To apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), a PFM analysis code has been developed at JAEA. Using this PFM analysis code, pascal version 3, the conditional probabilities of crack initiation (CPIs) and fracture for an RPV during pressurized thermal shock (PTS) events have been analyzed. Sensitivity analyses on certain input parameters were performed to clarify their effect on the conditional fracture probability. Comparisons between the conditional probabilities and the temperature margin (ΔTm) based on the current deterministic analysis method were made for various model plant conditions for typical domestic older types of RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.
Skip Nav Destination
Article navigation
February 2014
Research-Article
Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During Pressurized Thermal Shock Events
Koichi Masaki,
Koichi Masaki
1
1Present address: Mizuho Information & Research Institute, Inc., 2-3 Kanda Nishiki-cho, Chiyoda-ku, Tokyo 101-8443, Japan.
Search for other works by this author on:
Jinya Katsuyama,
Kunio Onizawa
Kunio Onizawa
Nuclear Safety Research Center,
Japan Atomic Energy Agency
,2-4 Shirakata-shirane, Tokai-mura
,Naka-gun, Ibaraki 319-1195
Japan
Search for other works by this author on:
Jinya Katsuyama
e-mail: katsuyama.jinya@jaea.go.jp
Kunio Onizawa
Nuclear Safety Research Center,
Japan Atomic Energy Agency
,2-4 Shirakata-shirane, Tokai-mura
,Naka-gun, Ibaraki 319-1195
Japan
1Present address: Mizuho Information & Research Institute, Inc., 2-3 Kanda Nishiki-cho, Chiyoda-ku, Tokyo 101-8443, Japan.
Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received February 13, 2013; final manuscript received August 2, 2013; published online November 27, 2013. Assoc. Editor: David L. Rudland.
J. Pressure Vessel Technol. Feb 2014, 136(1): 011208 (7 pages)
Published Online: November 27, 2013
Article history
Received:
February 13, 2013
Revision Received:
August 2, 2013
Citation
Masaki, K., Katsuyama, J., and Onizawa, K. (November 27, 2013). "Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During Pressurized Thermal Shock Events." ASME. J. Pressure Vessel Technol. February 2014; 136(1): 011208. https://doi.org/10.1115/1.4025615
Download citation file:
66
Views
Get Email Alerts
Cited By
Measurement of Steam-Generator-Tube Vibration Damping Caused by Anti-Vibration-Bar Supports
J. Pressure Vessel Technol (February 2025)
Related Articles
Guideline on Probabilistic Fracture Mechanics Analysis for Japanese Reactor Pressure Vessels
J. Pressure Vessel Technol (April,2020)
Verification of Probabilistic Fracture Mechanics Analysis Code for Reactor Pressure Vessel
J. Pressure Vessel Technol (August,2021)
Investigation on Constraint Effect of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shocks
J. Pressure Vessel Technol (February,2015)
Application of Probabilistic Fracture Mechanics to Reactor Pressure Vessel Using PASCAL4 Code
J. Pressure Vessel Technol (April,2021)
Related Proceedings Papers
Related Chapters
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Long.Term Reactivity Change and Control: On.Power Refuelling
Fundamentals of CANDU Reactor Physics
A Study of Irradiation-Induced Growth of Modified and Advanced Zr-Nb System Alloys after Irradiation in the VVER-1000 Reactor Core at Temelin NPP
Zirconium in the Nuclear Industry: 20th International Symposium