Severe Accident Steady-State Modeling of Advanced Nuclear Power Plant Based on ASTEC Code

2017 ◽  
Author(s):  
Wei Song ◽  
Jiaxu Zuo ◽  
Yan Chen ◽  
Chaojun Li ◽  
Peng Zheng
Keyword(s):  
Steady State ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Cooling Systems ◽  
Accident Management ◽  
Management Guidance ◽  
Level 2 ◽  
Reactor Block

Severe accident is an attractive topic today for the nuclear power plant (NPP) safety. In the nuclear safety regulatory work, it is planned to build a full scale severe accident model for the advanced nuclear power plant of China to study the new designs of severe accident prevention and mitigation systems and strategies, and to further deploy the application on the level 2 PSA and severe accident management guidance. This paper firstly introduces the modeling tool, ASTEC, and then presents the progress of modeling work, which is mainly on the steady state modeling and regulation including reactor block, primary and secondary cooling systems, regulation systems etc. Last but not least, the work plan for the future is given.

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.

Design of Severe Accident Management Systems for Beyond Design Basis External Hazards at Paks NPP

2016 ◽  
Author(s):  
Tamás János Katona ◽  
András Vilimi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Management Systems ◽  
Basic Question ◽  
Severe Accidents ◽  
Design Basis ◽  
Accident Management ◽  
External Events

Paks Nuclear Power Plant identified the post-Fukushima actions for mitigation and management of severe accidents caused by external events that include updating of some hazard assessments, evaluation of capacity / margins of existing severe accident management facilities, and construction of some mew systems and facilities. In all cases, the basic question was, what level of margin has to be ensured above design basis external hazard effects, and what level of or hazard has to be taken for the design. Paks Nuclear Power Plant developed certain an applicable in the practice concept for the qualification of already implemented and design the new post-Fukushima measures that is outlined in the paper. The concept and practice is presented on several examples.

Analysis of Severe Accident Management Strategy for a BWR-4 Nuclear Power Plant

2005 ◽  
Vol 152 (3) ◽  
pp. 253-265 ◽  
Author(s):  
Te-Chuan Wang ◽  
Shih-Jen Wang ◽  
Jyh-Tong Teng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Management Strategy ◽  
Nuclear Power ◽  
Severe Accident ◽  
Accident Management

Design of Severe Accident Management Systems for Beyond Design Basis External Hazards at Paks Nuclear Power Plant

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Tamás János Katona ◽  
András Vilimi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Exceedance Probability ◽  
Basic Question ◽  
Existing Structures ◽  
Design Basis ◽  
Accident Management ◽  
Definition Of

Paks Nuclear Power Plant (NPP) identified the post-Fukushima actions for mitigation and management of severe accidents caused by external events that include updating of some hazard assessments, evaluation of margins of existing severe accident management (SAM) facilities, and construction of some new systems and facilities. While developing the SAM strategy, the basic question was what is the sufficient margin above the design basis level of existing structures, systems, and components for avoiding the cliff-edge effects, and what level of or hazard should be taken for the design of new structures and systems dedicated for SAM. Paks NPP developed an applicable in the practice concept for the qualification of already implemented SAM measures and design the new post-Fukushima measures that are outlined in the paper. The concept is based on the generalization of the procedure and assumptions used in the definition of acceptable margins for seismic loads, analysis of the steepness of the hazard curves and features of the hazards. Justification of the definition of exceedance probability of the design basis effects for the design of SAM systems is given based on the first order reliability theory. The application of the concept is presented on several practical examples.

Characteristics of Hydrogen Monitoring Systems for Severe Accident Management at a Nuclear Power Plant

2018 ◽  
Vol 65 (2) ◽  
pp. 79-87 ◽  
Author(s):  
V. G. Petrosyan ◽  
E. A. Yeghoyan ◽  
A. D. Grigoryan ◽  
A. P. Petrosyan ◽  
M. R. Movsisyan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Monitoring Systems ◽  
Accident Management ◽  
Hydrogen Monitoring
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