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

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.

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

Journal of Power and Energy Systems ◽

10.1299/jpes.4.84 ◽

2010 ◽

Vol 4 (1) ◽

pp. 84-93

Author(s):

Masutake SOTSU ◽

Kenichi KURISAKA

Keyword(s):

Probabilistic Method ◽

Core Damage ◽

Damage Risk

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Discussion on Application of Accident Sequence Precursor Analysis in China

Volume 4: Nuclear Safety, Security, Non-Proliferation and Cyber Security; Risk Management ◽

10.1115/icone25-66579 ◽

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.

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A PSA Case Study: Promoting the Reliability of the CRSS on the CMRR Through the ATWS Mitigation System

Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management ◽

10.1115/icone26-81755 ◽

2018 ◽

Author(s):

Heng Yu ◽

Guan-bo Wang ◽

Da-zhi Qian ◽

Yu-chuan Guo ◽

Bo Hu

Keyword(s):

Research Reactor ◽

Control Rod ◽

Research Reactors ◽

The Core ◽

Core Damage ◽

The World ◽

Safety Features ◽

Control Rods ◽

Pool Type

An increasing number of PSA programs concerning research reactors have been launched across the world. As with many other reactors, the CMRR (China Mianyang Research Reactor), a typical pool-type research reactor, regards the control rod shutdown system (CRSS) as its primary shutdown system which enables the reactor subcritical by dropping control rods into the core after a specific initiating event is detected. As a result, the CRSS is an essential ingredient of engineered safety features. It is necessary to enhance the reliability of the CRSS, ensuring the reactor can be successfully shut down when the ATWS — the anticipated transients without scram occurs. Therefore, additional facilities should be designed to cope with the extremely severe circumstance. Accordingly, the purpose of this paper is to evaluate the promotion of the CMRR’s safety degree and the reliability of its CRSS from the PSA’s perspective with an ATWS mitigation system installed. Results indicate that, by introducing the ATWS mitigation system, the failure probability of the CRSS can decrease from 1.52e−05 per demand to 3.35e−06 per demand, while the aggregate CDF (core damage frequency) induced by all IE (initiating event) groups, is able to decrease to a relatively low value 1.17e−05/y from its previous value 3.11e−06/y. It is apparent that the reliability of the CRSS as well as the safety degree of the overall reactor can be enhanced effectively by adding the ATWS mitigation system to the elementary design of the normal CRSS.

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

Applied Mechanics and Materials ◽

10.4028/www.scientific.net/amm.341-342.1338 ◽

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

Annals of Nuclear Energy ◽

10.1016/j.anucene.2021.108662 ◽

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

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Damage Risk: An Evaluation of the Effects of Exposure to 85 versus 90 dBA of Noise

Journal of Speech and Hearing Research ◽

10.1044/jshr.1902.216 ◽

1976 ◽

Vol 19 (2) ◽

pp. 216-224 ◽

Cited By ~ 3

Author(s):

James T. Yates ◽

Jerry D. Ramsey ◽

Jay W. Holland

Keyword(s):

White Noise ◽

Statistical Difference ◽

Noise Exposure ◽

Threshold Shift ◽

Temporary Threshold Shift ◽

Noise Levels ◽

Potential Damage ◽

Full Day ◽

Damage Risk

The purpose of this study was to compare the damage risk of 85 and 90 dBA of white noise for equivalent full-day exposures. The damage risk of the two noise levels was determined by comparing the temporary threshold shift (TTS) of 12 subjects exposed to either 85 or 90 dBA of white noise for equivalent half- and full-day exposures. TTS was determined by comparing the pre- and postexposure binaural audiograms of each subject at 1, 2, 3, 4, 6, and 8 kHz. It was concluded that the potential damage risk, that is, hazardous effect, of 90 dBA is greater than 85 dBA of noise for equivalent full-day exposures. The statistical difference between the overall effects of equivalent exposures to 85 dBA as compared to 90 dBA of noise could not be traced to any one frequency. The damage risk of a full-day exposure to 85 dBA is equivalent to that of a half-day exposure to 90 dBA of noise. Within the limits of this study, TTS t was as effective as TTS 2 for estimating the damage risk of noise exposure.

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