Analysis of surveillance test interval by Markov process for SDS1 in CANDU nuclear power plants

Cho, Sung-Whan; Jiang, Jin

doi:10.1016/j.ress.2006.10.007

“…This is known as an SBO accident. Two independent fast-acting shutdown systems, namely shutoff rods (SDS1) and poison injection system (SDS2), will be activated automatically to safely shut down the reactor (Cho and Jiang 2008). Fig.…”

Section: System Failure C1f C2fmentioning

confidence: 99%

Comparison of Resilience vs Traditional Probabilistic Safety Assessment for Nuclear Power Facilities

Yan,

Dunnett,

rews

2022

Book of Extended Abstracts for the 32nd European Safety and Reliability Conference

0

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The safety of nuclear power facilities has consistently been of concern to the public as well as the scientific community due to the potentially catastrophic consequences of nuclear accidents. In this paper, a Petri net modelling-based resilience methodology developed by the authors is compared with the traditional Probabilistic Safety Assessment (PSA) methodologies namely Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) used for nuclear engineering risk assessment. A station blackout (SBO) accident is used as a case study to facilitate the comparison. The advantages and drawbacks of the two methods are compared and conclusions are drawn as to the best of these two options for future exploitation in ensuring the safety of nuclear reactors.

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“…This is known as an SBO accident. Two independent fast-acting shutdown systems, namely shutoff rods (SDS1) and poison injection system (SDS2), will be activated automatically to safely shut down the reactor (Cho and Jiang 2008). Fig.…”

Section: System Failure C1f C2fmentioning

confidence: 99%

Comparison of Resilience vs Traditional Probabilistic Safety Assessment for Nuclear Power Facilities

Yan,

Dunnett,

rews

2022

Book of Extended Abstracts for the 32nd European Safety and Reliability Conference

0

View full text Add to dashboard Cite

The safety of nuclear power facilities has consistently been of concern to the public as well as the scientific community due to the potentially catastrophic consequences of nuclear accidents. In this paper, a Petri net modelling-based resilience methodology developed by the authors is compared with the traditional Probabilistic Safety Assessment (PSA) methodologies namely Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) used for nuclear engineering risk assessment. A station blackout (SBO) accident is used as a case study to facilitate the comparison. The advantages and drawbacks of the two methods are compared and conclusions are drawn as to the best of these two options for future exploitation in ensuring the safety of nuclear reactors.

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“…Reference [22] presents the analysis of surveillance test interval by Markov processes for shutdown systems in CANDU nuclear power plants. In order to comply with regulatory requirements, the system availability is evaluated taking into account component failure rate data and the benefits of the tests.…”

Section: Introductionmentioning

confidence: 99%

Risk-Based Allowed Outage Time and Surveillance Test Interval Extensions for Angra 1

Gibelli

¹

,

Melo

²

,

Leite

³

2012

International Journal of Quality, Statistics, and Reliability

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In this work, Probabilistic Safety Assessment (PSA) is used to evaluate Allowed Outage Times (AOT) and Surveillance Test Intervals (STI) extensions for three Angra 1 nuclear power plant safety systems. The interest in such an analysis lies on the fact that PSA comprises a risk-based tool for safety evaluation and has been increasingly applied to support both the regulatory and the operational decision-making processes. Regarding Angra 1, among other applications, PSA is meant to be an additional method that can be used by the utility to justify Technical Specification relaxation to the Brazilian regulatory body. The risk measure used in this work is the Core Damage Frequency, obtained from the Angra 1 Level 1 PSA study. AOT and STI extensions are evaluated for the Safety Injection, Service Water and Auxiliary Feedwater Systems using the SAPHIRE code. In order to compensate for the risk increase caused by the extensions, compensatory measures as (1) test of redundant train prior to entering maintenance and (2) staggered test strategy are proposed. Results have shown that the proposed AOT extensions are acceptable for two of the systems with the implementation of compensatory measures whereas STI extensions are acceptable for all three systems.

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Analysis of surveillance test interval by Markov process for SDS1 in CANDU nuclear power plants

Cited by 18 publications

References 15 publications

Comparison of Resilience vs Traditional Probabilistic Safety Assessment for Nuclear Power Facilities

Comparison of Resilience vs Traditional Probabilistic Safety Assessment for Nuclear Power Facilities

Comparison of Resilience vs Traditional Probabilistic Safety Assessment for Nuclear Power Facilities

Risk-Based Allowed Outage Time and Surveillance Test Interval Extensions for Angra 1

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