Probabilistic fracture mechanics (PFM) analysis is expected to be a rational method for structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for structural integrity assessment of piping welds in nuclear power plants (NPP). In the past few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus, structural integrity assessments considering PWSCC have become important. In this study, PASCAL-SP was improved considering PWSCC by introducing several analytical functions such as the models for evaluation of crack initiation time, crack growth rate (CGR), and probability of crack detection. By using the improved version of PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were numerically evaluated. Moreover, the influence of leak detection and nondestructive examination (NDE) on failure probabilities was detected. Based on the obtained numerical results, it was concluded that the improved version of PASCAL-SP is useful for evaluating the failure probability of a pipe considering PWSCC.
Issue Section:
Research Papers
References
1.
JSME
, 2012
, “JSME Codes for Nuclear Power Generation Facilities—Rules on Fitness-for-Service for Nuclear Power Plant
,” Japan Society of Mechanical Engineers, Tokyo, Japan (in Japanese).2.
Itoh
, H.
, Kato
, D.
, Osakabe
, K.
, Nishikawa
, H.
, and Onizawa
, K.
, 2010
, “User's Manuals of Probabilistic Fracture Mechanics Analysis Code for Aged Piping, PASCAL-SP
,” Japan Atomic Energy Agency, Tokai, Japan, JAEA-Data/Code 2009-025 (in Japanese).3.
Katsuyama
, J.
, Itoh
, H.
, Li
, Y.
, Osakabe
, K.
, Onizawa
, K.
, and Yoshimura
, S.
, 2014
, “Benchmark Analysis on Probabilistic Fracture Mechanics Analysis Codes Concerning Fatigue Crack Growth in Aged Piping of Nuclear Power Plants
,” Int. J. Pressure Vessels Piping
, 117–118
, pp. 56
–63
.4.
Arakawa
, M.
, Machida
, H.
, Narumi
, K.
, and Onizawa
, K.
, 2011
, “Benchmark Analysis on the Failure Probability Assessment of Piping With Stress Corrosion Cracking
,” ASME
Paper No. PVP2011-57498.5.
Hojo
, K.
, Hayashi
, S.
, Nishi
, W.
, Kamaya
, M.
, Katsuyama
, J.
, Masaki
, K.
, Nagai
, M.
, Okamoto
, T.
, Takeda
, Y.
, and Yoshimura
, S.
, 2016
, “Benchmark Analysis of Probabilistic Fracture Mechanics for Cast Stainless Steel Pipe
,” Bull. JSME Mech. Eng. J.
, 3
(4
), p. 16-00083.6.
Udagawa
, M.
, Katsuyama
, J.
, and Onizawa
, K.
, 2013
, “User's Manuals of Probabilistic Fracture Mechanics Analysis Code for Ni-Based Alloy Welds, PASCAL-NP
,” Japan Atomic Energy Agency, Tokyo, Japan, Report No. JAEA-Data/Code 2013-013 (in Japanese).7.
ASME
, 2015
, “ASME Boiler and Pressure Vessel Code Section XI Rules for In-Service Inspection of Nuclear Power Plant Components
,” American Society of Mechanical Engineers, New York, Standard No. BPVC-XI-2015
. https://www.asme.org/products/codes-standards/bpvcxi-2015-bpvc-section-xirules-inservice8.
Harris
, D. O.
, Dedhia
, D.
, and Lu
, S. C.
, 1992
, “Theoretical and User's Manual for PC-PRAISE, A Probabilistic Fracture Mechanics Computer Code for Piping Reliability Analysis
,” Engineering Mechanics Technology, Inc., San Jose, CA, Report No. NUREG/CR-5864
.https://www.osti.gov/biblio/10162360-theoretical-user-manual-pc-praise-probabilistic-fracture-mechanics-computer-code-piping-reliability-analysis9.
JSME
, 2002
, “Codes for Nuclear Power Generation Facilities—Rules on Protection Design Against Postulated Pipe Rupture for Nuclear Power Plants
,” Japan Society of Mechanical Engineers, Tokyo, Japan, Standard No. JSME S ND1-2002 (in Japanese).10.
Amzallag
, C.
, Hong
, S. L.
, Pages
, C.
, and Gelpi
, A.
, 1999
, “Stress Corrosion Life Assessment of Alloy 600 PWR Components
,” Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors
, F. P. Ford, S. M. Bremer, and G. S. Was, eds., Newport Beach, CA, Aug. 1–5, pp. 243
–250
.11.
Rudland, D., 2011, “xLPR Version 1.0 Report - Technical Basis and Pilot Study Problem Results,” U.S. Nuclear Regulatory Commission, Washington, DC, ADAMS Accession No.
ML110660292
. https://www.nrc.gov/docs/ML1106/ML110660292.pdf12.
JSME
, 2010
, “JSME Code Case—SCC Crack Growth Rate of High Nickel Alloy in PWR Primary Water Environment
,” Japan Society of Mechanical Engineers, Tokyo, Japan, Standard No. JSME S NA1-2008 NA-CC-006 (in Japanese).13.
JSME
, 2010
, “JSME Code Case—Fatigue Crack Growth Rate of High Nickel Alloy in Primary Water Environment
,” Japan Society of Mechanical Engineers, Tokyo, Japan, Standard No. JSME S NA1-2008 NA-CC-004 (in Japanese).14.
Bamford
, W. H.
, 1980
, “Technical Basis for Revised Reference Crack Growth Rate Curves for Pressure Boundary Steels in LWR Environment
,” ASME J. Pressure Vessel Technol.
, 102
(4
), pp. 433
–442
.15.
Hunt
, S.
, Broussard
, J.
, O'Regan
, P.
, and Covill
, D.
, 2005
, “Weld Overlay Deposit on Alloy 82/182 Butt Welds to Reduce ID Surface Stresses, Proceedings of the Conference on Vessel Penetration Inspection, Crack Growth and Repair
,” U.S. Nuclear Regulatory Commission, Washington, DC, Report No. NUREG/CP-0191.16.
AFCEN
, 2012
, “RCC-MRx Code—Design and Construction Rules for Mechanical Components of Nuclear Installations
,” AFCEN, Paris, France, Standard No. RCC-MRx 2012.17.
Li
, Y.
, Doi
, H.
, Hasegawa
, K.
, Osakabe
, K.
, and Okada
, H.
, 2013
, “Stress Intensity Factors for Cracks With Large Aspect Ratio in Cylinder
,” ASME
Paper No. PVP2013-97665.18.
Zahoor
, A.
, 1989
, “Ductile Fracture Handbook,” Electric Power Research Institute
, Palo Alto, CA, Report No. NP-6301-D V3
. https://www.scribd.com/document/355828169/Ductile-Fracture-Handbook19.
EPRI
, 2009
, “Material Reliability Program: Development of Probability of Detection Curves for Ultrasonic Examination of Dissimilar Metal Welds
,” Electric Power Research Institute, Palo Alto, CA, Report No. MRP-262 Revision 1.20.
Fredette
, L. F.
, Scott
, P. M.
, Brust
, F. W.
, and Csontos
, A.
, 2009
, “An Analytical Evaluation of the Full Structural Weld Overlay as a Stress Improving Mitigation Strategy to Prevent Primary Stress Corrosion Cracking in Pressurized Water Reactor Piping
,” ASME
Paper No. PVP2009-77327.21.
Japan Nuclear Energy Safety Organization
, 2009
, “Technical Document About Improvement of Analytical Methodology for Failure Probability of Structures and Components Due to Seismic Stress Considering Aging Effect
,” Japan Nuclear Energy Safety Organization, Tokyo, Japan (in Japanese).22.
JSME
, 2008
, “Codes for Nuclear Power Generation Facilities—Rules on Protection Design and Construction for Nuclear Power Plants
,” Japan Society of Mechanical Engineers, Tokyo, Japan, Standard No. JSME S NC1-2008 (in Japanese).23.
Shim
, D. J.
, Xu
, S.
, and Lee
, D.
, 2014
, “Closed-Form Stress Intensity Factor Solutions for Circumferential Through-Wall Cracks in Cylinder
,” ASME
Paper No. PVP2014-28049.24.
Kupperman
, D. S.
, Sheen
, S. H.
, Shack
, W. J.
, Diercks
, D. R.
, Krishnaswamy
, P.
, Rudland
, D.
, and Wilkowski
, G. M.
, 2001
, “Barrier Integrity Research Program: Final Report
,” U.S. Nuclear Regulatory Commission, Washington, DC, Report No. NUREG/CR-6861
.https://www.nrc.gov/docs/ML0435/ML043580207.pdfCopyright © 2019 by ASME
You do not currently have access to this content.
