Abstract

In the past few decades, cracks because of stress corrosion cracking (SCC) have been detected in the dissimilar weld joints using nickel-based alloy in piping system of boiling water reactors (BWRs). Thus, the structural integrity assessment for such weld joints has become important. Nowadays, probabilistic fracture mechanics (PFM) analysis is recognized as a rational method for structural integrity assessment because it can consider inherent uncertainties of various influencing factors as probability distributions and quantitatively evaluate the failure probability of a cracked component. The Japan Atomic Energy Agency has developed a PFM analysis code PASCAL-SP for a weld joint of piping system in nuclear power plant. This study improves the analysis functions of PASCAL-SP for weld joint using nickel-based alloy in BWR susceptible to SCC. As an analysis example of the improved version of PASCAL-SP, the failure probability of a weld joint is quantitatively evaluated. Furthermore, sensitivity analyses are conducted concerning the effect of leak detection and in-service inspection. From the analysis results, it is concluded that the improved version of PASCAL-SP is useful for structural integrity assessment.

Issue Section:
Materials and Fabrication
Keywords:
dissimilar weld joint, failure probability, stress corrosion cracking, nickel-based alloy
Topics:
Alloys, Boiling water reactors, Failure, Fracture (Materials), Nickel, Probability, Sensitivity analysis, Piping systems, Stress corrosion cracking, Leakage, Fracture mechanics, Stress

References

1.
Japan Society of Mechanical Engineers
,
2016
, “
JSME Codes for Nuclear Power Generation Facilities - Rules on Fitness-for-Service for Nuclear Power Plant
,” JSME, Tokyo, Japan, Standard No. S NA1-2016 (in Japanese).
2.
Kobayashi
,
H.
, and
Kashima
,
K.
,
2000
, “
Overview of JSME Flaw Evaluation Code for Nuclear Power Plants
,”
Int. J. Pressure Vessels Piping
,
77
(
14–15
), pp.
937
–
944
.10.1016/S0308-0161(01)00016-3
3.
Westinghouse Electric Company
,
1999
, “
Westinghouse Owners Group Application of Risk Informed Methods to Piping Inservise Inspection Topical Report
,” Westinghouse Electric Company, Pittsburgh, PA, Standard No. WCAP-14572 Revision1-NP-A.
4.
Westinghouse Electric Company
,
2012
, “
Risk-Informed Extension of the Reactor Vessel Nozzle Inservice Inspection Interval
,” Westinghouse Electric Company, Cranberry Township, PA, Standard No. WCAP-17236-NP-A Revision 0.
5.
Li
,
Y.
,
Osakabe
,
K.
,
Katsumata
,
G.
,
Katsuyama
,
J.
,
Onizawa
,
K.
, and
Yoshimura
,
S.
,
2014
, “
Benchmark Analysis on Probabilistic Fracture Mechanics Analysis Codes Concerning Multiple Cracks and Crack Initiation in Aged Piping of Nuclear Power Plants
,”
ASME
Paper No. PVP2014-28513.10.1115/P VP2014-28513
6.
Mano
,
A.
,
Yamaguchi
,
Y.
,
Katsuyama
,
J.
, and
Li
,
Y.
,
2019
, “
Improvement of Probabilistic Fracture Mechanics Code PASCAL-SP With Regard to PWSCC
,”
ASME J. Nucl. Eng. Radiat. Sci.
,
5
(
3
), p.
031501
.10.1115/1.4042115
Google Scholar
Crossref
Search ADS
 
7.
Hojo
,
K.
,
Hayashi
,
S.
,
Nishi
,
W.
,
Kamaya
,
M.
,
Katsuyama
,
J.
,
Masaki
,
K.
,
Nagai
,
M.
,
Okamoto
,
T.
,
Takada
,
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
.10.1299/mej.16-00083
8.
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. PVP2011-57498.10.1115/P VP2011-57498
9.
Mano
,
A.
,
Katsuyama
,
J.
, and
Li
,
Y.
,
2020
, “
Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes –Analysis by PASCAL-SP
,”
ASME
Paper No. PVP2020-21427.10.1115/P VP2020-21427
10.
Tohoku Electric Power Corporation
,
2010
, “
The Results of Structural Integrity Assessment for the Primary Loop Recirculation System Piping of Onagawa Unit 1
,” Tohoku Electric Power Corporation, Sendai, Japan (in Japanese).
11.
Bush
,
S. H.
,
1992
, “
Overview of Piping Issues and Trends
,”
Fourth Symposium on Current Issues Related to Nuclear Power Plant Structure, Equipment & Piping
, Orlando, FL, Dec. 9–11, pp. III/1(1–24).
12.
Czajkowski
,
C. J.
,
1986
, “
Evaluation of Type 182 Weld Metal Cracks at the Pilgrim Nuclear Power Station
,”
Proceedings of the Technical Symposium of Corrosion 86
, No. 253, Houston, TX, Mar. 17, pp.
1
–
9
.
13.
International Atomic Energy Agency
,
2011
, “
Stress Corrosion Cracking in Light Water Reactors: Good Practice and Lessons Learned
,” IAEA, Vienna, Austria, Nuclear Energy Series No. NP-T-3.13.
14.
Japan Society of Mechanical Engineers
,
2015
, “
Code Case - SCC Crack Growth Rate of Nickel Alloy in PWR Primary Water Environment
,” JSME, Tokyo, Japan, Standard No. S NA-CC-006. (in Japanese).
15.
Japan Society of Mechanical Engineers
,
2015
, “
Code Case - Fatigue Crack Growth Rate of Nickel Alloy in PWR Primary Water Environment
,” JSME, Tokyo, Japan, Standard No. S NA-CC-004. (in Japanese).
16.
Japan Society of Mechanical Engineers
,
2015
, “
Code Case - SCC Crack Growth Rate of Nickel Alloy Weld in BWR Environment
,” JSME, Tokyo, Japan, Standard No. S NA-CC-009. (in Japanese).
17.
American Society of Mechanical Engineers
,
2019
, “
ASME Boiler and Pressure Vessel Code Section XI Rules for in-Service Inspection of Nuclear Power Plant Components
,” ASME, New York, Standard No. StBPVC-XI-2019.
18.
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
,” U.S. Nuclear Regulatory Commission, Washington, DC, Report No. NUREG/CR-5864.
19.
Moody
,
F. J.
,
1966
, “
Maximum Two-Phase Vessel Blowdown From Pipes
,”
ASME J. Heat Transfer-Trans. ASME
,
88
(
3
), pp.
285
–
294
.10.1115/1.3691539
Google Scholar
Crossref
Search ADS
 
20.
Kupperman
,
D. S.
,
Sheen
,
S. H.
,
Diercks
,
D. R.
,
Krishnaswamy
,
P.
,
Rudland
,
D.
, and
Wilkowski
,
G. M.
,
2004
, “
Barrier Integrity Research Program: Final Report
,” U.S. Nuclear Regulatory Commission, Washington, DC, Report No. NUREG/CR-6861.
21.
Udagawa
,
M.
,
Katsuyama
,
J.
,
Onizawa
,
K.
, and
Li
,
Y.
,
2015
, “
Failure Probability Analyses for PWSCC in Ni-Based Alloy Welds
,”
Int. J. Pressure Vessels Piping
,
131
, pp.
85
–
95
.10.1016/j.ijpvp.2015.04.013
Google Scholar
Crossref
Search ADS
 
22.
Udagawa
,
M.
,
Katsuyama
,
J.
, and
Onizawa
,
K.
,
2014
, “
Failure Probability Analyses for PWSCC and NiSCC in Ni-Based Alloy Welds
,”
Proceedings of the 10th International Workshop on the Integrity of Nuclear Components
, Busan, Korea, Apr. 16–18, pp.
219
–
226
.https://jopss.jaea.go.jp/search/servlet/search?5045011&language=1
23.
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, the Minerals, Metals and Materials Society
,
F. P.
Ford
,
S. M.
Bremer
, and
G. S.
Was
, eds., Newport Beach, CA, Aug. 1–5, pp.
243
–
250
.10.1002/9781118787618.ch25
Google Scholar
Crossref
Search ADS
 
24.
Couvant
,
T.
, and
Vaillant
,
F.
,
2011
, “
Initiation of PWSCC of Weld Alloy 182
,”
15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems -Water Reactors, the Minerals, Metals & Materials Society
, Colorado Springs, CO, Aug. 7–11, pp.
1141
–
1154
.10.1007/978-3-319-48760-1_69
25.
Saito
,
N.
,
Tanaka
,
S.
, and
Sakamoto
,
H.
,
1999
, “
Effect of Corrosion Potential on the SCC Initiation Lifetime of Alloy 182 Weld Metal
,” 9th International
Symposium on Environmental Degradation of Materials in Nuclear Power System – Water Reactors
, Newport Beach, CA, Aug. 1–5, pp.
493
–
499
.10.1002/9781118787618.ch51
Google Scholar
Crossref
Search ADS
 
26.
Akashi
,
M.
,
1994
, “
An Exponential Distribution Model for Assessing the Stress Corrosion Cracking Lifetime of BWR Component Materials
,”
Life Prediction of Corrodible Structures, R. N. Parkins, ed., Houston, Vol. II, pp. 1040–1049.
27.
Nagase
,
H.
,
Fujimori
,
H.
,
Saito
,
T.
,
Yamamoto
,
Y.
,
Takamori
,
K.
,
Dozaki
,
K.
,
Arai
,
T.
, and
Yuya
,
H.
,
2010
, “
Proposal of SCC Crack Growth Rate Curve for Nickel Base Alloy Metal in BWR Environments
,”
Britten Jpn. Soc. Mech. Eng.
,
76
(
764
), pp. 81–83 (in Japanese).
28.
Ogawa
,
T.
,
Itatani
,
Saito
,
T.
,
Aoike
,
M.
,
Nagase
,
S.
, and
Yoneda
,
H.
,
2014
, “
Technical Basis of Fatigue Crack Growth Rate Curvefor Ni-Base Alloy in BWR Water Environment
,”
ASME
Paper No. ICONE22-31143.10.1115/ICONE22-31143
29.
Japan Society of Mechanical Engineers
,
2015
, “
Code Case – Analytical Evaluation of Flaws in Nickel Based Alloy
,” JSME, Tokyo, Japan, Standard No. NA-CC-005-1. (in Japanese).
30.
Machida
,
H.
,
2007
, “
Reliability Assessment of PLR Piping Based on Domestic SCC Data
,”
ASME
Paper No. PVP2007-26059.10.1115/PVP2007-26059
Google Scholar
Crossref
Search ADS
 
31.
Ministry of International Trade and Industry
,
1980
, “
Technical Standard for Structural Design of Mechanical Components of Nuclear Power Facilities
,” MITI, Tokyo, Japan, Notification No. 501 (in Japanese).
32.
Japan Society of Mechanical Engineers
,
2016
, “
Codes for Nuclear Power Generation Facilities -Rules on Design and Construction for Nuclear Power Plants
,” JSME, Tokyo, Japan, Standard No. S NC1-2016 (in Japanese).
33.
Rudland
,
D.
,
Xu
,
H.
,
Wilkowski
,
G.
,
Scott
,
P.
,
Ghadiali
,
N.
, and
Brust
,
F.
,
2005
, “
Evaluation of Loss-of-Coolant Accident (LOCA) Frequencies Using the PRO-LOCA Code
,” Engineering Mechanics Corporation of Columbus, Columbus, OH, Report of Engineering Mechanics Corporation of Columbus in December 2005.
34.
Japan Society of Mechanical Engineers
,
2002
, “
Codes for Nuclear Power Generation Facilities -Rules on Protection Design Against Postulated Piping Failures in Nuclear Power Plants
,” JSME, Tokyo, Japan, Standard No. S ND1-2002 (in Japanese).
35.
Takahashi
,
Y.
,
2011
, “
Leak-Before-Break Assessment in JSME Standard
,”
ASME
Paper No. PVP2011-57978.10.1115/P VP2011-57978
Google Scholar
Crossref
Search ADS
 
36.
Nuclear Regulation Authority
,
2020
, “
Establishment of Interpretation for Cracks and Other Flaws Which Cause Fractures in Commercial Nuclear Power Plants and Surrounding Facilities
,” Nuclear Regulation Authority, Tokyo, Japan, Document No. 20033110 (in Japanese).
37.
Electric Power Research Institute
,
2009
, “
Material Reliability Program: Development of Probability of Detection Curves for Ultrasonic Examination of Dissimilar Metal Welds
,” Electric Power Research Institute, Palo Alto, CA, Standard No. MRP-262 Revision 1.
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