The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead an RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, three-dimensional computational fluid dynamics (3D-CFD) and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a practically more useful spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of conditional failure probabilities on the position inside the RPV is obtained. Using the spatial distribution of conditional failure probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

References

References
1.
Odette
,
G. R.
, and
Lucas
,
G. E.
,
1986
, “
Irradiation Embrittlement of Reactor Pressure Vessel Steels: Mechanisms, Models and Data Correlations, Radiation Embrittlement of Reactor Pressure Vessel Steels—An International Review L.E. Steele (Ed.)
,” ASTM International, Philadelphia, PA, Standard No. ASTM STP 909.
2.
SCIENTECH Inc.
,
1999
, “
RELAP5/Mod3 Code Manual, Vol. I: Code Structure, System Models and Solution Methods
,” The Thermal Hydraulics Group, ID.
3.
Smith
,
B. L.
,
2010
, “
Assessment of CFD Codes Used in Nuclear Reactor Safety Simulations
,”
Nucl. Eng. Technol.
,
42
(
4
), pp.
339
364
.
4.
González-Albuixech
,
V. F.
,
Qian
,
G.
,
Sharabi
,
M.
,
Niffenegger
,
M.
,
Niceno
,
B.
, and
Lafferty
,
N.
,
2015
, “
Comparison of PTS Analyses of RPVs Based on 3D-CFD and RELAP5
,”
Nucl. Eng. Des.
,
291
, pp.
168
178
.
5.
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
.
6.
González-Albuixech
,
V. F.
,
Qian
,
G.
,
Sharabi
,
M.
,
Niffenegger
,
M.
,
Niceno
,
B.
, and
Lafferty
,
N.
,
2016
, “
Coupled RELAP5, 3D CFD and FEM Analysis of Postulated Cracks in RPVs Subjected to PTS Loading
,”
Nucl. Eng. Des.
,
297
, pp.
111
122
.
7.
Qian
,
G.
, and
Niffenegger
,
M.
,
2013
, “
Integrity Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks by Considering Constraint Effect
,”
Eng. Fract. Mech.
,
112–113
, pp.
14
25
.
8.
Dickson
,
T. L.
,
Williams
,
P. T.
, and
Yin
,
S.
,
2007
, “
Fracture Analysis of Vessels-Oak Ridge FAVOR, v 06.1: Computer Code: User's Guide
,” U.S. Nuclear Regulatory Commission, Washington, DC, No. NUREG-ORNL/TM-2007/031.
9.
Onizawa
,
K.
,
Nishikawa
,
H.
, and
Itoh
,
H.
,
2010
, “
Development of Probabilistic Fracture Mechanics Analysis Codes for Reactor Pressure Vessels and Piping Considering Welding Residual Stress
,”
Int. J. Pressure Vessels Piping
,
87
(
1
), pp.
2
10
.
10.
Qian
,
G.
, and
González-Albuixech
,
V. F.
,
2013
, “
Procedures, Methods and Computer Codes for Probabilistic Assessment of Reactor Pressure Vessels Subjected to Pressurized Thermal Shocks
,”
Nucl. Eng. Des.
,
258
, pp.
35
50
.
11.
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
.
12.
Qian
,
G.
,
González-Albuixech
,
V. F.
,
Niffenegger
,
M.
, and
Sharabi
,
M.
,
2016
, “
Probabilistic Pressurized Thermal Shock Analysis for a Reactor Pressure Vessel Considering Plume Cooling Effect
,”
ASME J. Pressure Vessel Technol.
,
138
(
4
), p.
041204
.
13.
Qian
,
G.
,
Lei
,
W. S.
,
Niffenegger
,
M.
, and
González-Albuixech
,
V. F.
,
2018
, “
On the Temperature Independence of Statistical Model Parameters for Cleavage Fracture in Ferritic Steels
,”
Philos. Mag.
,
98
(
11
), pp.
959
1004
.
14.
Qian
,
G.
,
Lei
,
W. S.
,
Peng
,
L.
,
Yu
,
Z.
, and
Niffenegger
,
M.
,
2018
, “
Statistical Assessment of Notch Toughness Against Cleavage Fracture of Ferritic Steels
,”
Fatigue Fract. Eng. Mater. Struct.
,
41
(
5
), pp.
1120
1131
.
15.
Qian
,
G.
,
Cao
,
Y.
,
Niffenegger
,
M.
,
Chao
,
Y. J.
, and
Wu
,
W.
,
2018
, “
Comparison of Constraint Analyses With Global and Local Approaches Under Uniaxial and Biaxial Loadings
,”
Eur. J. Mech. A
,
69
, pp.
135
146
.
16.
IAEA, 2016, “
Report of the Integrated Regulatory Review Service (IRRS) Mission to Japan
,” International Atomic Energy Agency, Tokyo, Japan, Report No. IAEA-NS-IRRS-2016.
17.
ASTM,
1997
, “
Test Method for Determination of Reference Temperature, T0, for Ferritic Steels in the Transition Range
,” American Society for Testing and Materials, New York, Standard No.
ASTM-E1921-02
.
18.
U.S. NRC
,
2007
, “
Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)
,” Vol.
1
, U.S. Nuclear Regulatory Commission, Washington, DC, Report No.
NUREG-1806
.https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1806/v1/index.html
19.
ANSYS
,
2016
, “
PDF Documentation for Release 17.2
,” ANSYS Inc., Canonsburg, PA.
20.
Mahaffy
,
J.
,
Chung
,
B.
,
Song
,
C.
,
Dubois
,
F.
,
Graffard
,
E.
,
Ducros
,
F.
,
Heitsch
,
M.
,
Scheuerer
,
M.
,
Henriksson
,
M.
,
Komen
,
E.
,
Moretti
,
F.
,
Morii
,
T.
,
Muehlbauer
,
P.
,
Rohde
,
U.
,
Smith
,
B. L.
,
Watanabe
,
T.
, and
Zigh
,
G.
,
2007
, “
Best Practice Guidelines for the Use of CFD in Nuclear Reactor Safety Applications
,” Organisation for Economic Co-operation and Development/Nuclear Energy Agency/Committee on the Safety of Nuclear Installations, Paris, France, No. NEA/CSNI/R(2007)5.
21.
Ruan
,
X.
,
Nakasuji
,
T.
, and
Morishita
,
K.
,
2017
, “
Coupled 3D CFD and FEM Assessments of RPV Stress Intensity Factor During PTS Events
,”
2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017)
, Fukui and Kyoto, Japan, Apr. 24–28, Paper No. 17136.
22.
IAEA
,
2010
, “
Pressurized Thermal Shock in Nuclear Power Plants: Good Practices for Assessment
,” International Atomic Energy Agency, Vienna, Austria, Report No.
IAEA-TECDOC-1627
.https://www-pub.iaea.org/MTCD/Publications/PDF/te_1627_web.pdf
23.
Ruan
,
X.
,
Nakasuji
,
T.
, and
Morishita
,
K.
,
2017
, “
3D CFD and FEM Evaluations of RPV Stress Intensity Factor During PTS Loading
,”
E-J. Adv. Maint.
,
9
(
2
), pp.
84
90
.http://www.jsm.or.jp/ejam/Vol.9No.2/AA/AA133/AA133.pdf
24.
Stephens
,
R. I.
,
Fatemi
,
A.
,
Stephens
,
R. R.
, and
Fuchs
,
H. O.
,
2001
,
Metal Fatigue in Engineering
, 2nd ed.,
Wiley
,
New York
, p.
134
.
25.
ASME
,
1995
, “
ASME Boiler and Pressure Vessel Code, Section III
,”
Nuclear Power Plant Components
,
American Society of Mechanical Engineers
,
New York
.
26.
Qian
,
G.
, and
Niffenegger
,
M.
,
2014
, “
Deterministic and Probabilistic Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks
,”
Nucl. Eng. Des.
,
273
, pp.
381
395
.
27.
Simonen
,
F. A.
,
Doctor
,
S. R.
,
Schuster
,
G. J.
, and
Heasler
,
P. G.
,
2004
, “
A Generalized Procedure for Generating Flaw-Related Inputs for the FAVOR Code
,” U.S. Nuclear Regulatory Commission, Washington, DC, Report No.
NUREG/CR-6817
.https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6817/
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