Analysis of Potential Hydrogen Risk in the PWR Containment

2006 ◽  
Author(s):  
Deng Jian ◽  
Xuewu Cao
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Management Strategies ◽  
Severe Accident ◽  
Deterministic Approach ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Containment Pressure ◽  
Accident Sequences ◽  
Containment Integrity

Various studies have shown that hydrogen combustion is one of major risk contributors to threaten the integrity of the containment in a nuclear power plant. That hydrogen risk should be considered in severe accident strategies in current and future NPPs has been emphasized in the latest policies issued by the National Nuclear Safety Administration of China (NNSA). According to a deterministic approach, three typical severe accident sequences for a PWR large dry containment, such as the large break loss-of-coolant (LLOCA), the station blackout (SBO), and the small break loss-of-coolant (SLOCA) are analyzed in this paper with MELCOR code. Hydrogen concentrations in different compartments are observed to evaluate the potential hydrogen risk. The results show that there is a great amount of hydrogen released into the containment, which causes the containment pressure to increase and some potential inconsecutive burnings. Therefore, certain hydrogen management strategies should be considered to reduce the risk to threaten the containment integrity.

The Severe Accident Analysis of CPR1000 Based on MELCOR Code

2013 ◽  
Author(s):  
Longze Li ◽  
Mingjun Wang ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Failure Time ◽  
Water Pressure ◽  
Severe Accident ◽  
Feed Water ◽  
Calculation Results ◽  
Station Blackout ◽  
Accident Response ◽  
Basic Standard

The severe accident of CPR1000 caused by station blackout with the SG safety valve failure is simulated and analyzed using MELCOR code in this work. The CPR1000 power plant severe accident response process and the results with three different assumptions, which are no the seal leakage nor the auxiliary feed water, the seal leakage and auxiliary feed water exist, the seal leakage exist but no auxiliary feed water separately, are analyzed. According to the calculation results, without the seal leakage and auxiliary feed water, pressure vessel would fail at 9576 s. When auxiliary feed water was supplied, pressure vessel’s failure time would delay nearly 30000s. When the seal leakage exists, pressure vessel’s failure time would delay about 50 s. The results are meaningful and significant for comprehending the detailed process of severe accident for CPR1000 nuclear power plant, which is the basic standard for establishing the severe accident management guideline.

Transient Analysis of Dedicated Pressure Relief Valve Under High-Pressure Condition in Severe Accident for Tianwan 56 Nuclear Power Plant

2020 ◽  
Author(s):  
Youyou Xu ◽  
Jian Deng ◽  
Xiaoji Wang ◽  
Lingjun Wu ◽  
Ming Zhang ◽  
...  
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Feed Water ◽  
Pressure Condition ◽  
Relief Valve ◽  
Pressure Relief ◽  
Accident Sequences ◽  
Water Accident

Abstract In the management of severe accident of nuclear reactor, the pressure relief of reactor coolant system (RCS) is an important mitigation measure to prevent high pressure core melt (HPCM). In the safety system improvement of Tianwan56 nuclear power plant, the optimization measure of adding the dedicated pressure relief valve (DPRV) for severe accident were adopted. This improvement allows the reactor to release the pressure of RCS before the reactor vessel being damaged to mitigate the consequence of reactor melt accident under high-pressure condition. Based on the analysis of severe accident sequences, the total loss of feed water accident is confirmed to cover the various severe accident consequences which may lead to HPCM accident. This paper studied the transient characteristics of total loss of feed water accident sequences, and the factors such as valve opening delay on the operating temperature of the valve were researched. Finally, the representative and envelope operating condition of DPRV under severe accident was clarified. Besides, the temperature curve of fluid passing through the valve and the maximum temperature the valve experienced were obtained. This research provides the valuable and indispensable basis to the operability and integrity analysis of DPRV in severe accident.

Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

2006 ◽  
Author(s):  
Gert Sdouz
Keyword(s):  
Source Term ◽  
Management Strategies ◽  
Severe Accidents ◽  
Loss Of Coolant Accident ◽  
Management Measures ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Station Blackout ◽  
Accident Sequences ◽  
Comprehensive Study

The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the untightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the “Station Blackout”-sequence and the “Large Break LOCA”. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was demonstrated that the accident management measures have quite lower consequences. In addition it was shown that in the case of a “Large Break LOCA”-sequence the intact containment retains the nuclides up to a factor of 20 000. This is much more than in the case of a “Station Blackout”-sequence. Within the frame of the study 17 source terms have been generated to evaluate in detail accident management strategies for VVER-1000 reactors.

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.

The Development of a Mechanistic Model of MCCI Under Severe Accidents

2013 ◽  
Author(s):  
Wei Wei ◽  
Kelin Qi ◽  
Fuchang Shan ◽  
Yanfang Chen ◽  
Fude Guo
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Mechanistic Model ◽  
Severe Accident ◽  
Severe Accidents ◽  
Large Scale Analysis ◽  
Loss Of Coolant ◽  
Calculation Results ◽  
Station Blackout ◽  
Core Concrete

This paper describes a mechanistic model of the molten core-concrete interaction (MCCI) process under severe accidents, and selects the Daya-Bay nuclear power plant as the research object to calculate and analyze the process of the MCCI when the station blackout (SBO), or loss of coolant (LOCA) severe accident serial is happened. The calculation results of this procedure are compared with the large-scale analysis programs MELCOR to verify the reasonableness and correctness of the model. The results indicate that the model present in this paper can simulate the MCCI process correctly and reasonably under multi-serial severe accidents.

VVER 1000 Severe Accident Analyses Using MELCOR Code

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Marek Ruščák ◽  
Guido Mazzini ◽  
Milos Kynčl ◽  
Sevastyan Savanyuk ◽  
Miroslav Hrehor ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Time Course ◽  
Stress Test ◽  
Model Development ◽  
Severe Accident ◽  
Test Case ◽  
Station Blackout ◽  
Accident Scenarios

This work describes the computer model development of the water–water energetic reactor (VVER) 1000 nuclear power plant (NPP) in the methods for estimation of leakages and consequences of releases (MELCOR) 1.8.6 code and its subsequent use for the accident scenarios analysis leading to the core melting. The baseline accident scenario was a stress test case—the station blackout (SBO, the complete loss of alternating current electric power in a nuclear power plant). In addition to this, four other scenarios were analyzed in which the SBO was combined with other technological failures—the loss of steam generator feedwater system and small, medium, and large break coolant accidents (LOCA). The results provided detailed information on the time course of accident scenarios, their temperature and pressure parameters, hydrogen production, and the mass inventory released from the molten corium and debris into the containment of the NPP.

scholarly journals Study on Airborne Radionuclide Dispersion in Floating Nuclear Power Plant under the Loss-of-Coolant Accident

2021 ◽  
Vol 2021 ◽  
pp. 1-8
Author(s):  
Shuliang Zou ◽  
Na Liu ◽  
Binhai Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Effective Dose ◽  
Nuclear Power ◽  
Severe Accident ◽  
Emergency Plan ◽  
Dose Limit ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Dynamics Method

Floating nuclear power plant is a kind of nuclear power plant on a barge moored specifically in an area of the sea. In order to study the factors influencing airborne radionuclide dispersion induced by the loss-of-coolant accident in floating nuclear power plant, the floating nuclear power plant platform was taken as the research object, and the dispersion of airborne radionuclide under combined conditions of platform positions, wind directions, and break directions (north, south, west, and east) was simulated by the CFD (computational fluid dynamics) method. The results show that northern and southern breaks have less dangerous island area than western and eastern ones but have more platform dangerous area than the western and eastern ones. The risk of the southern break is the greatest, and that of the western break is the least. Rotating the floating nuclear power plant platform in a certain angle can reduce the damage of loss-of-coolant accident. The effects of the dose received by the personnel under the condition of the severe accident were evaluated based on previous research, showing that the inhalation effective dose and the effective dose of plume immersion exposure were less than the radiation dose limit of 0.25 Sv within two hours in the accident. The results of the study can provide reference for the design of floating nuclear power plant platform and the formulation of emergency plan.

scholarly journals Analysis of Processes in the Containment Using ATHLET-CD and COCOSYS Codes

2020 ◽  
pp. 27-37
Author(s):  
M. Vyshemirskyi ◽  
V. Pustovit ◽  
V. Kravchenko ◽  
D. Donskyi
Keyword(s):  
Nuclear Power ◽  
Fission Products ◽  
Severe Accident ◽  
Primary System ◽  
Coupled Calculation ◽  
Code Model ◽  
Loss Of Coolant Accident ◽  
Calculation Results ◽  
Station Blackout ◽  
Containment Pressure

A brief description of performed input deck modifications and results of stand-alone and coupled calculations of Dn 200 mm loss of coolant accident with simultaneous total station blackout accident scenario for Rivne Nuclear Power Plant Unit 1 (WWER‑440/V-213) with application of ATHLET-CD 3.1A and COCOSYS 2.4 codes are presented in the paper. ATHLET-CD stand-alone calculation was performed with constant containment pressure (a time dependent volume with constant pressure and temperature was used as a boundary volume for leakage). Further, mass and energy release and fission products from the primary system obtained during ATHLET‑CD stand-alone calculation were used to perform COCOSYS stand-alone calculation. In addition, coupled ATHLET-CD and COCOSYS calculation was performed. All the computer analyzes were performed until the lower head failure. ATHLET‑CD model was extended with core degradation module (ECORE), which allowed calculation of scenario until reactor pressure vessel failure. According to the results of comparative analysis, nearly the same behavior of the main parameters in the stand-alone and coupled calculation at an early phase of scenario was obtained. Some small differences occur due to distinction in behavior of water and steam mass flows released through the break and due to existence of heat transfer from the primary system structures to the containment compartments during coupled calculation of transient. As for middle and late phases of the accident, some differences between stand-alone and coupled calculation results for analyzed scenario are present. These differences are caused by different total fission products and aerosols release from the reactor coolant system to the containment compartments. The above information allows recommending application of coupled code/model versions for performing the computer severe accident analyses.

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