The Development and Application of PWR Severe Accident Simulation Model

2013 ◽  
Author(s):  
Yanfang Chen ◽  
Zhengquan Xie ◽  
Xusheng Lin ◽  
Fuchang Shan ◽  
Wei Wei ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Mathematical Models ◽  
Simulation Model ◽  
Nuclear Power ◽  
Severe Accident ◽  
Plant Type ◽  
Simulation Code ◽  
Accident Simulation ◽  
Simulation Codes

The severe accident simulation codes that developed by the engineers in RINPO are introduced in this chapter. The results of the severe accident caused by large LOCA plus losing safety core injection are presented. Comparison with the results of SCDAP/RELAP5/MOD3.2 of the same accident and the same nuclear power plant type has been made. From the comparison and the analysis we can make the conclusion that the trend-lines are correct and the mathematical models are reasonable in this simulation code.

Development and Verification of Severe Accident Simulation Capability for a NPP Simulator

2013 ◽  
Author(s):  
Zhifei Yang ◽  
Xiaofei Xie ◽  
Xing Chen ◽  
Shishun Zhang ◽  
Yehong Liao ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation System ◽  
Severe Accident ◽  
Normal Operation ◽  
Feed Water ◽  
Accident Simulation ◽  
Accident Conditions ◽  
Relap5 Code

It is reflected in the severe accident in Fukushima Daiichi that the emergency capacity of nuclear power plant needs to be enhanced. A nuclear plant simulator that can model the severe accident is the most effective means to promote this capacity. Until now, there is not a simulator which can model the severe accident in China. In order to enhance the emergency capacity in China, we focus on developing a full scope simulator that can model the severe accident and verify it in this study. The development of severe accident simulation system mainly includes three steps. Firstly, the integral severe accident code MELCOR is transplanted to the simulation platform. Secondly, the interface program must be developed to switch calculating code from RELAP5 code to MELCOR code automatically when meeting the severe accident conditions because the RELAP5 code can only simulate the nuclear power plant normal operation state and design basis accident but the severe accident. So RELAP5 code will be stopped when severe accident conditions happen and the current nuclear power plant state parameters of it should be transported to MELCOR code, and MELCOR code will run. Finally, the CPR1000 nuclear power plant MELCOR model is developed to analyze the nuclear power plant behavior in severe accident. After the three steps, the severe accident simulation system is tested by a scenario that is initiated by the station black out with reactor cooling pump seal leakage, HHSI, LHSI and auxiliary feed water system do not work. The simulation result is verified by qualitative analysis and comparison with the results in severe accident analysis report of the same NPP. More severe accident scenarios initiated by LBLOCA, MBLOCA, SBLOCA, SBO, ATWS, SGTR, MSLB will be tested in the future. The results show that the severe accident simulation system can model the severe accident correctly; it meets the demand of emergency capacity promotion.

Analysis of Inertization Strategies for the Filtered Containment Venting System in Cofrentes Nuclear Power Plant

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Kevin Fernández-Cosials ◽  
Gonzalo Jiménez ◽  
César Serrano ◽  
Luisa Ibáñez ◽  
Ángel Peinado
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Computational Model ◽  
Nuclear Power ◽  
Inert Atmosphere ◽  
Severe Accident ◽  
Nitrogen Injection

During a severe accident (SA) in a nuclear power plant (NPP), there are several challenges that need to be faced. To coup with a containment overpressure, the venting action will lower the pressure but it will release radioactivity to the environment. In order to reduce the radioactivity released, a filtered containment venting system (FCVS) can be used to retain iodine and aerosols radioactive releases coming from the containment atmosphere. However, during a SA, large quantities of hydrogen can also be generated. Hydrogen reacts violently with oxygen and its combustion could impair systems, components, or structures. For this reason, to protect the integrity of the FCVS against hydrogen explosions, an inertization system is found necessary. This system should create an inert atmosphere previous to any containment venting that impedes the contact of hydrogen and oxygen. In this paper, the inertization system for Cofrentes NPP is presented. It consists of a nitrogen injection located in three different points. A computational model of the FCVS as well as the inertization system has been created. The results show that if the nitrogen sweeps and the containment venting are properly synchronized, the hydrogen risk could be reduced to a minimum and therefore, the integrity of the FCVS would be preserved.

Estimation of Radiological Consequences at a Proposed Nuclear Power Plant Site Using Atmospheric Dispersion Code

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Kwame Gyamfi ◽  
Sylvester Attakorah Birikorang ◽  
Emmanuel Ampomah-Amoako ◽  
John Justice Fletcher
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Atmospheric Dispersion ◽  
Atmospheric Stability ◽  
Assessment System ◽  
Severe Accident ◽  
Dispersion Modeling ◽  
Stability Class ◽  
Total Effective Dose Equivalent

Abstract Atmospheric dispersion modeling and radiation dose calculation have been performed for a generic 1000 MW water-water energy reactor (VVER-1000) assuming a hypothetical loss of coolant accident (LOCA). Atmospheric dispersion code, International Radiological Assessment System (InterRAS), was employed to estimate the radiological consequences of a severe accident at a proposed nuclear power plant (NPP) site. The total effective dose equivalent (TEDE) and the ground deposition were calculated for various atmospheric stability classes, A to F, with the site-specific averaged meteorological conditions. From the analysis, 3.7×10−1 Sv was estimated as the maximum TEDE corresponding to a downwind distance of 0.1 km within the dominating atmospheric stability class (class A) of the proposed site. The intervention distance for evacuation (50 mSv) and sheltering (10 mSv) were estimated for different stability classes at different distances. The intervention area for evacuation ended at 0.5 km and that for sheltering at 1.5 km. The results from the study show that designated area for public occupancy will not be affected since the estimated doses were below the annual regulatory limits of 1 mSv.

Onset of Gas Blowthrough During Melt Expulsion From a Hole in a RPV: Numerical Analysis Using the SIMMER Code

2006 ◽  
Author(s):  
Frank Kretzschmar
Keyword(s):  
Numerical Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Material ◽  
Test Facility ◽  
Computer Codes ◽  
Direct Containment Heating ◽  
Heating Up

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

Application of Level 2 PSA in the Design of Cavity Injection System for Nuclear Power Plant

2018 ◽  
Author(s):  
Wentao Zhu ◽  
Wenjing Li
Keyword(s):  
Decision Making ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Assessment Tool ◽  
Quantitative Risk Assessment ◽  
Severe Accident ◽  
Injection System ◽  
Retention Strategy ◽  
Level 2

After Fukushima nuclear power plant accident, severe accident is getting more and more concerns all over the world. In order to mitigate severe accident and improve the safety of nuclear power plant, two different strategies are applied in different plants. One is in-vessel melt retention strategy, and the other is ex-vessel melt retention strategy. Tianwan nuclear power plant is an improved Gen II nuclear power plant and in-vessel melt retention strategy is adopted in the plant. In order to achieve this strategy, cavity injection system is designed for the plant. Probabilistic Safety Analysis is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. For this plant, in order to optimize the design of cavity injection system, improve the safety level of nuclear power plant, and meanwhile, improve the engineering implementation and economization, Level 2 PSA was used for this decision-making process. In this paper, the Level 2 PSA for this plant and the application for the design of cavity injection system are introduced.

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