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

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.

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

Advanced Materials Research ◽

10.4028/www.scientific.net/amr.614-615.626 ◽

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.

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Prediction of Hydrogen Distribution Under Severe Accident Conditions With CFD and Lumped Parameter Codes

Volume 3: Thermal-Hydraulics; Turbines, Generators, and Auxiliaries ◽

10.1115/icone20-power2012-54499 ◽

2012 ◽

Cited By ~ 2

Author(s):

Tobias Szabó ◽

Stefan Benz ◽

Frank Kretzschmar ◽

Peter Royl ◽

Thomas Jordan

Keyword(s):

Source Term ◽

Severe Accident ◽

Test Bed ◽

Hydrogen Distribution ◽

Water Steam ◽

Lumped Parameter ◽

Accident Conditions ◽

Energy Balances ◽

Mass And Energy Balances ◽

Good Agreement

In case of a severe accident, the containment is the ultimate barrier to the environment. Therefore, reliable simulations tools for containment thermal hydraulics, including hydrogen distribution are indispensable. We simulated the behavior of the containment atmosphere under severe accident conditions with a postulated source term of water, steam and hydrogen. We used a detailed 3D CFD code (GASFLOW) and a lumped parameter code (MELCOR) in order to compare and assess their modeling capabilities. A simplified generic containment including all important components was used as a test bed. We analyzed the calculated pressure histories, mass and energy balances, convective flow as well as steam and hydrogen distributions. Integral values were modeled in good agreement by both codes. The overall flow was reasonably predicted. However we observed discrepancies in the calculated steam and hydrogen concentrations.

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Innovative Practice of Severe Accidents Measures in CAP1400

Volume 7: Fuel Cycle, Decontamination and Decommissioning, Radiation Protection, Shielding, and Waste Management; Mitigation Strategies for Beyond Design Basis Events ◽

10.1115/icone25-66605 ◽

2017 ◽

Author(s):

Likai Fang ◽

Xin Liu ◽

Guobao Shi

Keyword(s):

Nuclear Power ◽

Heat Removal ◽

Severe Accident ◽

Mitigation Measures ◽

Spent Fuel ◽

Fukushima Accident ◽

Power Sources ◽

Severe Accidents ◽

Accident Management ◽

Accident Conditions

CAP1400 is GenIII passive PWR, which was developed based on Chinese 40 years of experience in nuclear power R&D, construction&operation, as well as introduction and assimilation of AP1000. Severe accidents prevention and mitigation measures were systematically considered during the design and analysis. In order to accommodate high power and further improve the safety of the plant, also considering feedback from Fukushima accident, some innovative measures and design requirements were also applied. Based on the probabilistic&deterministic analysis and engineering judgment, considerable severe accidents scenarios were considered. Both severe accidents initiated at power and shutdown condition were analyzed. Insights were also obtained to decide the challenge to the plant. All known severe accidents phenomena and their treatment were considered in the design. In vessel retention (IVR) was applied as one of the severe accident mitigation measures. To improve the margin of IVR success and verify the heat removal capability through reactor pressure vessel, both design innovative measures and experiments were used. The melt pool behavior and corium pool configuration were also studied by using CFD code and thermodynamic code. Hydrogen risk was mitigated by installation of hydrogen igniters, which were comprised of two serials, and were powered by multiple power sources. To further improve the safety, six extra hydrogen passive recombiners were also added in the containment. Hydrogen risk was analyzed both inside containment and outside containment considering leakage effect. Other severe accident phenomena were also considered by designed or analyzed to show the containment robustness to accommodate it. As one of the Fukushima accident feedback, full scope severe accident management guideline were developed by considering both power condition and shutdown condition, accident management for spent fuel pool was also considered. As the basis of accident management during severe accidents, survivability of equipments and instruments that are necessary in severe accident were assessed and will be further tested and/or analyzed. Such tests will consider severe accident conditions arised from hydrogen combustion.

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Current Severe Accident Research and Development Topics for Wolsong PHWR Safety in Korea

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4035690 ◽

2017 ◽

Vol 3 (2) ◽

Author(s):

Y. M. Song ◽

D. H. Kim ◽

S. Y. Park ◽

J. H. Song

Keyword(s):

Stress Test ◽

Severe Accident ◽

Test Results ◽

Computer Tool ◽

Accident Management ◽

Decision Supporting ◽

Accident Research ◽

Protection Capability ◽

Accident Conditions ◽

Management Expert

In Korea, pressurized heavy water-cooled reactors (PHWR) account for 17% of operating units and have taken an important role in providing national energy supply. The recent biggest issue in domestic PHWR community was the continued operation of the Wolsong-1 CANada Deuterium Uranium (CANDU) plant, which has recently been approved to operate for 10 more years after a 30 year design life. In relation to this issue, various actions from both post-Fukushima lessons and Wolsong-1 (WS1) stress test results are being taken. In KAERI R&D, the following topics are studied to support the basis for these actions. First, PHWR severe accident issues such as (1) primary heat transport system (PHTS) overpressure protection capability, (2) containment overpressure protection capability, and (3) bypass source term are evaluated. Second, a computer tool (called MAAP–ISAAC) has been modified and updated to support analyzing Wolsong severe accident issues. Third, a decision supporting tool, called Severe Accident Management Expert (SAMEX)–CANDU, has been developed to aid emergency response experts under severe accident conditions.

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

Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security ◽

10.1115/icone22-30941 ◽

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.

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Calculation of Evaporation From Fukushima NPP Spent Fuel Pools

Volume 8B: Heat Transfer and Thermal Engineering ◽

10.1115/imece2018-86561 ◽

2018 ◽

Author(s):

Mirza M. Shah

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Severe Accident ◽

Spent Fuel ◽

Decay Heat ◽

Wide Range ◽

Spent Fuel Pool ◽

Accident Conditions ◽

Safety Systems ◽

Good Agreement

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and post-accident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The post-accident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25 % lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

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