357. 単純化PSAモデルを用いた加圧水型軽水炉機器故障を原因とする炉心損傷頻度の評価

嶋田 善夫

doi:10.3327/taesj2002.1.90

嶋田善夫

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ジャーナルフリー

2002 年 1 巻 1 号 p. 90-94

DOI https://doi.org/10.3327/taesj2002.1.90

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発行日: 2002/03/25 受付日: - J-STAGE公開日: 2009/03/16 受理日: 2001/06/28 早期公開日: - 改訂日: -
訂正情報
訂正日: 2009/03/16 訂正理由: - 訂正箇所: 論文受付日訂正内容: 訂正後 : 20010628

訂正日: 2009/03/16 訂正理由: - 訂正箇所: 引用文献情報訂正内容: 訂正前 : 7) 牛島厚二,米林賢二,成宮祥介,他,"代表的3ループPWRプラントに対するリスク情報を活用した試験の適用性評価(その2)",原子力学会「1999秋の大会」予稿集,H17, (1999).

訂正後 : 7) 牛島厚二,米林賢二,成宮祥介,他,“代表的3ループPWRプラントに対するリスク情報を活用した試験の適用性評価(その2)”,原子力学会「1999秋の大会」予稿集,H17, (1999).

訂正日: 2009/03/16 訂正理由: - 訂正箇所: PDFファイル訂正内容: -

訂正日: 2009/03/16 訂正理由: - 訂正箇所: 論文受付日訂正内容: 訂正前 : 20010628

訂正日: 2009/03/16 訂正理由: - 訂正箇所: 引用文献情報訂正内容: Wrong : 1) USNRC, “Use of probabilistic risk assessment methods in nuclear activities: Final policy statement”, Fed. Regist., 60, p. 42622 (60 FR 42622), (1995).
2) USNRC, “An approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the current licensing basis”, Regulatory Guide 1.174, (1998).
3) USNRC, “An approach for plant-specific, risk-informed decisionmaking: Inservice testing”, Regulatory Guide 1.175, (1998).
4) USNRC, “An approach for using probabilistic risk assessment in risk-informed decisionmaking: Graded quality assurance”, Regulatory Guide 1.176, (1998).
5) USNRC, “An approach for plant-specific, risk-informed decisionmaking: Technical specifications”, Regulatory Guide 1.177, (1998).
6) USNRC, “An approach for plant-specific, risk-informed decisionmaking inservice inspection of piping”, Regulatory Guide 1.178, (1998).
8) R.J. Belles, J. W Cletcher, D. A. Copinger, et al., Precursors to Potential Severe Core Damage Accidents: 1994 a Status Report, NUREG/CR-4674, Vol. 21, (1995).
9) M. B. Sattison, “SAPHIRE models and software for ASP evaluations”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1657 (1996).
10) D. M. Rasmuson, P. D. O'Reilly, “Analysis of annual accident sequence precursor occurrence rates for 1983-1994”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1645 (1996).
11) W. P. Poore III, “LER screening algorithm for identification of potential accident sequence precursor events”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1653 (1996).
12) M. B. Sattison, T. A. Thatcher, J. K. Knudsen, et al., “Lowpower and shutdown models for the accident sequence preqursor (ASP) program”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1664 (1996).
13) R. J. Budnitz, J. K. Salas, G. Apostolakis, et al., “A methodology for analyzing precursors to fire-initiated and seismicinitiated accidents”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1671 (1996).
14) C. H. Lui, W. J. Galyean, T. D. Brown, et al., “Accident sequence precursor analysis level 2/3 model development”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1679 (1996).
15) K. D. Russell, C. L. Atwood, W. J. Galyean, et al., Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), NUREG/CR 6116, Vol. 1-Vol. 10, (1994-1995).
16) Science Applications International Corporation, Daily Events Evaluation Manual (Draft Report), 1-275-03-336-01 (1992).
17) S. A. Eide, S. V. Chmielewski, T. D. Swantzet, Generic Component Failure Data Base for Light Water and Liquid Sodium Reactor PRAs, EGG-SSRE-8875, (1990).
18) J. W. Hickman, et al., PRA Procedures Guide: A Guide to the Performance of ProbabiliStic Risk Assesments for Nuclear Power Plants, NUREG/CR-2300, Vol. 1-Vol. 2, (1983).
19) D. M. Ericson, Jr., T. A. Wheeler, T. T. Sype, et al., Analysis of Core Damage Frequency: Internal Events Methodology, NUREG/CR-4550, Vol. 1, Rev. 1, (1990).
20) IAEA, Procedures for Conducting Probabilistic Safety Assessments of Nuclear Power Plants (Level 1), Safety Series No. 50-P-4, (1992).

Right : 1) USNRC, “Use of probabilistic risk assessment methods in nuclear activities: Final policy statement”, Fed. Regist., 60, p. 42622 (60 FR 42622), (1995).
2) USNRC, “An approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the current licensing basis”, Regulatory Guide 1.174, (1998).
3) USNRC, “An approach for plant-specific, risk-informed decisionmaking: Inservice testing”, Regulatory Guide 1.175, (1998).
4) USNRC, “An approach for using probabilistic risk assessment in risk-informed decisionmaking: Graded quality assurance”, Regulatory Guide 1.176, (1998).
5) USNRC, “An approach for plant-specific, risk-informed decisionmaking: Technical specifications”, Regulatory Guide 1.177, (1998).
6) USNRC, “An approach for plant-specific, risk-informed decisionmaking inservice inspection of piping”, Regulatory Guide 1.178, (1998).
8) R.J. Belles, J. W Cletcher, D. A. Copinger, et al., Precursors to Potential Severe Core Damage Accidents: 1994 a Status Report, NUREG/CR-4674, Vol. 21, (1995).
9) M. B. Sattison, “SAPHIRE models and software for ASP evaluations”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1657 (1996).
10) D. M. Rasmuson, P. D. O'Reilly, “Analysis of annual accident sequence precursor occurrence rates for 1983-1994”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1645 (1996).
11) W. P. Poore III, “LER screening algorithm for identification of potential accident sequence precursor events”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1653 (1996).
12) M. B. Sattison, T. A. Thatcher, J. K. Knudsen, et al., “Lowpower and shutdown models for the accident sequence preqursor (ASP) program”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1664 (1996).
13) R.J. Budnitz, J.K. Salas, G. Apostolakis, et al., “A methodology for analyzing precursors to fire-initiated and seismic-initiated accidents”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1671 (1996).
14) C. H. Lui, W. J. Galyean, T. D. Brown, et al., “Accident sequence precursor analysis level 2/3 model development”, Proc. PSA 96 Int. Topical Meeting on Probabilistic Safety Assessment Moving toward Risk-Based Regulation, USA, Vol. 3, p. 1679 (1996).
15) K. D. Russell, C. L. Atwood, W. J. Galyean, et al., Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), NUREG/CR 6116, Vol. 1-Vol. 10, (1994-1995).
16) Science Applications International Corporation, Daily Events Evaluation Manual (Draft Report), 1-275-03-336-01 (1992).
17) S. A. Eide, S. V. Chmielewski, T. D. Swantzet, Generic Component Failure Data Base for Light Water and Liquid Sodium Reactor PRAs, EGG-SSRE-8875, (1990).
18) J. W. Hickman, et al., PRA Procedures Guide: A Guide to the Performance of ProbabiliStic Risk Assesments for Nuclear Power Plants, NUREG/CR-2300, Vol. 1-Vol. 2, (1983).
19) D. M. Ericson, Jr., T. A. Wheeler, T. T. Sype, et al., Analysis of Core Damage Frequency: Internal Events Methodology, NUREG/CR-4550, Vol. 1, Rev. 1, (1990).
20) IAEA, Procedures for Conducting Probabilistic Safety Assessments of Nuclear Power Plants (Level 1), Safety Series No. 50-P-4, (1992).

訂正日: 2009/03/16 訂正理由: - 訂正箇所: PDFファイル訂正内容: -

抄録

It is anticipated that the modification of frequency of surveillance tests, preventive maintenance or parts replacement of safety related components might affect the component failure probability and result in the change of core damage frequency. In this study, the change of core damage frequency is assessed using a simplified PSA model capable of calculating core damage frequency in a short time period, which was developed by the USNRC to address accident sequence precursors.
As a result of the analysis, the following is clarified: (1) Core damage frequency increases sensitive, when failure probability of motor driven pumps increase from 5th percentile lower bound to 95th percentile upper bound. (2) Change of core damage frequency is little affected, when failure probability of motor operated valves, turbine driven auxiliary feed water pump and emergency diesel generators vary between 5th percentile lower bound and 95th percentile upper bound.

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