Search

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

Information
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.5 No.4(Feb-25)
Vol.5 No.3(Nov-25)
Vol.5 No.2(Aug-26)
Vol.5 No.1(May-27)
< Other Issues

Featured Articles
EJAM5-3NT58 "Development of Future Prediction of Animation by coupling of Principle Component Analysis and Singular Spectrum Analysis"
EJAM5-3NT58 Validation of Multifrequency and Window-Multifrequency Techniques using Remote Field Eddy Currents for ISI of FBR Tubes
EJAM5-3NT59 Remote Controlled Robot with expandable features

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

EJAM Vol.2 pp.50-64 "Evaluation of Weld Residual Stress near the Cladding and J-weld in Reactor Pressure Vessel Head for the assessment of PWSCC Behavior"
 
Vol.2 No.1 previous GA 13 - AA 15 - SP2 (16 - 17 - 18) - NT 23 - 24 - 25 next Vol.2 No.3
Academic Articles
Regular Paper (Invited) Vol.2 (2010) p.50 - p.64
 

Evaluation of Weld Residual Stress near the Cladding and J-weld in Reactor Pressure Vessel Head for the assessment of PWSCC Behavior

 
Jinya KATSUYAMA1,*, Makoto UDAGAWA1, Hiroyuki NISHIKAWA1,†, Mitsuyuki NAKAMURA1,‡ and Kunio ONIZAWA1
 
1Nuclear Safety Research Center, Japan Atomic Energy Agency, 2-4 Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki 319-1195, Japan
 
Abstract
Weld residual stress is one of the most important factors in order to assess the structural integrity of safety-related components such as reactor pressure vessel (RPV) in long-term operation of nuclear power plant since the residual stress significantly affects crack initiation and growth behaviors. The inner surface of the RPV made of low alloy steel is protected against corrosion by weld-overlay cladding of austenitic stainless steel. At the J-weld of the vessel head penetrations, Ni-based alloys are used for weld material. The residual stresses generated within the cladding, J-weld and base material were measured as-welded and after PWHT conditions using the deep-hole-drilling method. Thermal-elastic-plastic-creep analyses considering the phase transformation were also performed to evaluate the weld residual stress. By comparing analytical results with the measured ones, it was shown that there was roughly a good agreement of residual stress distribution within the cladding, J-weld and base metal. It was also suggested that taking the phase transformation during welding and PWHT into account was important to improve the accuracy of weld residual stress analysis. Using the residual stress distributions, fracture mechanics analyses for primary water stress corrosion cracking (PWSCC) have been performed using probabilistic fracture mechanics analysis code. Effects of the weld residual stress and scatter of PWSCC growth rate on the crack penetration were evaluated through some case studies.
 
Keywords
weld residual stress, reactor pressure vessel, PWSCC, weld-overlay cladding, J-groove welding, deep-hole-drilling method, finite element analysis, structure integrity, crack growth
 
Full Paper: PDF
Article Information
Article history:
Received 20 May 2010
Accepted 21 July 2010