Both deterministic and probabilistic methods are used to assess the integrity of a reactor pressure vessel (RPV) subjected to pressurized thermal shocks (PTSs). The FAVOR code is applied to calculate the probabilities for crack initiation and failure of the RPV subjected to two transients, by considering crack distributions based on cracks observed in the Shoreham and pressure vessel research user facility (PVRUF) RPVs. The crack parameters, i.e., crack density, depth, aspect ratio, orientation, and location are assumed as random variables following different distributions. KI of the cracks with the same depth increases with its aspect ratio. Both KI and KIc at the crack tip increase with crack depth, which is the reason why a deeper crack does not necessarily lead to a higher failure probability. The underclad crack is the most critical crack and the deeper crack is the least critical one in this study. Considering uncertainties of the transients results in higher failure probabilities.
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December 2015
Research-Article
Probabilistic Pressurized Thermal Shocks Analyses for a Reactor Pressure Vessel
Guian Qian,
Guian Qian
1
Laboratory for Nuclear Materials,
Nuclear Energy and Safety Research Department,
OHSA/06,
e-mail: guian.qian@psi.ch
Nuclear Energy and Safety Research Department,
Paul Scherrer Institute
,OHSA/06,
5232 Villigen PSI
, Switzerland
e-mail: guian.qian@psi.ch
1Corresponding author.
Search for other works by this author on:
Markus Niffenegger
Markus Niffenegger
Laboratory for Nuclear Materials,
Nuclear Energy and Safety Research Department,
OHSA/06,
Nuclear Energy and Safety Research Department,
Paul Scherrer Institute
,OHSA/06,
5232 Villigen PSI
, Switzerland
Search for other works by this author on:
Guian Qian
Laboratory for Nuclear Materials,
Nuclear Energy and Safety Research Department,
OHSA/06,
e-mail: guian.qian@psi.ch
Nuclear Energy and Safety Research Department,
Paul Scherrer Institute
,OHSA/06,
5232 Villigen PSI
, Switzerland
e-mail: guian.qian@psi.ch
Markus Niffenegger
Laboratory for Nuclear Materials,
Nuclear Energy and Safety Research Department,
OHSA/06,
Nuclear Energy and Safety Research Department,
Paul Scherrer Institute
,OHSA/06,
5232 Villigen PSI
, Switzerland
1Corresponding author.
Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received August 6, 2014; final manuscript received March 31, 2015; published online May 20, 2015. Assoc. Editor: Kunio Hasegawa.
J. Pressure Vessel Technol. Dec 2015, 137(6): 061206 (7 pages)
Published Online: December 1, 2015
Article history
Received:
August 6, 2014
Revision Received:
March 31, 2015
Online:
May 20, 2015
Citation
Qian, G., and Niffenegger, M. (December 1, 2015). "Probabilistic Pressurized Thermal Shocks Analyses for a Reactor Pressure Vessel." ASME. J. Pressure Vessel Technol. December 2015; 137(6): 061206. https://doi.org/10.1115/1.4030299
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