Abstract

In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL was developed for structural integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. By reflecting the latest knowledge and findings, the evaluation functions are continuously improved and have been incorporated into PASCAL4 which is the most recent version of the PASCAL code. In this paper, the improvements made in PASCAL4 are explained in detail, such as the evaluation model of warm prestressing (WPS) effect, evaluation function of confidence levels for PFM analysis results by considering the epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions, and improved methods for KI calculations when considering complicated stress distributions. Moreover, using PASCAL4, PFM analysis examples considering these improvements are presented.

References

1.
Japan Electric Association
,
2016
, “
Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components
,”
Japan Electric Association
,
Tokyo, Japan
, Report No. JEAC4206-2016 (in Japanese).
2.
Structural Integrity Associates
,
1998
, “
Vessel Inspection Program Evaluation for Reliability
,” VIPER Version 1.2,
Structural Integrity Associates
,
San Jose, CA
.
3.
Williams
,
P. T.
,
Dickson
,
D. L.
,
Bass
,
B. R.
, and
Klasky
,
H. B.
,
2016
, “
Fracture Analysis of Vessels—Oak Ridge FAVOR, v16.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations
,”
Oak Ridge National Laboratory
,
Oak Ridge, TN
, Report No. ORNL/LTR-2016/309.
4.
Spencer
,
B. W.
,
Hoffman
,
W. M.
, and
Backman
,
M. A.
,
2019
, “
Modular System for Probabilistic Fracture Mechanics Analysis of Embrittled Reactor Pressure Vessels in the Grizzly Code
,”
Nucl. Eng. Des.
,
341
, pp.
25
37
.10.1016/j.nucengdes.2018.10.015
5.
United States Nuclear Regulatory Commission
,
2010
, “
Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events
,”
Title 10, Code of Federal Regulations, Part 50, Section 50.61a
,
United States Nuclear Regulatory Commission
,
Rockville, Maryland, USA
.
6.
Masaki
,
K.
,
Nishikawa
,
H.
,
Osakabe
,
K.
, and
Onizawa
,
K.
,
2011
, “
User's Manual and Analysis Methodology of Probabilistic Fracture Mechanics Analysis Code PASCAL3 for Reactor Pressure Vessel
,”
Japan Atomic Energy Agency
,
Tokai, Ibaraki, Japan
, Report No. JAEA-DATA/CODE 2010-033 (in Japanese).
7.
Japan Electric Association
,
2013
, “
Method of Surveillance Tests for Structural Materials of Nuclear Reactors
,”
Japan Electric Association
,
Tokyo, Japan
, Report No. JEAC4201-2007 (sup. 2013) (in Japanese).
8.
Katsuyama
,
J.
,
Osakabe
,
K.
,
Uno
,
S.
,
Li
,
Y.
, and
Yoshimura
,
S.
,
2017
, “
Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels
,”
ASME Paper No. PVP2017-65921
.10.1115/PVP2017-65921
9.
Japan Society of Mechanical Engineers
,
2016
, “
Codes for Nuclear Power Generation Facilities—Rules on Fitness-for-Service for Nuclear Power Plants
,”
JSME
,
Tokyo, Japan
, Standard No. JSME S NA1-2016.
10.
Marie
,
S.
, and
Chapuliot
,
S.
,
2008
, “
Improvement of the Calculation of the Stress Intensity Factors for Underclad and Through-Clad Defects in a Reactor Pressure Vessel Subjected to a Pressurized Thermal Shock
,”
Int. J. Pressure Vessels Piping
,
85
(
8
), pp.
517
531
.10.1016/j.ijpvp.2008.02.006
11.
American Society of Mechanical Engineers
,
2017
, “
ASME B&PV Code Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components
,”
ASME
,
New York
, Standard No. ASME BPVC XI 2017.
12.
Lu
,
K.
,
Mano
,
A.
,
Katsuyama
,
J.
,
Li
,
Y.
, and
Iwamatsu
,
F.
,
2018
, “
Development of Stress Intensity Factor Solutions for Subsurface Flaws in Plates Subjected to Polynomial Stress Distributions
,”
ASME J. Pressure Vessel Technol.
,
140
(
3
), p.
031201
.10.1115/1.4039125
13.
AFCEN
,
2015
, “
In-Service Inspection Rules for Mechanical Components of PWR Nuclear Islands
,”
RSE-M Appendix 5.4
,
AFCEN
,
Courbevoie, France
, pp.
21
23
.
14.
Marie
,
S.
, and
Nédélec
,
M.
,
2007
, “
A New Plastic Correction for the Stress Intensity Factor of an Under-Clad Defect in a PWR Vessel Subjected to a Pressurised Thermal Shock
,”
Int. J. Pressure Vessels Piping
,
84
(
3
), pp.
159
170
.10.1016/j.ijpvp.2006.09.019
15.
Lu
,
K.
,
Katsuyama
,
J.
, and
Li
,
Y.
,
2016
, “
Plasticity Correction on the Stress Intensity Factor Evaluation for Underclad Cracks Under Pressurized Thermal Shock Events
,”
ASME Paper No. PVP2016-63486
.10.1115/PVP2016-63486
16.
Lu
,
K.
,
Katsuyama
,
J.
,
Uno
,
S.
, and
Li
,
Y.
,
2017
, “
Probabilistic Fracture Mechanics Analysis Models for Japanese Reactor Pressure Vessels
,”
ASME Paper No. PVP2017-66003
.10.1115/PVP2017-66003
17.
Li
,
Y.
,
Katsumata
,
G.
,
Masaki
,
K.
,
Hayashi
,
S.
,
Itabashi
,
Y.
,
Nagai
,
M.
,
Suzuki
,
M.
, and
Kanto
,
Y.
,
2017
, “
Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL
,”
ASME Paper No. ICONE25-66468
.10.1115/ICONE25-66468
18.
Li
,
Y.
,
Uno
,
S.
,
Katsuyama
,
J.
,
Dickson
,
T. L.
, and
Kirk
,
M.
,
2017
, “
Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL Through Benchmark Analyses With FAVOR
,”
ASME Paper No. PVP2017-66004
.10.1115/PVP2017-66004
19.
Kanto
,
Y.
,
Li
,
Y.
, and
Yoshimura
,
S.
,
2016
, “
Summary of Results From Japanese Participants in Round-Robin Analyses by Probabilistic Fracture Mechanics for BWR Pressure Vessels During LTOP Event
,”
11th International Workshop on the Integrity of Nuclear Components
, Nagasaki, Japan, Apr. 11–13, pp.
91
98
.
20.
Chapuliot
,
S.
,
Izard
,
J.
,
Moinereau
,
D.
, and
Marie
,
S.
,
2010
, “
WPS Criterion Proposition Based on Experimental Data Base Interpretation
,” FONTEVRAUD 7, Avignon, France, Sept. 26–30, Paper No. A141, pp.
1
10
.
21.
Katsuyama
,
J.
,
Masaki
,
K.
,
Miyamoto
,
Y.
, and
Li
,
Y.
,
2018
, “
User's Manual and Analysis Methodology of Probabilistic Fracture Mechanics Analysis Code PASCAL4 for Reactor Pressure Vessel
,”
Japan Atomic Energy Agency
,
Tokai, Ibaraki, Japan
, Report No. JAEA-Data/Code 2017-015 (in Japanese).
22.
Iwata
,
K.
,
Tobita
,
T.
,
Takamizawa
,
H.
,
Chimi
,
Y.
,
Yoshimoto
,
K.
, and
Nishiyama
,
Y.
,
2016
, “
Specimen Size Effect on Fracture Toughness of Reactor Pressure Vessel Steel Following Warm Pre-Stressing
,”
ASME Paper No. PVP2016-63795
.10.1115/PVP2016-63795
23.
Natishan
,
M. E. N.
,
2001
, “
Materials Reliability Program (MRP) Establishing a Physically Based, Predictive Model for Fracture Toughness Transition Behavior of Ferritic Steels (MRP-53)
,”
EPRI
,
Palo Alto, CA
, EPRI Report No. 1003077.
24.
EricksonKirk
,
M.
,
Junge
,
M.
,
Arcieri
,
W.
,
Bass
,
B. R.
,
Beaton
,
R.
,
Bessette
,
D.
,
Chang
,
T. H.
,
Dickson
,
T.
,
Fletcher
,
C. D.
,
Kolaczkowski
,
A.
,
Malik
,
S.
,
Mintz
,
T.
,
Pugh
,
C.
,
Simonen
,
F.
,
Siu
,
N.
,
Whitehead
,
D.
,
Williams
,
P.
,
Woods
,
R.
and
Yin
,
S.
,
2006
, “
Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)
,”
United States Nuclear Regulatory Commission
,
Rockville, MD
, Report No. NUREG-1806.
25.
Raynaud
,
P.
,
Kirk
,
M.
,
Benson
,
M.
, and
Homiack
,
M.
,
2018
, “
Important Aspects of Probabilistic Fracture Mechanics Analyses
,”
Technical Letter, United States Nuclear Regulatory Commission
,
Rockville, MD
, Report No. TLR-RES/DE/CIB-2018-01.
26.
Lu
,
K.
,
Masaki
,
K.
,
Katsuyama
,
J.
, and
Li
,
Y.
,
2018
, “
Development of Crack Evaluation Models for Probabilistic Fracture Mechanics Analyses of Japanese Reactor Pressure Vessels
,”
ASME Paper No. PVP2018-84965
.10.1115/PVP2018-84965
27.
Cheverton
,
R. D.
,
Ball
,
D. G.
,
Bolt
,
S. E.
,
Iskander
,
S. K.
, and
Nanstad
,
R. K.
,
1985
, “
Pressure Vessel Fracture Studies Pertaining to the PWR Thermal-Shock Issue: Experiments TSE-5, TSE-5A, and TSE-6
,”
United States Nuclear Regulatory Commission
,
Rockville, MD
, Report No. NUREG/CR-4249 (ORNL-6163).
28.
Katsuyama
,
J.
,
Nishikawa
,
H.
,
Udagawa
,
M.
,
Nakamura
,
M.
, and
Onizawa
,
K.
,
2013
, “
Assessment of Residual Stress Due to Overlay-Welded Cladding and Structural Integrity of a Reactor Pressure Vessel
,”
ASME J. Pressure Vessel Technol.
,
135
(
5
), p.
051402
.10.1115/1.4024617
29.
Hirota
,
T.
,
Sakamoto
,
H.
, and
Ogawa
,
N.
,
2014
, “
Proposal for Update on Evaluation Procedure for Reactor Pressure Vessels Against Pressurized Thermal Shock Events in Japan
,”
ASME Paper No. PVP2014-28392
.10.1115/PVP2014-28392
30.
Zheng
,
X. J.
,
Kiciak
,
A.
, and
Glinka
,
G.
,
1997
, “
Weight Function and Stress Intensity Factors for Internal Surface Semi-Elliptical Crack in Thick-Walled Cylinder
,”
Eng. Fract. Mech.
,
58
(
3
), pp.
207
221
.10.1016/S0013-7944(97)00083-0
31.
Li
,
Y.
,
Hasegawa
,
K.
,
Xu
,
S.
, and
Scarth
,
D.
,
2014
, “
Weight Function Method With Segment-Wise Polynomial Interpolation to Calculate Stress Intensity Factors for Complicated Stress Distributions
,”
ASME J. Pressure Vessel Technol.
,
136
(
2
), p.
021202
.10.1115/1.4025816
32.
Chou
,
H. W.
, and
Huang
,
C. C.
,
2014
, “
Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events
,”
ASME Paper No. PVP2014-28350
.10.1115/PVP2014-28350
33.
The Japan Welding Society
,
1995
, “
Method of Crack Tip Opening Displacement (CTOD)
,”
The Japan Welding Society
,
Tokyo, Japan
, Report No. WES1108-1995 (in Japanese).
34.
Kanto
,
Y.
,
Jhung
,
M. J.
,
Ting
,
K.
,
He
,
Y. B.
,
Onizawa
,
K.
, and
Yoshimura
,
S.
,
2012
, “
Summary of International PFM Round Robin Analyses Among Asian Countries on Reactor Pressure Vessel Integrity During Pressurized Thermal Shock
,”
Int. J. Pressure Vessels Piping
,
90–91
, pp.
46
55
.10.1016/j.ijpvp.2011.10.007
35.
Simonen
,
F. A.
,
Doctor
,
S. R.
,
Schuster
,
G. J.
, and
Heasler
,
P. G.
,
2003
, “
A Generalized Procedure for Generating Flaw-Related Inputs for the FAVOR Code
,”
United States Nuclear Regulatory Commission
,
Rockville, MD
, Report No. NUREG/CR-6817.
You do not currently have access to this content.