Search


Information
ICMST-Tohoku 2018
Oct. 23 - 26, 2018
Sendai, Japan
ICMST-Shenzhen 2016
Nov 1 - 4, 2016
Shenzhen, China
(EXIT THIS PAGE)
ICMST-Kobe 2014
Nov 2(Sun) - 5(Wed), 2014
Kobe, Japan
Nuclear Regulation Authority Outline of the New Safety Standards for Light Water Reactors for Electric Power Generation
For Public Comment
Outline of New Safety Standard (Design Basis)
For Public Comment
New Safety Standards (SA) Outline (Draft)
For Public Comment
Outline of New Safety Standard(Earthquake and Tsunami)(DRAFT)
Issues
 

Vol.10 No.2(Aug)
Vol.10 No.1(May)
Vol.9 No.4(Feb)
Vol.9 No.3(Nov)

< Other Issues

 

Occasional Topics
OTjapan Measures for Tsunami Striking Nuclear Power Station in Japan
Special Article: The Great Tohoku Earthquake (1)
OTjapan The Tragedy of “To Be” Principle in the Japanese Nuclear Industry
EJAMOT_CN3_Figure1_The_outside_view_of_CEFR OTChinaPlanning and Consideration on SFR R&D Activities in China
< All Occasional Topics

Featured Articles
EJAM7-3NT72 A New Mechanical Condition-based Maintenance Technology Using Instrumented Indentation Technique
EJAM7-3NT73 Survey robots for Fukushima Daiichi Nuclear Power Plant

JSM
Contacts
(EJAM): ejam@jsm.or.jp
(JSM): secretariat@jsm.or.jp
HP: http://www.jsm.or.jp
(in English)

 

Vol.11 No.4previousAA SP22 (AA165-166-167-168-169-170-171) NT96

Academic Articles
Regular Paper Vol. 11 No. 4 (2020) p.124 - p131
 

Verification Benchmark Analysis of Structural Reliability Evaluation Codes for Fast Reactor Components

 

Shigeru TAKAYA1,*, Naoto SASAKI2, Toshiki OKAMOTO3 and Hideo MACHIDA3

 
1 Japan Atomic Energy Agency, 4002 Narita, Oarai, Ibaraki 311-1393, Japan
2 Ascend Co., Ltd., 4002 Narita, Oarai, Ibaraki 311-1313, Japan
3 TEPCO Systems Corporation, 2-37-28 Eitai, Koto-ku, Tokyo 135-0034, Japan

 
Abstract
This paper describes benchmark analysis of independently programmed structural reliability evaluation codes, REAL-P and GENPEP. An upper core structure of a prototype fast breeder reactor in Japan, MONJU, was chosen, and crack initiation time and crack propagation due to fatigue-creep interaction damage was evaluated in deterministic and probabilistic manners. Evaluation procedures follow the new guidelines on reliability evaluation of fast reactor components issued by JSME. The results obtained by two codes were compared, and the effects of differences in treatments of which details are not prescribed in the guidelines on results were discussed. As result, although slight difference was recognized in crack initiation evaluation especially due to difference in fairing treatment of fatigue life curves, the results estimated by two codes generally agreed very well for both deterministic and probabilistic evaluations. It was shown that the effects of differences in treatments of which details are not prescribed in the guidelines on results are small for structural reliability evaluation of fatigue-creep interaction damage, which is one of typical degradation mechanisms for fast reactor passive components.
 
Keywords
System Based Code concept, Inservice inspection, Fatigue-creep interaction damage, Crack initiation, Crack propagation, Deterministic evaluation, Probabilistic evaluation, Plant safety
 
Full Paper: PDF
Article Information
Article history:
Received 22 October 2018
Accepted 04 February 2020