2191 R0 XI.M4

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XI.M4 BWR VESSEL ID ATTACHMENT WELDS

Program Description

This program is a condition monitoring program for detecting cracking due to stress corrosion cracking (SCC), intergranular stress corrosion cracking (IGSCC), and cyclical loading mechanisms in the reactor vessel inside diameter (ID) attachment welds of boiling water reactors (BWRs). The program includes inspection and flaw evaluation in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, and the guidance in “BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines” (Boiling Water Reactor Vessel and Internals Project (BWRVIP)-48-A) to provide reasonable assurance of the long-term integrity and safe operation of BWR vessel ID attachment welds.

The guidance in BWRVIP-48-A includes inspection recommendations and evaluation methodologies for certain attachment welds between the vessel wall and the brackets that attach components to the vessel. In some cases, the attachment is a weld attached directly to the vessel wall; in other cases, the attachment includes a weld build-up pad on the vessel wall. The BWRVIP-48-A report includes information on the geometry of the vessel ID attachments; evaluates susceptible locations and the safety consequence of failure; provides recommendations regarding the method, extent, and frequency of augmented examinations; and discusses acceptable methods for evaluating the structural integrity significance of indications detected during examinations.

Evaluation and Technical Basis

1. Scope of Program: This program manages the effects of cracking caused by SCC, IGSCC, or cyclical loading mechanisms for those BWR vessel ID attachment welds that are covered by BWRVIP-48-A. The program is an augmented inservice inspection (ISI) program that uses the inspection and flaw evaluation criteria in BWRVIP-48-A to detect cracking and monitor the effects of cracking on the intended functions of these components.

2. Preventive Actions: This program is a condition monitoring program and has no preventive actions. To mitigate SCC and IGSCC, reactor coolant water chemistry is monitored and controlled in accordance with activities that meet the guidelines in https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr2191/r0/index.html Generic Aging Lessons Learned for Subsequent License Renewal] (GALL-SLR) Report aging management program (AMP) XI.M2, “Water Chemistry.”

3. Parameters Monitored or Inspected: The program monitors for cracks caused by SCC, IGSCC, and cyclical loading mechanisms. Inspections performed in accordance with the guidance in BWRVIP-48-A and the requirements of the ASME Code, Section XI, Table IWB-2500-1 are used to interrogate the components for discontinuities that may indicate the presence of cracking.

4. Detection of Aging Effects: The extent and schedule of the inspections prescribed by BWRVIP-48-A and ASME Code, Section XI, are designed to maintain structural integrity, to discover aging effects, and to repair or replace the component before a loss of intended function. The vessel ID attachment welds are visually examined in accordance with the requirements of ASME Code, Section XI, Table IWB-2500-1, Examination Category B-N-2. The inspection and evaluation guidelines of BWRVIP-48-A recommend more stringent inspections for certain attachment welds. The nondestructive examination techniques that are appropriate for the augmented examinations, including the uncertainties inherent in delivering and executing these techniques and applicable for inclusion in flaw evaluations, are included in BWRVIP-03(Archived).

5. Monitoring and Trending: Inspections scheduled in accordance with ASME Code, Section XI, Subarticle IWB-2400, and BWRVIP-48-A provide for the timely detection of cracking. If indications are detected, the scope of examination is expanded. Any indications are evaluated in accordance with ASME Code, Section XI, and the guidance in BWRVIP-48-A. Guidance for the evaluation of crack growth in stainless steels, nickel alloys, and low-alloy steels is provided in BWRVIP-14-A, BWRVIP-59-A, and BWRVIP-60-A, respectively.

6. Acceptance Criteria: The relevant acceptance criteria are provided in BWRVIP-48-A and ASME Code, Section XI, Subarticle IWB-3520.

7. Corrective Actions: Results that do not meet the acceptance criteria are addressed in the applicant’s corrective action program under those specific portions of the quality assurance (QA) program that are used to meet Criterion XVI, “Corrective Action,” of [Title 10 of the Code of Federal Regulation (10 CFR) Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the corrective actions element of this AMP for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

Repair and replacement activities are conducted in accordance with the guidance in BWRVIP-52-A.

8. Confirmation Process: The confirmation process is addressed through those specific portions of the QA program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the confirmation process element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

9. Administrative Controls: Administrative controls are addressed through the QA program that is used to meet the requirements of 10 CFR Part 50, Appendix B, associated with managing the effects of aging. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the administrative controls element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

10. Operating Experience: Cracking due to SCC, IGSCC, and cyclical loading has occurred in BWR components. The program guidelines are based on an evaluation of available information, including BWR inspection data and information on the causes of SCC, IGSCC, and cracking due to cyclical loading, to determine which attachment welds may be susceptible to cracking from any of these mechanisms. Implementation of this program provides reasonable assurance that cracking will be adequately managed and that the intended functions of the vessel ID attachments will be maintained consistent with the current licensing basis for the subsequent period of extended operation.

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry operating experience including research and development such that the effectiveness of the AMP is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report.

References

10 CFR Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants." Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR 50.55a, “Codes and Standards.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

ASME. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” New York, New York: The American Society of Mechanical Engineers. 2008.

EPRI. BWRVIP-03, Revision 6 (EPRI 105696-R6)(Archived), “BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines.” Palo Alto, California: Electric Power Research Institute. December 2003.

_____. BWRVIP-14-A (EPRI 1016569), “Evaluation of Crack Growth in BWR Stainless Steel RPV Internals.” Palo Alto, California: Electric Power Research Institute. September 2008.

_____. BWRVIP-48-A (EPRI 1009948), “BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. November 2004.

_____. BWRVIP-52-A (EPRI 1012119), “BWR Vessel and Internals Project, Shroud Support and Vessel Bracket Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. September 2005.

_____. BWRVIP-59-A (EPRI 1014874), “Evaluation of Crack Growth in BWR Nickel-Base Austenitic Alloys in RPV Internals.” Palo Alto, California: Electric Power Research Institute. May 2007.

_____. BWRVIP-60-A (EPRI 1008871), “BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Low Alloy Steel RPV Internals.” Palo Alto, California: Electric Power Research Institute. June 2003.

Boiling Water Reactor

Stress Corrosion Cracking

American Society of Mechanical Engineers

Generic Aging Lessons Learned - The US NRC GALL Report is NUREG-1801. The US NRC Subsequent License Renewal GALL is NUREG-2191

Subsequent License Renewal - The extension of the original license from 40 years to 60 years is known as License Renewal. Extensions of the license beyond 60 years are known as Subsequent License Renewal.

Boiling Water Reactor Vessel Internals Program - The BWRVIP is an industry program for General Electric boiling water reactors that has developed procedures for performing inspections of reactor vessel internals.

Quality Assurance

Aging Management Program

Electric Power Research Institute