scholarly journals Study of hydrodynamics performance of a multi-venturi filtering system for BWR severe accident venting strategies

2020 ◽  
Vol 35 (1) ◽  
pp. 16-23
Author(s):  
Alejandro Reyes-Garcia ◽  
Eduardo Sainz-Mejia ◽  
Javier Ortiz-Villafuerte ◽  
Javier Palacios-Hernandez ◽  
Roberto Lopez-Solis
Keyword(s):  
Pressure Drop ◽  
Mass Flow ◽  
Reference Model ◽  
Geometrical Model ◽  
Severe Accident ◽  
Cfd Models ◽  
Station Blackout ◽  
Experimental Values ◽  
Good Agreement

The aim of this project was to determine the capacity of a multi-venturi scrubber filtering system to cope with vented gas mass-flow rate coming from a BWR Mark II primary containment during a long-term station blackout. The multi-venturi filtering system CFD models were developed in the environment of the open source platforms SALOME and OpenFoam. The first geometrical model was created based on the dimensions of a well-known experimental setup, and the results of the pressure drop along the streamwise co-ordinate showed a maximum difference of 10 % in relation to the experimental values for different cases of liquid to gas mass ratios. Then a full scale multi-venturi model was developed. To study the performance of this system during conditions expected in a severe accident, a gas mixture similar to that occurring in a BWR Mark II containment at venting pressure was used as inlet gas. The gas mass-flow that can be cleansed by individual venturis and the pressure required to activate those venturis were computed. The pressure drop profiles in each sector were also determined as the function of different liquid loadings. The results showed good agreement with the capacity of the design taken as the reference model.

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

2017 ◽  
Author(s):  
Jun Ishikawa ◽  
Tomoyuki Sugiyama ◽  
Yu Maruyama
Keyword(s):  
Source Term ◽  
Severe Accident ◽  
Water Pool ◽  
Energy Agency ◽  
Severe Accidents ◽  
Station Blackout ◽  
Carbon Dioxide Co2 ◽  
Accident Sequences ◽  
Term Analysis

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

Simulation of Hydrogen Distribution due to In-Vessel Severe Accident in WWER-1000 NPP Containment: A Comparison of CONTAIN and MELCOR Codes Results

2018 ◽  
Author(s):  
Omid Noorikalkhoran ◽  
Massimiliano Gei
Keyword(s):  
Severe Accident ◽  
Second Step ◽  
Hydrogen Distribution ◽  
Short Term ◽  
Long Time ◽  
Design Basis Accident ◽  
Safety Features ◽  
Design Pressure ◽  
Good Agreement

During a severe accident or Beyond Design Basis Accident (BDBA), the reaction of water with zirconium alloy as fuel clad, radiolysis of water, molten corium-concrete interaction (MCCI) and post-accident corrosion can generate a source of hydrogen. In the present work, hydrogen distribution due to in-vessel reaction (between zircaloy and steam) has been simulated inside a WWER-1000 reactor containment. In the first step, the thermal hydraulic parameters of containment have been simulated for a DECL (Double Ended Cold Leg) accident (DBA phase) in both short and long time and the effects of spray as Engineering Safety Features (ESFs) on mitigating the parameters have been studied. In the second step, it has been assumed that the accident developed into an in-vessel core melting accident. While in pre-phase of core melting (severe accident phase), hydrogen will be produced as a result of zircaloy and steam reaction (BDBA phase), the hydrogen distribution has been simulated for 23 cells inside the reactor containment by using CONTAIN 2.0 (Best estimate code) and MELCOR 1.8.6 codes. Finally, the results have been compared to FSAR results. As it can be seen from the comparisons, both CONTAIN and MELCOR codes can predict the results in good agreement with FSAR (ANGAR code) results. CONTAIN shows peak pressure around 0.36 MPa in short-term and this amount is about 0.38 and 0.4 MPa for MELCOR and ANGAR (FSAR) results respectively. All these values are under design pressure that is around 0.46 MPa. Cell 20 has the maximum mole fraction of hydrogen in long-term about 9.5% while the maximum amount of hydrogen takes place in cell 22. The differences between the results of codes are because of different equations, Models, Numerical methods and assumptions that have been considered by the codes. The simulated Hydrogen Distribution Map (HDM) can be used for upgrading the location of HCAV systems and Hydrogen Mitigator features (like the recombiners and ignitors) inside the containment to reduce the risk of hydrogen explosion.

Severe Accident Context Quantification for Long-Term Station Blackout in Boiling Water Reactor Nuclear Power Plants

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Gueorgui I. Petkov ◽  
Monica Vela-Garcia
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Accident Scenario ◽  
Station Blackout ◽  
Fukushima Daiichi

The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the performance evaluation of teamwork (PET) procedure for dynamic context quantification and determination of alternatives, coordination, and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions with the use of thermo-hydraulic (TH) model and severe accident (SA) codes (melcor and maap). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and a hypothetic unmitigated LT SBO at peach bottom #1 boiling water reactor (BWR) reactor nuclear power plants (NPPs). The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, “State-of-the-Art Reactor Consequence Analysis” and TH calculations made by using maap code at the EC Joint Research Centre.

scholarly journals Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

2013 ◽  
Vol 2013 ◽  
pp. 1-15 ◽  
Author(s):  
Andrej Prošek ◽  
Leon Cizelj
Keyword(s):  
Management Strategies ◽  
Severe Accident ◽  
Stress Tests ◽  
Pressurized Water ◽  
Makeup Water ◽  
Reactor Coolant ◽  
The Core ◽  
Station Blackout ◽  
Accident Sequences

Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs) leaks assumed) to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS). For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

scholarly journals Safety Analyses of VVER-1200/V491 reactor for longterm station blackout along with small LOCAs

2016 ◽  
Vol 6 (4) ◽  
pp. 8-17
Author(s):  
Thi Hoa Bui ◽  
Tan Hung Hoang ◽  
Minh Giang Hoang
Keyword(s):  
Steam Generator ◽  
Heat Removal ◽  
Safety Analysis ◽  
Severe Accident ◽  
Management Guideline ◽  
Accident Management ◽  
Station Blackout

Performance of  Passive Heat Removal through Steam Generator (PHRS-SG) of VVER-1200/V491 reactor presented in Safety Analysis Report for Ninh Thuan 1 shows that in case of long term station black out (SBO),  VVER-1200/V491 reactor can be cooldown and remained in safety mode at least 24 hours based on PHRS-SG performance. Anyway, long term station blackout along with small break in main coolant pipe of VVER-1200/V491 is assumed to be happening as an extension design condition that needs to be investigated. This study focuses on investigation of SBO along with different size of small break of LOCAs with expectation of finding the range of break size that the reactor is still kept in safety mode during 24 hours. During the investigation, some indicators for fuel damage such as the timing of HA1 actuation or mass of coolant inventory discharged are introduced as necessary information contributed to Severe Accident Management Guideline (SAMG).

scholarly journals Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

2007 ◽  
Vol 2007 ◽  
pp. 1-9
Author(s):  
Algirdas Kaliatka ◽  
Eugenijus Uspuras ◽  
Sigitas Rimkevicius
Keyword(s):  
Cooling System ◽  
Severe Accident ◽  
Diesel Generator ◽  
Core Damage ◽  
Design Basis Accident ◽  
Station Blackout ◽  
Core Cooling ◽  
Key Events ◽  
Channel Type

Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed.

Severe Accident Context Evaluation for BWR NPPS Based on Long-Term Station Blackout

2016 ◽  
Author(s):  
Gueorgui I. Petkov ◽  
Monica Vela-Garcia
Keyword(s):  
Safety Assessment ◽  
Power Plants ◽  
Human Performance ◽  
Severe Accident ◽  
Research Centre ◽  
Joint Research ◽  
Accident Scenario ◽  
Station Blackout ◽  
Fukushima Daiichi

The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the Performance Evaluation of Teamwork procedure for dynamic context quantification and determination of alternatives, coordination and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions (timelines) with the use of thermo-hydraulic model and severe accident codes (MELCOR and MAAP). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and an hypothetic unmitigated LT SBO at Peach Bottom #1 Boiling Water Reactor Nuclear Power Plants. The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, “State-of-the-Art Reactor Consequence Analysis” and thermo-hydraulic calculations made by using MAAP code at the EC Joint Research Centre, Institute for Energy and Transport, Nuclear Reactor Safety Assessment Unit.

scholarly journals TRANSIENT CALCULATIONS OF SPERT III EXPERIMENTS

2021 ◽  
Vol 247 ◽  
pp. 07017
Author(s):  
Andreas Pautz ◽  
Winfried Zwermann
Keyword(s):  
Cross Sections ◽  
Reference Model ◽  
Neutron Transport ◽  
Control Rod ◽  
Combined Application ◽  
Experimental Values ◽  
Core Characteristics ◽  
Control Rods ◽  
Static Core ◽  
Good Agreement

Cold-startup and hot-standby reactivity accident tests conducted at the SPERT III E-core research reactor are analysed with the coupled neutron-kinetic/thermal-hydraulic code system DYN3D-ATHLET. Homogenised 2-group cross sections for DYN3D are thereby generated with the Monte Carlo neutron transport code Serpent 2 in combination with the ENDF/B-VII.1 cross section library. Results in terms of maximum power, energy release, and reactivity compensation are in good agreement with the experimental values. The time-dependent contributions to the reactivity feedback are investigated for both a cold-startup test and a hot-standby test. These findings prove the suitability of the combined application of the simulation codes to predict the reactor dynamic behaviour in the event of prompt-critical and super-prompt critical transients even for small reactor cores. Furthermore, static core characteristics of the SPERT III E-core reactor at cold-startup condition are analysed with using a static DYN3D model, a detailed Serpent reference model, and a simplified Serpent model consistent with the DYN3D model. The critical control rod position and the excess reactivities of both the control rods and the transient rod obtained with the Serpent reference model are consistent with the experimental values. For the same parameters, the DYN3D model is in good agreement with the Serpent simplified model.

Analyses of Feedwater Trip With SBO Sequence of VVER1000 Reactor

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Guido Mazzini ◽  
Miloš Kynčl ◽  
Marek Ruščák
Keyword(s):  
Nuclear Power ◽  
Lessons Learned ◽  
Severe Accident ◽  
Research Centre ◽  
Stress Tests ◽  
Research Organizations ◽  
Trace Model ◽  
Station Blackout ◽  
Project Prevention

In the Czech Republic, as a follow-up, a consortium of research organizations and universities has decided to simulate selected stress tests’ scenarios, in station blackout (SBO) and the loss of ultimate heat sink (LoUHS), with the aim to verify the national stress report and to analyze time response of respective source term releases. These activities are carried out in the frame of the project prevention, preparedness, and mitigation of consequences of severe accident (SA) at Czech NPPs in relation to lessons learned from stress tests after Fukushima, financed by the Ministry of Interior. The Research Centre Rez has been working on a methods for estimation of leakages and consequences of releases (MELCOR) model for VVER1000 nuclear power plant (NPP) starting with a plant systems nodalization. The aim was to benchmark the MELCOR model with the validated TRAC/RELAP advanced computational engine (TRACE) model, first comparing the steady state and continuing in a long-term SBO plus another event until the beginning of the SA. The presented work is based on the previous paper from the ICONE 23rd Conference hosted in Japan. It focuses mainly on the comparison of the thermohydraulics of the two models created in MELCOR and TRACE codes as outcome of the “Fukushima project.” After that, preliminary general results of the SA progression showing the hydrogen production and the relocation phenomena will be shortly discussed. This scenario is considered closed after some seconds to the break of the lower head. It is important to note that this paper is a substantial update of a previous one (Mazzini et al., 2015, “Analyses of SBO Sequence of VVER1000 Reactor Using TRACE and MELCOR Codes,” 23rd International Conference on Nuclear Engineering (ICONE23)) and although it contains the same descriptive sections, all the results are new.

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