scholarly journals CFD Recombiner Modelling and Validation on the H2-Par and Kali-H2Experiments

2011 ◽  
Vol 2011 ◽  
pp. 1-13 ◽  
Author(s):  
Stéphane Mimouni ◽  
Namane Mechitoua ◽  
Mehdi Ouraou
Keyword(s):  
Gaseous Mixture ◽  
Hydrogen Gas ◽  
Severe Accident ◽  
Experimental Program ◽  
Pressurized Water Reactor ◽  
Gas Temperature ◽  
Flammability Limit ◽  
Pressurized Water ◽  
Containment Integrity ◽  
Good Agreement

A large amount of Hydrogen gas is expected to be released within the dry containment of a pressurized water reactor (PWR), shortly after the hypothetical beginning of a severe accident leading to the melting of the core. According to local gas concentrations, the gaseous mixture of hydrogen, air and steam can reach the flammability limit, threatening the containment integrity. In order to prevent mechanical loads resulting from a possible conflagration of the gas mixture, French and German reactor containments are equipped with passive autocatalytic recombiners (PARs) which preventively oxidize hydrogen for concentrations lower than that of the flammability limit. The objective of the paper is to present numerical assessments of the recombiner models implemented in CFD solvers NEPTUNE_CFD and Code_Saturne. Under the EDF/EPRI agreement, CEA has been committed to perform 42 tests of PARs. The experimental program named KALI-H2, consists checking the performance and behaviour of PAR. Unrealistic values for the gas temperature are calculated if the conjugate heat transfer and the wall steam condensation are not taken into account. The combined effects of these models give a good agreement between computational results and experimental data.

Analysis of RCS Depressurization Effects on Containment Temperature for Qinshan Phase II Nuclear Power Plant

2014 ◽  
Author(s):  
Liu Lili ◽  
Zhang Ming ◽  
Deng Jian
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Hydrogen Generation ◽  
Hydrogen Combustion ◽  
Severe Accident ◽  
Generation Rate ◽  
Pressurized Water Reactor ◽  
Gas Temperature ◽  
Pressurized Water

A severe accident code was applied for modeling of a typical pressurized water reactor (PWR) nuclear power plant, and the effects of RCS depressurization on the gas temperature of the relief tank cell in the containment during a station blackout (SBO) induced accident was analyzed. The sensitivity calculation indicated that the hydrogen generation rate obviously increased due to RCS depressurization in a critical stage. The results show that RCS depressurization can play an important role in hydrogen generation rate and total accumulation, and the temperature of the containment atmosphere is highly influenced by hydrogen combustion. High temperature induced by hydrogen combustion may degrade the equipment and instruments capabilities. Based on this analysis, a feasible strategy of RCS depressurization for mitigating the accident consequence is provided for developing the capacity of the SBO treatment of Qinshan Phase Nuclear Power Plant (QSP-II NPP).

Analysis of Aerosol Collection by Droplets: Application to Fission Products Removal in Case of Severe Accident

2008 ◽  
Author(s):  
Emmanuel Porcheron ◽  
Pascal Lemaitre
Keyword(s):  
Fission Products ◽  
Severe Accident ◽  
Experimental Program ◽  
Pressurized Water Reactor ◽  
Flow Conditions ◽  
Detailed Characterization ◽  
Two Phase ◽  
Pressurized Water ◽  
Mass Transfers

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Suˆrete´ Nucle´aire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulate typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to study the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between the spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

Fog Inerting Effects on Hydrogen Combustion in a PWR Ice Condenser Containment

1995 ◽  
Vol 117 (2) ◽  
pp. 502-507 ◽  
Author(s):  
W. Luangdilok ◽  
R. B. Bennett
Keyword(s):  
Test Data ◽  
Combustible Mixture ◽  
Pressurized Water Reactor ◽  
Flammability Limit ◽  
Flammability Limits ◽  
Pressurized Water ◽  
Large Loss ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Good Agreement

A mechanistic fog inerting model has been developed to account for the effects of fog on the upward lean flammability limits of a combustible mixture based on the thermal theory of flame propagation. Benchmarking of this model with test data shows reasonably good agreement between the theory and the experiment. Applications of the model and available fog data to determine the upward lean flammability limits of the H2–air–steam mixture in the ice condenser upper plenum region of a pressurized water reactor (PWR) ice condenser containment during postulated large loss of coolant accident (LOCA) conditions indicate that combustion may be suppressed beyond the downward flammability limit (8 percent H2 by volume).

Analysis of a Loss of Feedwater Case at the Ringhals-3 NPP Using RELAP5/PARCS Coupled Codes

2008 ◽  
Author(s):  
Jo´zsef Ba´na´ti ◽  
Mathias Sta˚lek ◽  
Christophe Demazie`re ◽  
Magnus Holmgren
Keyword(s):  
Temperature Fields ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
History Effects ◽  
Coupled Codes ◽  
Fuel Assemblies ◽  
Development And Validation ◽  
Secondary Side ◽  
Relap5 Code ◽  
Good Agreement

This paper deals with the development and validation of a coupled RELAP5/PARCS model of the Swedish Ringhals-3 pressurized water reactor against a Loss of Feedwater transient, which occurred on August 16, 2005. At first, the stand-alone RELAP5 and PARCS models are presented. All the 157 fuel assemblies are modeled in individually in both codes. The model is furthermore able to handle possible asymmetrical conditions of the flow velocity and temperature fields between the loops. On the neutronic side, the dependence of the material constants on history effects, burnup, and instantaneous conditions is accounted for, and the full heterogeneity of the core is thus taken into account. The reflectors are also explicitly represented. The coupling between the two codes is touched upon, with emphasis on the mapping between the hydrodynamic/heat structures and the neutronic nodes. The transient was initiated by a malfunction of the feedwater valve at the 2nd steam generator. Consequently, the turbines were tripped and, because of the low level in the SG-2 the reactor was scrammed. Activation of the auxiliary feedwater provided proper amount of cooling from the secondary side, resulting in safe shutdown conditions. Capabilities of the RELAP5 code were more challenged in this transient, where the influences of the feedback from the neutron kinetic side were also taken into account in the analysis. The calculated values of the parameters show good agreement with the measured data.

Comparative Calculations of TROI TS-2 and TS-3 Steam Explosion Experiments With TEXAS-V

2017 ◽  
Author(s):  
Taehoon Kim ◽  
Sukyoung Pak ◽  
Yongjin Cho
Keyword(s):  
Void Fraction ◽  
Steam Explosion ◽  
Peak Pressure ◽  
Severe Accident ◽  
Experiment Data ◽  
Trigger Time ◽  
Vessel Reactor ◽  
The Moment ◽  
Containment Integrity ◽  
Good Agreement

During a severe accident, contact of the molten corium with the coolant water may cause an energetic steam explosion which is a rapid increase of explosive vaporization by transfer to the water of a significant part of the energy in the corium melt. This steam explosion has been considered as an adverse effect when the water is used to cool the molten corium and could threaten reactor vessel, reactor cavity, containment integrity. In this study, TROI TS-2 and TS-3 experiments as part of the OECD/SERENA-2 project were analyzed with TEXAS-V. Input parameters were based on actual TROI experiment data. In mixing simulations, calculated results were compared to melt front behavior, void fraction in trigger time and other parameters in experiment results. In explosion simulations, corresponding to TROI experiments an external triggering was employed at the moment that melt front reached heights of 0.4 m. Calculated results of peak pressure and impulse at the bottom were compared with TROI experiment results. Melt front behaviors of the melt was different from the experimental results in both TS-2 and TS-3. Void fraction in triggering time in TS-2 was in good agreement with the experiment results and in TS-3 was slightly overestimated. The peak pressure and impulse at bottom were successfully predicted by TEXAS-V. These calculations will allow establishing whether the limitations and differences observed in the simulations of the experiments are important for the reactor case.

scholarly journals Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

Study of the Ageing Effects on the Lower Head Failure in a PWR Reactor

2020 ◽  
Author(s):  
R. Lo Frano ◽  
S. Paci ◽  
P. Darnowski ◽  
P. Mazgaj
Keyword(s):  
Nuclear Power ◽  
Quantitative Estimation ◽  
Computer Code ◽  
Severe Accident ◽  
Symmetric Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Term Operation ◽  
Ageing Effects ◽  
Lower Head

Abstract The paper studies influence the ageing effects on the failure of a Reactor Pressure Vessel (RPV) during a severe accident with a core meltdown in a Nuclear Power Plant (NPP). The studied plant is a generic high-power Generation III Pressurized Water Reactor (PWR) developed in the frame of the EU NARSIS project. A Total Station Blackout (SBO) accident was simulated with MELCOR 2.2 severe accident integral computer code. Results of the analysis, temperatures in the lower head and pressures in the lower plenum were used as initial and boundary conditions for the Finite Element Method (FEM) simulations. Two FEM models were developed, a simple two-dimensional axis-symmetric model of the lower head to study fundamental phenomena and complex 3D model to include interactions with the RPV and reactor internals. Ageing effects of a lower head were incorporated into the FEM models to investigate its influence onto lower head response. The ageing phenomena are modelled in terms of degraded mechanical material properties as σ(T), E(T). The primary outcome of the study is the quantitative estimation of the influence of ageing process onto the timing of reactor vessel failure. Presented novel methodology and results can have an impact on future consideration about Long-Term Operation (LTO) of NPPs.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

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