Transformation of classical PSA and DSA into the form of conditional event tree: An approach of human action in time dependent core damage risk

2022 ◽  
Vol 165 ◽  
pp. 108662
Author(s):  
Alireza Najafi ◽  
Athena Shahsavand ◽  
Seyed Ali Hosseini ◽  
Amir Saeed Shirani ◽  
Faramarz Yousefpour ◽  
...  
Keyword(s):  
Human Action ◽  
Time Dependent ◽  
Event Tree ◽  
Conditional Event ◽  
Core Damage ◽  
Damage Risk

Evaluation of Monju Core Damage Risk With Change of AOT Using Probabilistic Method

2009 ◽  
Author(s):  
Masutake Sotsu ◽  
Kenichi Kurisaka
Keyword(s):  
Probabilistic Method ◽  
Fast Breeder Reactor ◽  
Core Damage ◽  
Safety Regulations ◽  
Cooling Loops ◽  
Damage Risk

MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops which can supply 280 MW of electricity. Limiting conditions of operation (LCO) defined in the safety regulations in MONJU given the allowed outage time (AOT) are evaluated using a PSA technique. The result indicates the possibility of limit extension and some prospects that we should examine.

scholarly journals Evaluation of MONJU Core Damage Risk with Change of AOT Using Probabilistic Method

2010 ◽  
Vol 4 (1) ◽  
pp. 84-93
Author(s):  
Masutake SOTSU ◽  
Kenichi KURISAKA
Keyword(s):  
Probabilistic Method ◽  
Core Damage ◽  
Damage Risk

scholarly journals Evaluation of MONJU Core Damage Risk due to Control Rod Function Failure

2012 ◽  
Vol 6 (3) ◽  
pp. 462-471
Author(s):  
Masutake SOTSU ◽  
Kenichi KURISAKA
Keyword(s):  
Control Rod ◽  
Core Damage ◽  
Rod Function ◽  
Damage Risk

Discussion on Application of Accident Sequence Precursor Analysis in China

2017 ◽  
Author(s):  
Jiang Hu ◽  
Guoxu Zhang ◽  
Yushu Zhang ◽  
Binbin Zhang ◽  
Zhi Xiao
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Operating Experience ◽  
Evaluation Process ◽  
Management Platform ◽  
Core Damage ◽  
Current State ◽  
Core Cooling ◽  
Damage Risk ◽  
Degraded Condition

An accident sequence precursor (ASP) is an initiating event or degraded condition that, when coupled with one or more postulated events, could result in inadequate core cooling or even severe reactor core damage. Through the systematical investigation and evaluation of nuclear power plant (NPP) operating experience, the ASP analysis could provide a comprehensive, risk-informed view of NPP operating experience, a measure for trending core damage risk, a partial validation of the current state of practice in risk assessment, and also a feedback to regulatory activities. This paper firstly gives a brief review on the ASP evaluation process of U.S. and German, and then introduces the current progresses on the establishment of ASP regulatory framework and the development of ASP management platform, which are carrying out by the related nuclear regulation agency in China. Also, this paper gives some insightful discussions on issues such as the ASP application and problems maybe faced.

Development of Probabilistic Risk Assessment Methodology of Decay Heat Removal Function Against Combination Hazard of Low Temperature and Snow for Sodium-Cooled Fast Reactors

2017 ◽  
Author(s):  
Hiroyuki Nishino ◽  
Hidemasa Yamano ◽  
Kenichi Kurisaka
Keyword(s):  
Risk Assessment ◽  
Low Temperature ◽  
Heat Removal ◽  
Ambient Air ◽  
Probabilistic Risk Assessment ◽  
Event Tree ◽  
Snow Removal ◽  
Decay Heat ◽  
Core Damage ◽  
Probabilistic Risk

A Probabilistic Risk Assessment (PRA) should be performed not only for earthquake and tsunami which are major natural events in Japan, but also for other natural external hazards. However, PRA methodologies for other external hazards and their combination have not been sufficiently developed. This study is intended to develop PRA methodology for a combination of low temperature and snow for a Sodium-cooled Fast Reactor (SFR) that uses the ambient air as its ultimate heat sink for decay heat removal under accident conditions. Annual excess probabilities of low temperature and of snow are statistically estimated based on the meteorological records of low temperature, snow depth and daily snowfall depth. To identify core damage sequence, an event tree was developed by considering the impact of low temperature and snow on decay heat removal systems (DHRSs), e.g., plugged intake and/or outtake for the DHRS and for the emergency diesel generator (EDG), unopenable door on the access routes due to accumulated snow, failure of the intake filters due to accumulated snow, possibility of freezing of the water in cooling circuits. Recovery actions (i.e., snow removal and filter replacement) to prevent loss of DHRS function were also considered in developing the event tree. Furthermore, considering that a dominant contributor to snow risk can be failure of snow removal around the intake and outtake induced by loss of the access routes, this study has investigated effects of electric heaters installed around the intake and outtake as an additional countermeasure. By using the annual excess probabilities and failure probabilities, the event tree was quantified. The result showed that a dominant core damage sequence is failure of the electric heaters and loss of the access routes for snow removal against the combination hazard at daily snowfall depth of 2 m/day, duration time (snow and low temperature) of 1 day.

scholarly journals Research on Time-Dependent Failure Modeling Method of Integrating Discrete Dynamic Event Tree With Fault Tree

2019 ◽  
Vol 7 ◽  
Author(s):  
Anqi Xu ◽  
Zhijian Zhang ◽  
Min Zhang ◽  
He Wang ◽  
Huazhi Zhang ◽  
...  
Keyword(s):  
Fault Tree ◽  
Modeling Method ◽  
Time Dependent ◽  
Event Tree ◽  
Discrete Dynamic ◽  
Dynamic Event ◽  
Failure Modeling ◽  
Dependent Failure

Sensitivity Analyses of Steam Generator Tube Thickness on Induced-SGTR in SBO Accident

2013 ◽  
Vol 341-342 ◽  
pp. 1338-1341
Author(s):  
Jin Rong Qiu ◽  
Chang Hong Peng ◽  
Yun Guo
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Fission Products ◽  
Sensitivity Analyses ◽  
Steam Generator Tube ◽  
Core Damage ◽  
Reference Plant ◽  
Station Blackout ◽  
Level I ◽  
Damage Risk

For nuclear power plant, station black-out (SBO) is the events that contribute significantly to the level-I core damage risk. For an SBO, it is assumed that both the off-site power and on-site diesel generators fail to supply alternating-current power for the plant systems. SBO induced steam generator tube rupture (SGTR) is a concern because the steam generator (SG) tubes are parts of the reactor coolant pressure boundary and failure of the SG tubes may lead to fission products bypassing the containment. The SG tube integrity may be challenged by high temperature and high pressure conditions and may have a potential to fail due to creep rupture. This study focuses on the probability of SBO induced SGTR accidents under the station blackout (SBO) with RCS integrity, seal LOCA and steam relief valves remaining stuck open for the reference plant. At last, the sensitivity of the tube thick is studied.

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