Stress Corrosion Cracking in Welds of Reactor Vessel Nozzle at Ohi-3 and of Other Vessel’s Nozzle at Japan’s PWR Plants

2009 ◽  
Author(s):  
Takao Nakamura ◽  
Keiji Taniguchi ◽  
Shinro Hirano ◽  
Narita Marekazu ◽  
Tomonobu Sato
Keyword(s):  
Weld Metal ◽  
Grain Boundaries ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Visual Inspection ◽  
Reactor Vessel ◽  
Corrosion Cracking ◽  
Periodic Inspection ◽  
Depth Direction ◽  
Alloy 690

During the 13th periodic inspection, which started in February 2008, KEPCO’s Ohi unit 3 (1,180MWe PWR) implemented voluntary ECT in addition to the visual inspection of the RV hot and cold leg nozzle welds to confirm the integrity of the concerned section. As a result of inspection, in March 2008, a flaw extending in the depth direction along dendritic grain boundaries of the weld metal was found in the RV A-loop hot leg nozzle. With traces of machining, which could cause residual tensile stresses, it was suspected that SCC initiated and grew at the concerned section. After grinding the section to remove the entire flaw, WJP was applied as the corrective action. Ohi-3 restarted operation on November 2008. It is planned to apply repair welding to the ground out section with alloy 690 during the next periodic inspection [1]. Several Japan’s PWR plants have experienced similar incidents in the nozzle welds. This paper presents the details of repair technologies, which have been developed to address PWSCC found in Ohi-3 RV hot leg nozzle and previous similar incidents in the RV and other vessel’s nozzle welds at Japan’s PWR plants.

Mechanical Stress Improvement Process (MSIP) Used to Prevent and Mitigate Primary Water Stress Corrosion Cracking (PWSCC) in Reactor Vessel Piping at V.C. Summer

2003 ◽  
Author(s):  
E. A. Ray ◽  
K. Weir ◽  
C. Rice ◽  
T. Damico
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Mechanical Stress ◽  
Stress Corrosion Cracking ◽  
Reactor Vessel ◽  
Corrosion Cracking ◽  
The Other ◽  
Water Reactor ◽  
Improvement Process ◽  
Primary Water

During the October 2000 refueling outage at the V.C. Summer Nuclear Station, a leak was discovered in one of the three reactor vessel hot leg nozzle to pipe weld connections. The root cause of this leak was determined to be extensive weld repairs causing high tensile stresses throughout the pipe weld; leading to primary water stress corrosion cracking (PWSCC) of the Alloy 82/182 (Inconel). This nozzle was repaired and V.C. Summer began investigating other mitigative or repair techniques on the other nozzles. During the next refueling outage V.C. Summer took mitigative actions by applying the patented Mechanical Stress Improvement Process (MSIP) to the other hot legs. MSIP contracts the pipe on one side of the weldment, placing the inner region of the weld into compression. This is an effective means to prevent and mitigate PWSCC. Analyses were performed to determine the redistribution of residual stresses, amount of strain in the region of application, reactor coolant piping loads and stresses, and effect on equipment supports. In May 2002, using a newly designed 34-inch clamp, MSIP was successfully applied to the two hot-leg nozzle weldments. The pre- and post-MSIP NDE results were highly favorable. MSIP has been used extensively on piping in boiling water reactor (BWR) plants to successfully prevent and mitigate SCC. This includes Reactor Vessel nozzle piping over 30-inch diameter with 2.3-inch wall thickness similar in both size and materials to piping in pressurized water reactor (PWR) plants such as V.C. Summer. The application of MSIP at V.C. Summer was successfully completed and showed the process to be predictable with no significant changes in the overall operation of the plant. The pre- and post-nondestructive examination of the reactor vessel nozzle weldment showed no detrimental effects on the weldment due to the MSIP.

Weld Residual Stresses and Primary Water Stress Corrosion Cracking in Bimetal Nuclear Pipe Welds

2007 ◽  
Author(s):  
Frederick W. Brust ◽  
Paul M. Scott
Keyword(s):  
Water Stress ◽  
Residual Stresses ◽  
Weld Metal ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Pressure Vessel ◽  
Large Diameter ◽  
Corrosion Cracking ◽  
Primary Water

There have been incidents recently where cracking has been observed in the bi-metallic welds that join the hot leg to the reactor pressure vessel nozzle. The hot leg pipes are typically large diameter, thick wall pipes. Typically, an inconel weld metal is used to join the ferritic pressure vessel steel to the stainless steel pipe. The cracking, mainly confined to the inconel weld metal, is caused by corrosion mechanisms. Tensile weld residual stresses, in addition to service loads, contribute to PWSCC (Primary Water Stress Corrosion Cracking) crack growth. In addition to the large diameter hot leg pipe, cracking in other piping components of different sizes has been observed. For instance, surge lines and spray line cracking has been observed that has been attributed to this degradation mechanism. Here we present some models which are used to predict the PWSCC behavior in nuclear piping. This includes weld model solutions of bimetal pipe welds along with an example calculation of PWSCC crack growth in a hot leg. Risk based considerations are also discussed.

On the role of intergranular nanocavities in long-term stress corrosion cracking of Alloy 690

2021 ◽  
pp. 117453
Author(s):  
Zhao Shen ◽  
Edward Roberts ◽  
Naganand Saravanan ◽  
Phani Karamched ◽  
Takumi Terachi ◽  
...  
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Alloy 690

Developing Response Functions for Crack Opening Displacement to Calculate Leak-Rates Through Complex Cracks in Pipes With Weld Overlays

2010 ◽  
Author(s):  
Arindam Chakraborty ◽  
Wasimreza Momin ◽  
Angah Miessi ◽  
Peihua Jing ◽  
Haiyang Qian
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Crack Opening ◽  
Corrosion Cracking ◽  
Flaw Size ◽  
Pipe Material ◽  
Response Functions ◽  
Plant Design ◽  
Alloy 690 ◽  
Complex Crack

Leak-Before-Break (LBB) is employed in design of nuclear power reactor piping to eliminate consideration of the dynamic effects of pipe rupture from the plant design basis for the affected piping system. LBB cannot be applied if environmental conditions that could lead to degradation by stress corrosion cracking exists. For Alloy 600/82/182 dissimilar metal welds (DMW) in pressurized water reactor plants, primary water stress corrosion cracking (PWSCC) is found to be active. Application of weld overlay (WOL) of non-susceptible Alloy 690/52/152 material has been shown to mitigate PWSCC growth in DMW. Therefore, LBB can be considered for a DMW with Alloy 690/52/152 overlay. However, WOL sizing design postulates a complex crack which is through wall in the overlay material and part through or full circumferential in the DMW base material. This significantly reduces the critical flaw size and in turn the maximum allowable flaw size for leak rate. The current industry practice conservatively ignores the full circumferential crack in the original pipe material and assumes a through wall crack along the entire pipe thickness. This assumptions leads to significantly reduced leakage due to smaller crack opening. The problem becomes more critical with small diameter pipes. The current work calculates the crack opening displacements (CODs) for a pipe with complex crack. Since it is a function of several geometry and materials parameters, response functions are generated to calculate CODs.

Review of Ferrite Number (FN) Requirements and Proposed Changes to Code Case N-504-4 and Nonmandatory Appendix Q

2019 ◽  
Author(s):  
Steven L. McCracken ◽  
David Segletes
Keyword(s):  
Stainless Steel ◽  
Weld Metal ◽  
Carbon Content ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Corrosion Cracking ◽  
Delta Ferrite ◽  
Ferrite Number

Abstract ASME Section XI Nonmandatory Appendix Q and Code Case N-504-4 are routinely used to install full structural weld overlays in the nuclear power industry for repair or mitigation of stress corrosion cracking in austenitic stainless steel weldments. Both Appendix Q and N-504-4 specify a Ferrite Number (FN) and carbon content requirement for the stainless steel weld metal used for the weld overlay to ensure acceptable resistance to stress corrosion cracking. The Ferrite Number (FN) is used in the ASME Code for establishing the delta ferrite content in the deposited weld metal. Field experience indicates there is often confusion and differing opinion concerning how the Ferrite Number and carbon content requirements of Appendix Q and N-504-4 are satisfied. This is in part due to unavailability of the original technical basis for these requirements. This paper provides a background for the delta ferrite and carbon content requirements, information on influence of delta ferrite and carbon content on stress corrosion cracking and U.S. Nuclear Regulatory Commission (NRC) guidance on the issue. Finally, this paper details a proposed revision of Nonmandatory Appendix Q and N-504-4 to clarify the FN and carbon content requirements.

Sign in / Sign up

Export Citation Format

Share Document