MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

2014 ◽  
Author(s):  
Pavlin P. Groudev ◽  
Antoaneta E. Stefanova ◽  
Petya I. Vryashkova
Keyword(s):  
Parametric Study ◽  
Cold Water ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Fuel Cladding ◽  
Accident Management ◽  
Management Guidance ◽  
Fuel Behavior ◽  
Accident Conditions

This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed. The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident. The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.

The Analysis of Severe Accident Induced Steam Generator Tube Rupture and LERF Risk

2012 ◽  
Vol 614-615 ◽  
pp. 626-631
Author(s):  
Chang Hong Peng ◽  
Ying Hao Yang
Keyword(s):  
Steam Generator ◽  
Creep Rupture ◽  
Severe Accident ◽  
Frequency Model ◽  
Early Release ◽  
Accident Management ◽  
Management Guidance ◽  
The Mean ◽  
Accident Conditions ◽  
Thermal Hydraulic Analysis

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg and SG tubes. The first step performs thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step applies the creep rupture models to test the potential for the degraded SG to rupture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in LERF (Large and Early Release Frequency) model, and the overall LERF risk due to the Induced SGTR was calculated. In the final step, implementation of severe accident management guidance (SAMG), such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the result of induced SGTR and LERF effectively.

Natural Circulation and Creep Damage Analyses of Severe Accident for a Typical PWR Plant

2009 ◽  
Author(s):  
Osamu Kawabata ◽  
Masao Ogino
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Reactor Core ◽  
Creep Damage ◽  
Reactor Vessel ◽  
Reactor System ◽  
Severe Accident ◽  
Steam Generator Tube ◽  
Hot Gas ◽  
Core Meltdown

When the primary reactor system remain pressurized during core meltdown for a typical PWR plant, loop seals formed in the primary reactor system would lead to natural circulations in hot leg and steam generator. In this case, the hot gas released from the reactor core moves to a steam generator, and a steam generator tube would be failed with cumulative creep damage. From such phenomena, a high-pressure scenario during core meltdown may lead to large release of fission products to the environment. In the present study, natural circulation and creep damage in the primary reactor system accompanying the hot gas generation in the reactor core were discussed and the combining analysis with MELCOR and FLUENT codes were performed to examine the natural circulation behavior. For a typical 4 loop PWR plant, MELCOR code which can analyze for the severe accident progression was applied to the accident analyses from accident initiation to reactor vessel failure for the accident sequence of the main steam pipe break which is maintained at high pressure during core meltdown. In addition, using the CFD code FLUENT, fluid dynamics in the reactor vessel plenum, hot leg and steam generator of one loop were simulated with three-dimensional coordinates. And the hot gas natural circulation flow and the heat transfer to adjoining structures were analyzed using results provided by the MELCOR code as boundary conditions. The both ratios of the natural circulation flow calculated in the hot leg and the steam generator using MELCOR code and FLUENT code were obtained to be about 2 (two). And using analytical results of thermal hydraulic analysis with both codes, creep damage analysis at hottest temperature points of steam generator tube and hot leg were carried out. The results in both cases showed that a steam generator tube would be failed with creep rupture earlier than that of hot leg rupture.

External Cooling: The SWR 1000 Severe Accident Management Strategy

2004 ◽  
Author(s):  
Nikolay Ivanov Kolev
Keyword(s):  
Management Strategy ◽  
External Pressure ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Core Flooding ◽  
External Cooling ◽  
Structural Resistance ◽  
Safety Margins ◽  
Accident Management ◽  
Management Concept

This paper provides the description of the basics behind design features for the severe accident management strategy of the SWR 1000. The hydrogen detonation/deflagration problem is avoided by containment inertization. In-vessel retention of molten core debris via water cooling of the external surface of the reactor vessel is the severe accident management concept of the SWR 1000 passive plant. During postulated bounding severe accidents, the accident management strategy is to flood the reactor cavity with Core Flooding Pool water and to submerge the reactor vessel, thus preventing vessel failure in the SWR 1000. Considerable safety margins have been determined by using state of the art experiment and analysis: regarding (a) strength of the vessel during the melt relocation and its interaction with water; (b) the heat flux at the external vessel wall; (c) the structural resistance of the hot structures during the long term period. Ex-vessel events are prevented by preserving the integrity of the vessel and its penetrations and by assuring positive external pressure at the predominant part of the external vessel in the region of the molten corium pool.

CFD Simulation of PANDA and MISTRA Cooler Tests of ERCOSAM-SAMARA Project

2014 ◽  
Author(s):  
A. S. Filippov ◽  
S. Y. Grigoryev ◽  
O. V. Tarasov ◽  
T. A. Iudina
Keyword(s):  
Cfd Simulation ◽  
Molar Fraction ◽  
Transient Behavior ◽  
Severe Accident ◽  
Steam Condensation ◽  
Multi Stage ◽  
Accident Management ◽  
Accident Conditions ◽  
Good Agreement ◽  
Three Phases

The ERCOSAM and SAMARA projects (EURATOM (EU) and ROSATOM (Russia)) include a set of multi-stage experiments carried out at different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). The tests sequences are aimed at investigating hydrogen concentration build-up and stratification during a postulated severe accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and passive auto-catalytic recombiners. Each test includes four phases, of which the first three phases simulate the establishment of severe accident conditions in NPP containment (injection of steam and helium (simulator of hydrogen), stratification of the gas mixture). During the fourth phase of the experiment one of the SAMs simulators is activated. All experiments were simulated at Nuclear Safety Institute of the Russian Academy of Science (IBRAE RAN) with FLUENT and, partially, OpenFOAM codes. In this paper the tests with coolers carried out on PANDA and MISTRA facilities are considered. Their simulations required development of a set of models of volumetric and near-wall condensation phenomena. The models were validated vs. already known tests and vs. integrated experiments of ERCOSAM-SAMARA projects. A brief description of the models and the used CFD methods is provided. Then the results of simulations of the four phases of the tests are presented. Some peculiarities of gas motion and helium distribution obtained in the experiments as well as in their simulations are analyzed. These phenomena concern steam condensation and helium redistribution by convective flows due to the cooler activation in the installation. Local ‘pockets’ of helium are formed with a molar fraction larger than the maximum achieved at the first three phases of the experiments. The accounting of initial and boundary conditions along with calibration of the models provided as a whole a good agreement between calculations and experimental data on transient behavior of gas composition in the facility at the first three phases and at the final fourth phase.

Experimental Study on the Improved In-Vessel Corium Retention Concepts for the Severe Accident Management

2002 ◽  
Author(s):  
K. H. Kang ◽  
R. J. Park ◽  
K. M. Koo ◽  
S. B. Kim ◽  
H. D. Kim
Keyword(s):  
Heat Removal ◽  
Elevated Pressure ◽  
Reactor Vessel ◽  
Experimental Results ◽  
Severe Accident ◽  
Experimental Facility ◽  
Dual Strategy ◽  
Accident Management ◽  
Lower Head ◽  
Inner Diameter

Feasibility experiments were performed for the assessment of improved In-Vessel Corium Retention (IVR) concepts using an internal engineered gap device and also a dual strategy of In/Ex-vessel cooling using the LAVA experimental facility. The internal engineered gap device made of carbon steel was installed inside the LAVA lower head vessel and it made a uniform gap with the vessel by 10 mm. In/Ex-vessel cooling in the dual strategy experiment was performed installing an external guide vessel outside the LAVA lower head vessel at a uniform gap of 25 mm. The LAVA lower head vessel was a hemispherical test vessel simulated with a 1/8 linear scale mock-up of the reactor vessel lower plenum with an inner diameter of 500 mm and thickness of 25 mm. In both of the tests, Al2O3 melt was delivered into about 50K subcooled water inside the lower head vessel under the elevated pressure. Temperatures of the internal engineered gap device and the lower head vessel were measured by K-type thermocouples embedded radially in the 3mm depth of the lower head vessel outer surface and in the 4mm depth of the internal engineered gap device, respectively. In the dual strategy experiment, the Ex-vessel cooling featured pool boiling in the gap between the lower head vessel and the external guide vessel. It could be found from the experimental results that the internal engineered gap device was intact and so the vessel experienced little thermal and mechanical attacks in the internal engineered gap device experiment. And also the vessel was effectively cooled via mutual boiling heat removal in- and ex-vessel in the dual strategy experiment. Compared with the previous LAVA experimental results performed for the investigation of the inherent in-vessel gap cooling, it could be confirmed that the Ex-vessel cooling measure was dominant over the In-vessel cooling measure in this study. It is concluded that the improved cooling measures using a internal engineered gap device and a dual strategy promote the cooling characteristics of the lower head vessel and so enhance the integrity of the vessel in the end.

Analysis on Ex-Vessel Steam Explosion in CPR1000+ Unit

2010 ◽  
Author(s):  
Juanhua Zhang ◽  
Jiming Lin ◽  
Shishun Zhang
Keyword(s):  
Physical Model ◽  
Steam Explosion ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Cavity Wall ◽  
Severe Accident ◽  
Severe Accidents ◽  
Structural Capacity ◽  
Impulse Load

Reactor Pit Flooding System (RPF) is adopted under the severe accidents situation in CPR1000+ units. It can move the heat generated from the reactor core via external reactor vessel cooling (ERVC) to keep the integrity of RPV and achieve the in-vessel corium retention (IVR). But if IVR function of RPF is failed, there is Ex-Vessel Steam Explosion (EX-SE) risk. The Ex-Vessel Steam Explosion is analyzed by MC3D software which is for fuel and cooling interaction (FCI). The physical model of CPR1000+ for Steam Explosion is built firstly and then the phenomenon of Ex-Vessel Steam Explosion under typical severe accident is analyzed. The conclusion of this study is that the impulse load of pressure on the cavity wall induced by steam explosion is about 310KPas ∼ 440KPas. Referencing the structure capacity of AP600 containment, if the structural capacity of CPR1000+ containment is equal to AP600, the impulse load of pressure is lower than it. So it could be preliminarily estimated that steam explosion will not threaten the integrality of CPR1000+ containment.

Numerical Simulation of Concrete Ablation and Corium Cooling for Molten Corium-Concrete Interaction (MCCI)

2020 ◽  
Author(s):  
Zhichun Xu ◽  
Yapei Zhang ◽  
G. H. Su ◽  
Wenxi Tian ◽  
Suizheng Qiu
Keyword(s):  
Water Injection ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Core Material ◽  
Melt Temperature ◽  
Concrete Type ◽  
Depth Study ◽  
Water Reactors

Abstract In a postulated severe accident situation in Light Water Reactors (LWRs), if the core fuel cannot be effectively cooled, the reactor core material will be heated and form a molten corium in the lower head. When the lower plenum of the reactor vessel fails, the molten corium may flow into the cavity under the reactor vessel and react with the concrete. This process, known as Molten Corium Concrete Interaction (MCCI), is characterized by concrete ablation and oxidation of metal in the corium, both of which produce a large amount of combustible and non-condensable gases, threatening the integrity of the containment. Thus in-depth study of the characteristics of concrete ablation and corium cooling have great significance. In the present study, an MCCI analysis code, MOQUICO (molten corium concrete interaction and corium cooling code, QUI means quintic) has been developed. The MACE M3b and OECD/MCCI CCI-3 tests were analyzed to validate the developed code. The melt temperature, axial and radial ablation depths, upward heat flux were calculated and were in good agreement with the experimental measurements, which proved that the code is capable of simulating MCCI and related phenomena of LWRs. Sensitivity analyses on the factors of decay heat, concrete type and water injection moment were performed and analyzed.

Severe Accident Steady-State Modeling of Advanced Nuclear Power Plant Based on ASTEC Code

2017 ◽  
Author(s):  
Wei Song ◽  
Jiaxu Zuo ◽  
Yan Chen ◽  
Chaojun Li ◽  
Peng Zheng
Keyword(s):  
Steady State ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Cooling Systems ◽  
Accident Management ◽  
Management Guidance ◽  
Level 2 ◽  
Reactor Block

Severe accident is an attractive topic today for the nuclear power plant (NPP) safety. In the nuclear safety regulatory work, it is planned to build a full scale severe accident model for the advanced nuclear power plant of China to study the new designs of severe accident prevention and mitigation systems and strategies, and to further deploy the application on the level 2 PSA and severe accident management guidance. This paper firstly introduces the modeling tool, ASTEC, and then presents the progress of modeling work, which is mainly on the steady state modeling and regulation including reactor block, primary and secondary cooling systems, regulation systems etc. Last but not least, the work plan for the future is given.

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