Review and categorization of existing studies on the estimation of probabilistic failure metrics for Reactor Coolant Pressure Boundary piping and steam generator tubes in Nuclear Power Plants

Cheng, Wen-Chi; Sakurahara, Tatsuya; Zhang, Sai; Farshadmanesh, Pegah; Reihani, Seyed; Kee, Ernie; Mohaghegh, Zahra; Heckmann, Klaus; Sievers, Jürgen; Lydell, Bengt; Zammali, Chokri; Yuan, Xian-Xun; Duan, Xinjian; Alzbutas, Robertas; Lee, Gyeong-Geun; Karim, Julia Abdul; Morozov, Vladimir; Takasugi, Cole; Jevremovic, Tatjana

doi:10.1016/j.pnucene.2019.103105

Title: Review and categorization of existing studies on the estimation of probabilistic failure metrics for Reactor Coolant Pressure Boundary piping and steam generator tubes in Nuclear Power Plants

Journal Article · Sat Aug 24 00:00:00 EDT 2019 · Progress in Nuclear Energy

DOI:https://doi.org/10.1016/j.pnucene.2019.103105· OSTI ID:1848156

Cheng, Wen-Chi ^[1]; Sakurahara, Tatsuya ^[1];

^[1]; Farshadmanesh, Pegah ^[1]; Reihani, Seyed ^[1]; Kee, Ernie ^[1];

^[1]; Heckmann, Klaus ^[2];

^[2]; Lydell, Bengt ^[3]; Zammali, Chokri ^[4]; Yuan, Xian-Xun ^[5]; Duan, Xinjian ^[6]; Alzbutas, Robertas ^[7]; Lee, Gyeong-Geun ^[8]; Karim, Julia Abdul ^[9]; Morozov, Vladimir ^[10]; Takasugi, Cole ^[11]; Jevremovic, Tatjana ^[11]

Univ. of Illinois at Urbana-Champaign, IL (United States). Dept. of Nuclear, Plasma, and Radiological Engineering
Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany). Containment Dept. Safety Research Division
SIGMA-PHASE, INC., Vail, AZ (United States)
Tunisian Electricity and Gas Company (Tunisia). Alternative Energies Project
Ryerson Univ., Toronto (Canada). Dept. of Civil Engineering
Candu Energy Inc., Mississauga (Canada)
Lithuanian Energy Inst., Kaunas (Lithuania). Lab. of Nuclear Installation Safety
Korean Atomic Energy Research Inst. (KAERI), Daejeon (Korea, Republic of)
Malaysian Nuclear Agency, Bangi (Malaysia)
Russian Federation
International Atomic Energy Agency (IAEA), Vienna (Austria)

This article presents the first output of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP), titled Methodology for Assessing Pipe Failure Rates in Advanced Water-Cooled Reactors (AWCRs), launched in 2018. This IAEA CRP is aimed at developing a good-practices document for estimating the probabilistic failure metrics of piping in AWCRs. The reliability of piping that comprise the Reactor Coolant Pressure Boundary (RCPB) is important for maintaining safe and stable operations of Nuclear Power Plants (NPPs) because failure of those piping components could lead to undesirable consequences, such as plant shutdown, costly repair, the occurrence of Loss-of-Coolant Accidents (LOCAs) and, possibly, subsequent core damage or large release events. Probabilistic failure metrics (e.g., failure rate, failure frequency, or failure probability) of RCPB components are the key inputs to the Probabilistic Safety/Risk Assessment (PSA/PRA) and risk management of NPPs. The estimation of probabilistic failure metrics, however, is challenging, especially for AWCRs, due to the lack of operating experience. Therefore, as the first step of the IAEA CRP activities, this paper is developed to provide a literature review of the existing studies (from 2000 to April 2019) on the estimation of probabilistic failure metrics for RCPB piping and Steam Generator (SG) tubes of NPPs and to categorize them based on four criteria: (1) explicitness of incorporation of physical failure mechanisms; (2) types of failure characterization; (3) types of physical models for degradation; and (4) explicitness of consideration of maintenance. The existing studies are also analyzed from the viewpoint of the following key aspects: (i) uncertainty analysis, (ii) sensitivity analysis, (iii) validation strategies, and (iv) the areas of applications.

View Accepted Manuscript (DOE)

View Accepted Manuscript (Publisher)

Cite

Export

Save

Research Organization:: Univ. of Illinois at Urbana-Champaign, IL (United States)

Sponsoring Organization:: USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: NE0008700; 17–12614; I31030

OSTI ID:: 1848156

Alternate ID(s):: OSTI ID: 1682451

Journal Information:: Progress in Nuclear Energy, Vol. 118, Issue C; ISSN 0149-1970

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

References (102)

PFM APPLICATION FOR THE PWSCC INTEGRITY OF Ni-BASE ALLOY WELDS-DEVELOPMENT AND APPLICATION OF PINEP-PWSCC Hong, Jong-Dae; Jang, Changheui; Kim, Tae Soon Nuclear Engineering and Technology, Vol. 44, Issue 8 https://doi.org/10.5516/NET.07.2012.017	journal	December 2012
A Bayesian approach to modeling and predicting pitting flaws in steam generator tubes Yuan, X. -X.; Mao, D.; Pandey, M. D. Reliability Engineering & System Safety, Vol. 94, Issue 11 https://doi.org/10.1016/j.ress.2009.06.001	journal	November 2009
FLOODING PSA BY CONSIDERING THE OPERATING EXPERIENCE DATA OF KOREAN PWRs Choi, Sun-Yeong; Yang, Joon-Eon Nuclear Engineering and Technology, Vol. 39, Issue 3 https://doi.org/10.5516/NET.2007.39.3.215	journal	June 2007
Application of piping failure database to nuclear safety issues in Korea Park, Jeong Soon; Choi, Young Hwan International Journal of Pressure Vessels and Piping, Vol. 90-91 https://doi.org/10.1016/j.ijpvp.2011.10.008	journal	February 2012
Markov models for evaluating risk-informed in-service inspection strategies for nuclear power plant piping systems Fleming, Karl N. Reliability Engineering & System Safety, Vol. 83, Issue 1 https://doi.org/10.1016/j.ress.2003.08.009	journal	January 2004
A comprehensive framework for evaluation of piping reliability due to erosion–corrosion for risk-informed inservice inspection Vinod, Gopika; Bidhar, S. K.; Kushwaha, H. S. Reliability Engineering & System Safety, Vol. 82, Issue 2 https://doi.org/10.1016/S0951-8320(03)00163-7	journal	November 2003
Korean Consortium's preliminary research for enhancing a probabilistic fracture mechanics code, PRO-LOCA Kim, Sun-Hye; Park, Jung-Soon; Lee, Jin-Ho International Journal of Pressure Vessels and Piping, Vol. 131 https://doi.org/10.1016/j.ijpvp.2015.04.012	journal	July 2015
Pedestrian: Probabilistic fracture mechanics analysis code based on direct sampling with replacement Nagai, Masaki; Miura, Naoki; Yamamoto, Masato International Journal of Pressure Vessels and Piping, Vol. 167 https://doi.org/10.1016/j.ijpvp.2018.10.008	journal	November 2018
Structural integrity assessment procedure for Europe – of the SINTAP programme overview Webster, Stephen; Bannister, Adam Engineering Fracture Mechanics, Vol. 67, Issue 6 https://doi.org/10.1016/S0013-7944(00)00070-9	journal	December 2000
ProSINTAP – A probabilistic program implementing the SINTAP assessment procedure Dillström, Peter Engineering Fracture Mechanics, Vol. 67, Issue 6 https://doi.org/10.1016/S0013-7944(00)00078-3	journal	December 2000
Probabilistic fracture analysis of cracked pipes with circumferential flaws Rahman, S. International Journal of Pressure Vessels and Piping, Vol. 70, Issue 3 https://doi.org/10.1016/S0308-0161(96)00034-8	journal	March 1997
The probability of leakage in piping systems of pressurized water reactors on the basis of fracture mechanics and operating experience Beliczey, S.; Schulz, H. Nuclear Engineering and Design, Vol. 102, Issue 3 https://doi.org/10.1016/0029-5493(87)90188-9	journal	July 1987
Study on Risk-Informed In-Service Inspection for BWR Piping Koriyama, Tamio; Li, Yinsheng; Hamaguchi, Yoshikane Journal of Nuclear Science and Technology, Vol. 46, Issue 8 https://doi.org/10.1080/18811248.2007.9711594	journal	August 2009
Assessment of the integrity of structures containing defects Milne, I.; Ainsworth, R. A.; Dowling, A. R. International Journal of Pressure Vessels and Piping, Vol. 32, Issue 1-4 https://doi.org/10.1016/0308-0161(88)90071-3	journal	January 1988
Experimental and Analytical Studies on the Effect of Excessive Loading on Fatigue Crack Propagation in Piping Materials Yamaguchi, Yoshihito; Katsuyama, Jinya; Onizawa, Kunio Journal of Pressure Vessel Technology, Vol. 135, Issue 4 https://doi.org/10.1115/1.4024454	journal	June 2013
Database development and uncertainty treatment for estimating pipe failure rates and rupture frequencies Fleming, Karl N.; Lydell, Bengt O. Y. Reliability Engineering & System Safety, Vol. 86, Issue 3 https://doi.org/10.1016/j.ress.2004.01.013	journal	December 2004
Best-Estimate Evaluation of Large-Break Loss-of-Coolant Accident for Advanced Natural Circulation Nuclear Reactor Bhasin, Vivek; Srivastava, A.; Rastogi, R. Nuclear Science and Engineering, Vol. 160, Issue 3 https://doi.org/10.13182/NSE160-318	journal	November 2008
Procedure for probabilistic safety assessment of leaks and breaks of piping systems Berg, H. -P.; Gersinska, R.; Sievers, J. International Journal of Pressure Vessels and Piping, Vol. 87, Issue 2-3 https://doi.org/10.1016/j.ijpvp.2009.12.006	journal	February 2010
Benchmark analysis on PFM analysis codes for aged piping of nuclear power plants Itoh, Hiroto; Li, Yinsheng; Osakabe, Kazuya Journal of Mechanical Science and Technology, Vol. 26, Issue 7 https://doi.org/10.1007/s12206-012-0514-4	journal	July 2012
Benchmark analysis on probabilistic fracture mechanics analysis codes concerning fatigue crack growth in aged piping of nuclear power plants Katsuyama, J.; Itoh, H.; Li, Y. International Journal of Pressure Vessels and Piping, Vol. 117-118 https://doi.org/10.1016/j.ijpvp.2013.10.010	journal	May 2014
A nonlinear mixed-effects model for degradation data obtained from in-service inspections Yuan, X. -X.; Pandey, M. D. Reliability Engineering & System Safety, Vol. 94, Issue 2 https://doi.org/10.1016/j.ress.2008.06.013	journal	February 2009
Risk–benefit analyses of SG tube maintenance based on probabilistic fracture mechanics Isobe, Y.; Sagisaka, M.; Yoshimura, S. Nuclear Engineering and Design, Vol. 207, Issue 3 https://doi.org/10.1016/S0029-5493(01)00338-7	journal	August 2001
A probabilistic approach to leak-before-break in CANDU pressure tubes Walker, J. R. International Journal of Pressure Vessels and Piping, Vol. 43, Issue 1-3 https://doi.org/10.1016/0308-0161(90)90104-P	journal	January 1990
A study on the reliability assessment methodology for pipelines with active corrosion defects Caleyo, F.; González, J. L.; Hallen, J. M. International Journal of Pressure Vessels and Piping, Vol. 79, Issue 1 https://doi.org/10.1016/S0308-0161(01)00124-7	journal	January 2002
Optimal condition-based maintenance decisions for systems with dependent stochastic degradation of components Hong, H. P.; Zhou, W.; Zhang, S. Reliability Engineering & System Safety, Vol. 121 https://doi.org/10.1016/j.ress.2013.09.004	journal	January 2014
Probabilistic Failure Analysis of Austenitic Nuclear Pipelines against Stress Corrosion Cracking Priya, C.; Rao, K. B.; Anoop, M. B. Proceedings of the Institution of Mechanical Engineers, Part C: Journal of Mechanical Engineering Science, Vol. 219, Issue 7 https://doi.org/10.1243/095440605X31526	journal	July 2005
Piping Failure Analysis for the Korean Nuclear Piping Including the Effect of In-Service Inspection Choi, Sun Yeong; Choi, Young Hwan Key Engineering Materials, Vol. 270-273 https://doi.org/10.4028/www.scientific.net/KEM.270-273.1731	journal	August 2004
Determination of rupture frequency for feeders subject to flow accelerated corrosion Wang, Min; Li, Ming; Lehockey, Edward Nuclear Engineering and Design, Vol. 241, Issue 6 https://doi.org/10.1016/j.nucengdes.2011.03.018	journal	June 2011
Probabilistic fracture assessment of piping systems based on FITNET FFS procedure Qian, Guian; Niffenegger, Markus Nuclear Engineering and Design, Vol. 241, Issue 3 https://doi.org/10.1016/j.nucengdes.2011.01.019	journal	March 2011
Probabilistic ageing and risk analysis tools for nuclear piping Qian, Guian; Chou, Hsoung-Wei; Niffenegger, Markus Nuclear Engineering and Design, Vol. 300 https://doi.org/10.1016/j.nucengdes.2016.02.033	journal	April 2016
Probabilistic elastic-plastic fracture analysis of circumferentially cracked pipes with finite-length surface flaws Rahman, Sharif Nuclear Engineering and Design, Vol. 195, Issue 3 https://doi.org/10.1016/S0029-5493(99)00214-9	journal	February 2000
Improvement of plant reliability based on combination of prediction and inspection of wall thinning due to FAC Uchida, Shunsuke; Okada, Hidetoshi; Naitoh, Masanori Nuclear Engineering and Design, Vol. 337 https://doi.org/10.1016/j.nucengdes.2018.06.007	journal	October 2018
A procedure for estimation of pipe break probabilities due to IGSCC Bergman, M.; Brickstad, B.; Nilsson, F. International Journal of Pressure Vessels and Piping, Vol. 74, Issue 3 https://doi.org/10.1016/S0308-0161(97)00122-1	journal	December 1997
OPDE—The international pipe failure data exchange project Lydell, Bengt; Riznic, Jovica Nuclear Engineering and Design, Vol. 238, Issue 8 https://doi.org/10.1016/j.nucengdes.2008.01.002	journal	August 2008
A Multi-State Physics Modeling approach for the reliability assessment of Nuclear Power Plants piping systems Di Maio, Francesco; Colli, Davide; Zio, Enrico Annals of Nuclear Energy, Vol. 80 https://doi.org/10.1016/j.anucene.2015.02.007	journal	June 2015
An empirical stress-intensity factor equation for the surface crack Newman, J. C.; Raju, I. S. Engineering Fracture Mechanics, Vol. 15, Issue 1-2 https://doi.org/10.1016/0013-7944(81)90116-8	journal	January 1981
Life Prediction and Monitoring of Nuclear Power Plant Components for Service-Related Degradation Simonen, Fredric A.; Gosselin, Stephen R. Journal of Pressure Vessel Technology, Vol. 123, Issue 1 https://doi.org/10.1115/1.1344237	journal	October 2000
Comparison of approaches for estimating pipe rupture frequencies for risk-informed in-service inspections Simola, Kaisa; Pulkkinen, Urho; Talja, Heli Reliability Engineering & System Safety, Vol. 84, Issue 1 https://doi.org/10.1016/j.ress.2003.10.008	journal	April 2004
Effect of Partial Welding on the Residual Stress and Structural Integrity of Piping Welds Katsuyama, Jinya; Masaki, Koichi; Onizawa, Kunio Journal of Pressure Vessel Technology, Vol. 135, Issue 6 https://doi.org/10.1115/1.4025087	journal	October 2013
Effects of alternative inspection strategies on piping reliability Khaleel, M. A.; Simonen, F. A. Nuclear Engineering and Design, Vol. 197, Issue 1-2 https://doi.org/10.1016/S0029-5493(99)00261-7	journal	April 2000
A Probabilistic Physics-of-Failure Approach to Prediction of Steam Generator Tube Rupture Frequency Chatterjee, Kaushik; Modarres, Mohammad Nuclear Science and Engineering, Vol. 170, Issue 2 https://doi.org/10.13182/NSE11-27	journal	February 2012
Benchmark analysis and numerical investigation on probabilistic fracture mechanics analysis codes for NPPs piping Li, Yinsheng; Itoh, Hiroto; Osakabe, Kazuya International Journal of Pressure Vessels and Piping, Vol. 99-100 https://doi.org/10.1016/j.ijpvp.2012.08.004	journal	November 2012
Probabilistic assessment of thermal fatigue in nuclear components Sudret, B.; Guédé, Z. Nuclear Engineering and Design, Vol. 235, Issue 17-19 https://doi.org/10.1016/j.nucengdes.2005.05.016	journal	August 2005
Life prediction of steam generator tubing due to stress corrosion crack using Monte Carlo Simulation Hu, Jun; Liu, Fei; Cheng, Guangxu Nuclear Engineering and Design, Vol. 241, Issue 10 https://doi.org/10.1016/j.nucengdes.2011.08.016	journal	October 2011
Probabilistic failure analysis of nuclear piping with empirical study of Taiwan's BWR plants You, Jang-Shyong; Wu, Wen-Fang International Journal of Pressure Vessels and Piping, Vol. 79, Issue 7 https://doi.org/10.1016/S0308-0161(02)00061-3	journal	July 2002
Leak-before-break analyses of PWR and BWR piping concerning size effects Heckmann, Klaus; Sievers, Jürgen Nuclear Engineering and Design, Vol. 326 https://doi.org/10.1016/j.nucengdes.2017.11.027	journal	January 2018
A New Strategy for In-Service Inspection of Nuclear Piping Considering Piping Failure Frequency Choi, Sun Yeong; Choi, Young Hwan Key Engineering Materials, Vol. 353-358 https://doi.org/10.4028/www.scientific.net/KEM.353-358.2088	journal	September 2007
Implementation of condition-dependent probabilistic risk assessment using surveillance data on passive components Lewandowski, Radoslaw; Denning, Richard; Aldemir, Tunc Annals of Nuclear Energy, Vol. 87 https://doi.org/10.1016/j.anucene.2015.07.035	journal	January 2016
Simulation-Informed Probabilistic Methodology for Common Cause Failure Analysis Sakurahara, Tatsuya; Schumock, Grant; Reihani, Seyed Reliability Engineering & System Safety, Vol. 185 https://doi.org/10.1016/j.ress.2018.12.007	journal	May 2019
Failure probabilistic analysis of steam generator heat-transfer tubing with pitting corrosion Zhou, B. H.; Zhai, Z. Q. Engineering Failure Analysis, Vol. 18, Issue 5 https://doi.org/10.1016/j.engfailanal.2011.03.022	journal	July 2011
Probabilistic damage estimation in piping components against thermal creep and fatigue Appalanaidu, Y.; Gupta, Sayan Nuclear Engineering and Design, Vol. 273 https://doi.org/10.1016/j.nucengdes.2014.04.002	journal	July 2014
Predicting tube repair at French nuclear steam generators using statistical modeling Mathon, C.; Chaudhary, A.; Gay, N. Nuclear Engineering and Design, Vol. 269 https://doi.org/10.1016/j.nucengdes.2013.08.044	journal	April 2014
The Impact of Probabilistic Modeling in Life-Cycle Management of Nuclear Piping Systems Pandey, M. D.; Lu, D.; Komljenovic, D. Journal of Engineering for Gas Turbines and Power, Vol. 133, Issue 1 https://doi.org/10.1115/1.4000897	journal	September 2010
Development of Probabilistic Fracture Mechanics Analysis Code for Pipes with Stress Corrosion Cracks Machida, Hideo; Arakawa, Manabu; Yamashita, Norimichi Journal of Power and Energy Systems, Vol. 3, Issue 1 https://doi.org/10.1299/jpes.3.103	journal	January 2009
Reliability Models of Aging Passive Components Informed by Materials Degradation Metrics to Support Long-Term Reactor Operations Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y. Nuclear Science and Engineering, Vol. 171, Issue 1 https://doi.org/10.13182/NSE11-18	journal	May 2012
Probabilistic and non-probabilistic failure assessment curves of primary coolant pipe contained internal circumferential surface crack in pressurized water reactor nuclear power plant Tian, Yun; Zhang, Guodong; Miao, Xinting Nuclear Engineering and Design, Vol. 322 https://doi.org/10.1016/j.nucengdes.2017.07.011	journal	October 2017
Stress-Intensity Factors for Internal and External Surface Cracks in Cylindrical Vessels Raju, I. S.; Newman, J. C. Journal of Pressure Vessel Technology, Vol. 104, Issue 4 https://doi.org/10.1115/1.3264220	journal	November 1982
A Review of the Progress with Statistical Models of Passive Component Reliability Lydell, Bengt O. Y. Nuclear Engineering and Technology, Vol. 49, Issue 2 https://doi.org/10.1016/j.net.2016.12.008	journal	March 2017
Acceptance criterion for probabilistic structural integrity assessment: Prediction of the failure pressure of steam generator tubing with fretting flaws Duan, Xinjian; Wang, Min; Kozluk, Michael Nuclear Engineering and Design, Vol. 281 https://doi.org/10.1016/j.nucengdes.2014.10.007	journal	January 2015
Probabilistic leak-before-break analysis with correlated input parameters Qian, Guian; Niffenegger, Markus; Karanki, Durga Rao Nuclear Engineering and Design, Vol. 254 https://doi.org/10.1016/j.nucengdes.2012.10.005	journal	January 2013
A probabilistic model of wall thinning in CANDU feeders due to flow-accelerated corrosion Yuan, X. -X.; Pandey, M. D.; Bickel, G. A. Nuclear Engineering and Design, Vol. 238, Issue 1 https://doi.org/10.1016/j.nucengdes.2007.06.004	journal	January 2008
Steady-State Model for Erosion-Corrosion of Feedwater Piping Abdulsalam, M.; Stanley, J. T. CORROSION, Vol. 48, Issue 7 https://doi.org/10.5006/1.3315976	journal	July 1992
An overview of the European flaw assessment procedure SINTAP and its validation Ainsworth, R. A.; Bannister, A. C.; Zerbst, U. International Journal of Pressure Vessels and Piping, Vol. 77, Issue 14-15 https://doi.org/10.1016/S0308-0161(01)00009-6	journal	December 2000
Combining discrete-time Markov processes and probabilistic fracture mechanics in RI-ISI risk estimates Cronvall, Otso; Männistö, Ilkka International Journal of Pressure Vessels and Piping, Vol. 86, Issue 11 https://doi.org/10.1016/j.ijpvp.2009.07.004	journal	November 2009
On the separation of aleatory and epistemic uncertainties in probabilistic assessments Jyrkama, Mikko I.; Pandey, Mahesh D. Nuclear Engineering and Design, Vol. 303 https://doi.org/10.1016/j.nucengdes.2016.04.013	journal	July 2016
Prognostics for integrity of steam generator tubes using the general path model Kim, Hyeonmin; Kim, Jung Taek; Heo, Gyunyoung Nuclear Engineering and Technology, Vol. 50, Issue 1 https://doi.org/10.1016/j.net.2017.10.006	journal	February 2018
Structural integrity aspects of reactor safety Vaze, K. K. Sadhana, Vol. 38, Issue 5 https://doi.org/10.1007/s12046-013-0173-x	journal	October 2013
Probabilistic fracture mechanics analysis of thermally aged nuclear piping in a pressurized water reactor Li, Shuxiao; Zhang, Hailong; Li, Shilei Nuclear Engineering and Design, Vol. 265 https://doi.org/10.1016/j.nucengdes.2013.06.034	journal	December 2013
Probabilistic modelling of steam generator tube pitting corrosion Datla, Suresh V.; Jyrkama, Mikko I.; Pandey, Mahesh D. Nuclear Engineering and Design, Vol. 238, Issue 7 https://doi.org/10.1016/j.nucengdes.2008.01.013	journal	July 2008
A Procedure for Analysis of leak Before Break in Pipes Subjected to Fatigue or Igscc Accounting for Complex Crack Shapes Bergman, M.; Brickstad, B. Fatigue & Fracture of Engineering Materials and Structures, Vol. 18, Issue 10 https://doi.org/10.1111/j.1460-2695.1995.tb00846.x	journal	October 1995
Case studies of risk-informed inservice inspection of nuclear piping systems You, Jang-Shyong; Kuo, Hung-Ta; Wu, Wen-Fang Nuclear Engineering and Design, Vol. 236, Issue 1 https://doi.org/10.1016/j.nucengdes.2005.06.014	journal	January 2006
Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method Lin, Y. C.; Xie, Y. J.; Wang, X. H. Nuclear Engineering and Design, Vol. 229, Issue 2-3 https://doi.org/10.1016/j.nucengdes.2003.12.007	journal	April 2004
Safety assessment of austenitic steel nuclear power plant pipelines against stress corrosion cracking in the presence of hybrid uncertainties Anoop, M. B.; Balaji Rao, K.; Lakshmanan, N. International Journal of Pressure Vessels and Piping, Vol. 85, Issue 4 https://doi.org/10.1016/j.ijpvp.2007.09.001	journal	April 2008
FITNET FFS procedure: A unified European procedure for structural integrity assessment Gutiérrez-Solana, F.; Cicero, S. Engineering Failure Analysis, Vol. 16, Issue 2 https://doi.org/10.1016/j.engfailanal.2008.02.007	journal	March 2009
A study on the effect of flaw detection probability assumptions on risk reduction achieved by non-destructive inspection Cronvall, O.; Simola, K.; Männistö, I. Reliability Engineering & System Safety, Vol. 105 https://doi.org/10.1016/j.ress.2012.03.012	journal	September 2012
Updating of risk-informed ISI programme for Ignalina RBMK-1500 nuclear power plant in Lithuania: Results and challenges Klimašauskas, A.; Alzbutas, R.; Kopustinskas, V. Nuclear Engineering and Design, Vol. 236, Issue 24 https://doi.org/10.1016/j.nucengdes.2006.03.019	journal	December 2006
Reliability analysis of stainless steel piping using a single stress corrosion cracking damage parameter Guedri, A. International Journal of Pressure Vessels and Piping, Vol. 111-112 https://doi.org/10.1016/j.ijpvp.2013.03.011	journal	November 2013
Pipe and vessel failure probability Thomas, H. M. Reliability Engineering, Vol. 2, Issue 2 https://doi.org/10.1016/0143-8174(81)90017-2	journal	April 1981
Benchmark analyses of probabilistic fracture mechanics for cast stainless steel pipe Hojo, Kiminobu; Hayashi, Shotaro; Nishi, Wataru Mechanical Engineering Journal, Vol. 3, Issue 4 https://doi.org/10.1299/mej.16-00083	journal	January 2016
Uncertainties in risk analysis: Six levels of treatment Paté-Cornell, M. Elisabeth Reliability Engineering & System Safety, Vol. 54, Issue 2-3 https://doi.org/10.1016/S0951-8320(96)00067-1	journal	November 1996
Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications Fleming, K. N.; Lydell, B. O. Y. Nuclear Engineering and Design, Vol. 305 https://doi.org/10.1016/j.nucengdes.2016.06.007	journal	August 2016
Development of probabilistic fracture mechanics analysis codes for reactor pressure vessels and piping considering welding residual stress Onizawa, Kunio; Nishikawa, Hiroyuki; Itoh, Hiroto International Journal of Pressure Vessels and Piping, Vol. 87, Issue 1 https://doi.org/10.1016/j.ijpvp.2009.11.011	journal	January 2010
PoD curves, their derivation, applications and limitations Georgiou, G. A. Insight - Non-Destructive Testing and Condition Monitoring, Vol. 49, Issue 7 https://doi.org/10.1784/insi.2007.49.7.409	journal	July 2007
Pipe failure probability—the Thomas paper revisited Lydell, B. O. Y. Reliability Engineering & System Safety, Vol. 68, Issue 3 https://doi.org/10.1016/S0951-8320(00)00016-8	journal	June 2000
The probability distribution of maintenance cost of a system affected by the gamma process of degradation: Finite time solution Cheng, Tianjin; Pandey, Mahesh D.; van der Weide, J. A. M. Reliability Engineering & System Safety, Vol. 108 https://doi.org/10.1016/j.ress.2012.06.005	journal	December 2012
RoverD: Use of Test Data in GSI-191 Risk Assessment Kee, Ernie; Hasenbein, John; Zolan, Alex Nuclear Technology, Vol. 196, Issue 2 https://doi.org/10.13182/NT16-34	journal	November 2016
A new repair criterion for steam generator tubes with axial cracks based on probabilistic integrity assessment Kim, Hyun-Su; Oh, Chang-Kyun; Chang, Yoon-Suk Nuclear Engineering and Design, Vol. 313 https://doi.org/10.1016/j.nucengdes.2016.11.027	journal	March 2017
Reliability analysis of nuclear piping system using semi-Markov process model Veeramany, Arun; Pandey, Mahesh D. Annals of Nuclear Energy, Vol. 38, Issue 5 https://doi.org/10.1016/j.anucene.2010.12.012	journal	May 2011
Comparative analysis of deterministic and probabilistic fracture mechanical assessment tools Heckmann, K.; Saifi, Q. Kerntechnik, Vol. 81, Issue 5 https://doi.org/10.3139/124.110725	journal	October 2016
Failure rate updates using condition-based prognostics in probabilistic safety assessments Kim, Hyeonmin; Kim, Jung Taek; Heo, Gyunyoung Reliability Engineering & System Safety, Vol. 175 https://doi.org/10.1016/j.ress.2018.03.022	journal	July 2018
Comments on probabilities of leaks and breaks of safety-related piping in PWR plants Beliczey, S.; Schulz, H. International Journal of Pressure Vessels and Piping, Vol. 43, Issue 1-3 https://doi.org/10.1016/0308-0161(90)90103-O	journal	January 1990
Stress Intensity Factors for Circumferential Surface Cracks in Pipes Bergman, M. Fatigue & Fracture of Engineering Materials and Structures, Vol. 18, Issue 10 https://doi.org/10.1111/j.1460-2695.1995.tb00845.x	journal	October 1995
Leak-before-break in the pressure tubes of CANDU reactors Moan, G. D.; Coleman, C. E.; Price, E. G. International Journal of Pressure Vessels and Piping, Vol. 43, Issue 1-3 https://doi.org/10.1016/0308-0161(90)90089-Z	journal	January 1990
An accurate analysis of maintenance cost of structures experiencing stochastic degradation Cheng, T.; Pandey, M. D. Structure and Infrastructure Engineering, Vol. 8, Issue 4 https://doi.org/10.1080/15732479.2011.563088	journal	April 2012
Uncertainty in probabilistic risk assessment Winkler, Robert L. Reliability Engineering & System Safety, Vol. 54, Issue 2-3 https://doi.org/10.1016/S0951-8320(96)00070-1	journal	November 1996
Evaluation of environmental effects on fatigue life of piping Simonen, F. A.; Khaleel, M. A.; Phan, H. K. Nuclear Engineering and Design, Vol. 208, Issue 2 https://doi.org/10.1016/S0029-5493(01)00373-9	journal	September 2001
Condition-based probabilistic safety assessment of a spontaneous steam generator tube rupture accident scenario Di Maio, Francesco; Antonello, Federico; Zio, Enrico Nuclear Engineering and Design, Vol. 326 https://doi.org/10.1016/j.nucengdes.2017.10.020	journal	January 2018
Application of particle filtering for prognostics with measurement uncertainty in nuclear power plants Kim, Gibeom; Kim, Hyeonmin; Zio, Enrico Nuclear Engineering and Technology, Vol. 50, Issue 8 https://doi.org/10.1016/j.net.2018.08.002	journal	December 2018
Evaluation of Nuclear Piping Failure Frequency in Korean Pressurized Water Reactors Choi, Sun Yeong; Choi, Young Hwan Key Engineering Materials, Vol. 297-300 https://doi.org/10.4028/www.scientific.net/KEM.297-300.1645	journal	November 2005
Reliability data update using condition monitoring and prognostics in probabilistic safety assessment Kim, Hyeonmin; Lee, Sang-Hwan; Park, Jun-Seok Nuclear Engineering and Technology, Vol. 47, Issue 2 https://doi.org/10.1016/j.net.2014.12.008	journal	March 2015
Assessment of Steam Generator Tube Flaw Size and Leak Rate Models Revankar, Shripad T.; Riznic, Jovica R. Nuclear Technology, Vol. 167, Issue 1 https://doi.org/10.13182/NT09-A8859	journal	July 2009
Effect of through-wall stress gradients on piping failure probabilities Khaleel, M. A.; Simonen, F. A. Nuclear Engineering and Design, Vol. 197, Issue 1-2 https://doi.org/10.1016/S0029-5493(99)00259-9	journal	April 2000

Similar Records

Integrating renewal process modeling with Probabilistic Physics-of-Failure: Application to Loss of Coolant Accident (LOCA) frequency estimations in nuclear power plants

Journal Article · Mon Apr 22 00:00:00 EDT 2019 · Reliability Engineering and System Safety · OSTI ID:1848156

Sakurahara, Tatsuya; O'Shea, Nicholas; Cheng, Wen-Chi; +4 more

Methods development to evaluate the risk of upgrading to DCS: The human factor

Technical Report · Sat Apr 01 00:00:00 EST 1995 · OSTI ID:1848156

Ostrom, L. T.; Wilhelmsen, C. A.

Advanced Test Reactor probabilistic risk assessment methodology and results summary

Technical Report · Wed Jan 01 00:00:00 EST 1992 · OSTI ID:1848156

Eide, S A; Atkinson, S A; Thatcher, T A

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
Nuclear Science & Technology
Reactor Coolant Pressure Boundary
Piping
Steam generator tubes
Loss of coolant accident
Probabilistic safety assessment
Probabilistic risk assessment

Title: Review and categorization of existing studies on the estimation of probabilistic failure metrics for Reactor Coolant Pressure Boundary piping and steam generator tubes in Nuclear Power Plants

Citation Formats

References (102)

Similar Records

Related Subjects