some lessons and knowledge were gained. (author)

SCDAP/RELAP5 lower core plate model

International Nuclear Information System (INIS)

Coryell, E.W.; Griffin, F.P.

1999-01-01

The SCDAP/RELAP5 computer code is a best-estimate analysis tool for performing nuclear reactor severe accident simulations. This report describes the justification, theory, implementation, and testing of a new modeling capability which will refine the analysis of the movement of molten material from the core region to the vessel lower head. As molten material moves from the core region through the core support structures it may encounter conditions which will cause it to freeze in the region of the lower core plate, delaying its arrival to the vessel head. The timing of this arrival is significant to reactor safety, because during the time span for material relocation to the lower head, the core may be experiencing steam-limited oxidation. The time at which hot material arrives in a coolant-filled lower vessel head, thereby significantly increasing the steam flow rate through the core region, becomes significant to the progression and timing of a severe accident. This report is a revision of a report INEEL/EXT-00707, entitled ''Preliminary Design Report for SCDAP/RELAP5 Lower Core Plate Model''

LMFBR accident delineation study: approach and preliminary results

International Nuclear Information System (INIS)

Williams, D.C.; Sholtis, J.A.; Rios, M.; Worledge, D.H.; Conrad, P.W.; Varela, D.W.; Pickard, P.S.

1979-01-01

Event trees have been constructed for all phases of LMFBR accidents. The trees proved useful for identifying meaningful initiating accident categories and containment responses. In these areas, quantification appears feasible, given an adequate data base. Event trees were also used to represent in-core phenomenological questions governing accident progression and energetics, but here quantification appears impracticable because pervasive phenomenological uncertainties exist. Infrequent accident initiation is the dominant factor in assuring low risk. Nevertheless, containment promises an additional measure of risk reduction provided severe energetics are highly unlikely. The delineation served to systematize LMFBR safety issues and should aid in evaluating LMFBR R and D priorities

Fuel- and clad-motion diagnostics: licensing needs

International Nuclear Information System (INIS)

Bari, R.A.; Meyer, J.F.

1976-01-01

The paper addresses the current state of uncertainty with respect to fuel and clad motion during a hypothetical core-disruptive accident in a liquid metal fast breeder reactor as it relates to licensing needs. It should be noted that the paper does not represent an official position of the U.S. Nuclear Regulatory Commission, but rather, represents, in part, opinions and conclusions of its contractors. Particular attention is given to the needs for an assessment of the course of events during a hypothetical core-disruptive accident in the Clinch River Breeder Reactor. However, some of the issues discussed are likely to be relevant to larger breeder reactors as well. The issues addressed are related to the needs associated with analyses of the loss-of-flow (LOF) accident without scram and the transient overpower (TOP) accident, without scram

Nuclear Reactor RA Safety Report, Vol. 16, Maximum hypothetical accident

International Nuclear Information System (INIS)

1986-11-01

Fault tree analysis of the maximum hypothetical accident covers the basic elements: accident initiation, phase development phases - scheme of possible accident flow. Cause of the accident initiation is the break of primary cooling pipe, heavy water system. Loss of primary coolant causes loss of pressure in the primary circuit at the coolant input in the reactor vessel. This initiates safety protection system which should automatically shutdown the reactor. Separate chapters are devoted to: after-heat removal, coolant and moderator loss; accident effects on the reactor core, effects in the reactor building, and release of radioactive wastes [sr

Accidents of loss of flow for the ETTR-2 reactor; deterministic analysis

International Nuclear Information System (INIS)

El-Messiry, A.M.

2000-01-01

The main objective for reactor safety is to keep the fuel in a thermally safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accident (DBA) as loss of flow accident (LOFA), is required assessing reactor safety. The present work concerns this transients applied to Egypt Test and Research Reactor ETRR-3 (new reactor). An accident analysis code FLOWTR is developed to investigate the thermal behaviour of the core during such flow transients. The active core is simulated by two channels: 1 - hot channel (HC), and 2 - average channel (AC) representing the remainder of the core. Each channel is divided into four axial sections. The external loop, core plenums, and core chimney are simulated by different dynamic loops. The code includes modules for pump cast down, flow regimes, decay heat, temperature distributions, and feedback coefficients. FLOWTR is verified against results from RETRAN code, THERMIC code and commissioning tests for null transient case. The comparison shows a good agreement. The study indicates that for LOFA transients, provided the scram system is available, the core is shutdown safely by low flow signal (496.6 kg/s) at 1.4 s were the HC temperature reaches the maximum value, 45.64 o C after shutdown. On the other hand, if the scram system is unavailable, and at t = 47.33 s, the core flow decreases to 67.41 kg/s, the HC temperature increases to 164.02 o C, and the HC clad surface heat flux exceeds its critical value of 400.00 W/cm 2 resulting of fuel burnout. (author)

A Study on the Operation Strategy for Combined Accident including TLOFW accident

International Nuclear Information System (INIS)

Kim, Bo Gyung; Kang, Gook Young; Yoon, Ho Joon

2014-01-01

It is difficult for operators to recognize the necessity of a feed-and-bleed (F-B) operation when the loss of coolant accident and failure of secondary side occur. An F-B operation directly cools down the reactor coolant system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. The plant is not always necessary the F-B operation when the secondary side is failed. It is not necessary to initiate an F-B operation in the case of a medium or large break because these cases correspond to low RCS pressure sequences when the secondary side is failed. If the break size is too small to sufficiently decrease the RCS pressure, the F-B operation is necessary. Therefore, in the case of a combined accident including a secondary cooling system failure, the provision of clear information will play a critical role in the operators' decision to initiate an F-B operation. This study focuses on the how we establish the operation strategy for combined accident including the failure of secondary side in consideration of plant and operating conditions. Previous studies have usually focused on accidents involving a TLOFW accident. The plant conditions to make the operators confused seriously are usually the combined accident because the ORP only focuses on a single accident and FRP is less familiar with operators. The relationship between CET and PCT under various plant conditions is important to decide the limitation of initiating the F-B operation to prevent core damage

Applicability of Phebus FP results to severe accident safety evaluations and management measures

International Nuclear Information System (INIS)

Schwarz, M.; Clement, B.; Jones, A.V.

2001-01-01

The international Phebus FP (Fission Product) programme is the largest research programme in the world investigating core degradation and radioactive product release should a core meltdown accident occur in a light water reactor plant. Three integral experiments have already been performed. The experimental database obtained so far contains a wealth of information to validate the computer codes used for safety and accident management assessment

A review of MAAP4 code structure and core T/H model

International Nuclear Information System (INIS)

Song, Yong Mann; Park, Soo Yong

1998-03-01

The modular accident analysis program (MAAP) version 4 is a computer code that can simulate the response of LWR plants during severe accident sequences and includes models for all of the important phenomena which might occur during accident sequences. In this report, MAAP4 code structure and core thermal hydraulic (T/H) model which models the T/H behavior of the reactor core and the response of core components during all accident phases involving degraded cores are specifically reviewed and then reorganized. This reorganization is performed via getting the related models together under each topic whose contents and order are same with other two reports for MELCOR and SCDAP/RELAP5 to be simultaneously published. Major purpose of the report is to provide information about the characteristics of MAAP4 core T/H models for an integrated severe accident computer code development being performed under the one of on-going mid/long-term nuclear developing project. The basic characteristics of the new integrated severe accident code includes: 1) Flexible simulation capability of primary side, secondary side, and the containment under severe accident conditions, 2) Detailed plant simulation, 3) Convenient user-interfaces, 4) Highly modularization for easy maintenance/improvement, and 5) State-of-the-art model selection. In conclusion, MAAP4 code has appeared to be superior for 3) and 4) items but to be somewhat inferior for 1) and 2) items. For item 5), more efforts should be made in the future to compare separated models in detail with not only other codes but also recent world-wide work. (author). 17 refs., 1 tab., 12 figs

Disruption Physics and Mitigation on DIII-D

International Nuclear Information System (INIS)

Whyte, D.G.; Humphreys, D.A.; Kellman, A.G.

2005-01-01

The contributions of the DIII-D tokamak toward the understanding and control of disruptions are reviewed. Disruptions are found to be deterministic, and the underlying causes of disruption can therefore be predicted and avoided. With sufficiently rapid detection, possible damage from disruptions can be mitigated using an understanding of disruption phenomenology and plasma physics. Regimes of high β are readily available in DIII-D and provide access to relatively high energy density disruptions, despite DIII-D's moderate magnetic field and size. DIII-D, with all-graphite wall armor and wall conditioning between discharges, has proven highly resilient to the deleterious effects that disruptions can have on plasma operations. Simultaneously, exploitation and adaptation of DIII-D's extensive core and edge plasma diagnostic set have allowed for unique plasma measurements during disruptions. These measurements have tied into the development of several physical models used to understand aspects of disruptions, such as magnetohydrodynamic growth at the disruption onset, radiation energy balance through the thermal quench, and halo currents during the current quench. Based on this fundamental understanding, DIII-D has developed techniques to mitigate the harmful effects of disruptions by radiative dissipation of the plasma energy and extrapolated these techniques for possible use on larger devices like ITER

MDEP Common Position CP-EPRWG-04. Common position on EPR containment heat removal system in accident conditions

International Nuclear Information System (INIS)

2015-01-01

The importance of the integrity of the containment as a fundamental barrier to protect the people and environment against the effects of a nuclear accident is well established. In this regard, an essential objective is that the necessity for off-site counter-measures to reduce radiological consequences be limited or even eliminated. The design should provide engineering means to address those sequences which would otherwise lead to large or early releases, even in case of severe external hazards. The plant shall be designed so that it can be brought into a controlled and stable state and the containment function can be maintained, under accident conditions in which there is a significant amount of radioactive material in the containment, i.e. resulting from severe degradation of the reactor core. It is expected that due consideration to these requirements is to be given while tailoring long term loss of electrical power mitigation strategies. In order to reliably maintain the containment barrier, the regulators believe that: - safety features specifically designed for fulfilling safety functions required in core melt accidents shall be independent to the extent reasonably practicable from the Systems, Structures and Components (SSC) of the other levels of defense; - safety features specifically designed for fulfilling safety functions required in core melt accidents shall be safety classified and adequately qualified for the core melt accident environmental conditions for the time frame for which they are required to operate. In the light of the Fukushima Daiichi accident, the regulators believe that those safety features shall be designed with an adequate margin as compared to the levels of natural hazards considered for the site hazard evaluation; - the systems and components necessary for ensuring the containment function in a core melt accident shall have reliability commensurate with the function that they are required to fulfil. This may require redundancy of

Accident Source Terms for Pressurized Water Reactors with High-Burnup Cores Calculated using MELCOR 1.8.5.

Energy Technology Data Exchange (ETDEWEB)

Gauntt, Randall O.; Goldmann, Andrew; Kalinich, Donald A.; Powers, Dana A.

2016-12-01

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the effects of high fuel burnup in contrast with low burnup on fission product releases to the containment. Supporting this emphasis, recent data available on fission product release from high-burnup (HBU) fuel from the French VERCOR project are used in this study. The results of these analyses are treated as samples from a population of accident sequences in order to employ approximate order statistics characterization of the results. These trends and tendencies are then compared to the NUREG-1465 alternative source term prescription used today for regulatory applications. In general, greater differences are observed between the state-of-the-art calculations for either HBU or low-burnup (LBU) fuel and the NUREG-1465 containment release fractions than exist between HBU and LBU release fractions. Current analyses suggest that retention of fission products within the vessel and the reactor coolant system (RCS) are greater than contemplated in the NUREG-1465 prescription, and that, overall, release fractions to the containment are therefore lower across the board in the present analyses than suggested in NUREG-1465. The decreased volatility of Cs 2 MoO 4 compared to CsI or CsOH increases the predicted RCS retention of cesium, and as a result, cesium and iodine do not follow identical behaviors with respect to distribution among vessel, RCS, and containment. With respect to the regulatory alternative source term, greater differences are observed between the NUREG-1465 prescription and both HBU and LBU predictions than exist between HBU and LBU

[Trampoline accident with anterior knee dislocation caused popliteal artery disruption].

Science.gov (United States)

Pedersen, Peter Heide; Høgh, Annette Langager

2011-10-17

Only a few reports describe the risk of neurovascular damage following knee dislocation while trampolining. A 16 year-old male in a trampoline accident, sustained multi-ligament damage and occlusion of the popliteal artery. The occlusion did not show clinically until 24 hours after the trauma. He underwent vascular surgery (short saphenous bypass). We recommend implementing algorithms, for the management of suspected knee dislocation and possible accompanying neurovascular injuries in all trauma centers.

Chemical considerations in severe accident analysis

International Nuclear Information System (INIS)

Malinauskas, A.P.; Kress, T.S.

1988-01-01

The Reactor Safety Study presented the first systematic attempt to include fission product physicochemical effects in the determination of expected consequences of hypothetical nuclear reactor power plant accidents. At the time, however, the data base was sparse, and the treatment of fission product behavior was not entirely consistent or accurate. Considerable research has since been performed to identify and understand chemical phenomena that can occur in the course of a nuclear reactor accident, and how these phenomena affect fission product behavior. In this report, the current status of our understanding of the chemistry of fission products in severe core damage accidents is summarized and contrasted with that of the Reactor Safety Study

Swedish approach to information needs in severe accident situations

Energy Technology Data Exchange (ETDEWEB)

Soederman, E. (ES-Konsult AB, Stockholm (Sweden)); Karnik, P. (ES-Konsult AB, Stockholm (Sweden))

1992-07-01

In Sweden, systems for mitigating severe accidents have been installed at all plants and procedures have been implemented for accident management. This work has included the assessment of needs of information and the survivability of existing instrumentation during the various phases of an accident scenario. The approach has been pragmatic and based on existing knowledge of accident phenomenology and MAAP code calculations together with plant staff experience of detailed plant design and installation. During the early phases of accidents, which is defined to remain up to maximum fuel temperatures in the order of 800 C, the ordinary instrumentation is to a great extent useful. The reactor vessel level measurement is however identified to be weak in BWRs as soon as the core is partly uncovered. This has lead to the development of a Core Cooling Monitor. In later phases of accident scenarios, the general basis has been that no intrumentation inside the containment can survive. It has been analysed what information is strictly needed. It has been found that detailed information of the status inside the pressure vessel is of little importance after vessel penetration. Certain important information needs have been identified, that was not safely accessible from existing instrumentation. This had lead to complementary installations, using instruments inserted into the containment through protected guide tubes. Also for sampling of gas and water complementary installations have been made. (orig.)

Swedish approach to information needs in severe accident situations

International Nuclear Information System (INIS)

Soederman, E.; Karnik, P.

1992-01-01

In Sweden, systems for mitigating severe accidents have been installed at all plants and procedures have been implemented for accident management. This work has included the assessment of needs of information and the survivability of existing instrumentation during the various phases of an accident scenario. The approach has been pragmatic and based on existing knowledge of accident phenomenology and MAAP code calculations together with plant staff experience of detailed plant design and installation. During the early phases of accidents, which is defined to remain up to maximum fuel temperatures in the order of 800 C, the ordinary instrumentation is to a great extent useful. The reactor vessel level measurement is however identified to be weak in BWRs as soon as the core is partly uncovered. This has lead to the development of a Core Cooling Monitor. In later phases of accident scenarios, the general basis has been that no intrumentation inside the containment can survive. It has been analysed what information is strictly needed. It has been found that detailed information of the status inside the pressure vessel is of little importance after vessel penetration. Certain important information needs have been identified, that was not safely accessible from existing instrumentation. This had lead to complementary installations, using instruments inserted into the containment through protected guide tubes. Also for sampling of gas and water complementary installations have been made. (orig.)

Root causes and impacts of severe accidents at large nuclear power plants.

Science.gov (United States)

Högberg, Lars

2013-04-01

The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long-lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities.

Root Causes and Impacts of Severe Accidents at Large Nuclear Power Plants

International Nuclear Information System (INIS)

Hoegberg, Lars

2013-01-01

The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities

Root Causes and Impacts of Severe Accidents at Large Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Hoegberg, Lars

2013-04-15

The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from developing into severe core damage, even if they are initiated by very unlikely natural or man-made events, and, recognizing that accidents with severe core damage may nevertheless occur; (2) to prevent large-scale and long lived ground contamination by limiting releases of radioactive nuclides such as cesium to less than about 100 TBq. To achieve these objectives the importance of maintaining high global standards of safety management and safety culture cannot be emphasized enough. All three severe accidents discussed in this paper had their root causes in system deficiencies indicative of poor safety management and poor safety culture in both the nuclear industry and government authorities.

Innovations in systems engineering and analysis for the simulation of beyond design-base accidents

International Nuclear Information System (INIS)

Frisch, W.; Beraha, D.

1990-01-01

An important target in improving reactor safety is to have the most realistic simulation possible of beyond design-base accidents in the computer. This paper presents new developments in ATHLET and further developments (description of the thermo-fluid-dynamic conditions in the core and cooling circuits during serious incidents in the computer programme ATHLET-SA) and extensions (link-up to RALOC). RALOC is a computer programme for describing thermodynamic conditions inside the containment during design-base accidents and accidents involving core meltdown. Further research is dedicated to code acceleration. (DG) [de