1801 R0 XI.M7

Return to AMP Table

XI.M7 BWR STRESS CORROSION CRACKING

Program Description

The program to manage intergranular stress corrosion cracking (IGSCC) in boiling water reactor (BWR) coolant pressure boundary piping made of stainless steel (SS) is delineated in NUREG-0313, Rev. 2, and US Nuclear Regulatory Commission (US NRC) Generic Letter (GL) 88-01 and its Supplement 1. The program includes (a) preventive measures to mitigate IGSCC, and (b) inspection and flaw evaluation to monitor IGSCC and its effects. The staff-approved boiling water reactor vessel and internals project (BWRVIP)-75(Archived) report allows for modifications of inspection scope in the GL 88-01 program.

Evaluation and Technical Basis

1. Scope of Program: The program focuses on (a) managing and implementing countermeasures to mitigate IGSCC and (b) performing inservice inspection (ISI) to monitor IGSCC and its effects on the intended function of BWR components. The program is applicable to all BWR piping made of austenitic SS that is 4 in. or larger in nominal diameter and contains reactor coolant at a temperature above 930C (2000F) during power operation, regardless of code classification. The program also applies to pump casings, valve bodies and reactor vessel attachments and appurtenances, such as head spray and vent components. NUREG-0313 and US NRC GL 88-01, respectively, describe the technical basis and staff guidance regarding mitigation of IGSCC in BWRs. Attachment A of US NRC GL 88-01 delineates the staff-approved positions regarding materials, processes, water chemistry, weld overlay reinforcement, partial replacement, stress improvement of cracked welds, clamping devices, crack characterization and repair criteria, inspection methods and personnel, inspection schedules, sample expansion, leakage detection, and reporting requirements.

2. Preventive Actions: The comprehensive program outlined in NUREG-0313 and US NRC GL 88-01 addresses improvements in all three elements that, in combination, cause IGSCC. These elements consist of a susceptible (sensitized) material, a significant tensile stress, and an aggressive environment. Sensitization of nonstabilized austenitic SSs containing greater than 0.03 wt.% carbon involves precipitation of chromium carbides at the grain boundaries during certain fabrication or welding processes. The formation of carbides creates an envelope of chromium depleted region that, in certain environments, is susceptible to stress corrosion cracking (SCC). Residual tensile stresses are introduced from fabrication processes, such as welding, surface grinding, or forming. High levels of dissolved oxygen or aggressive contaminants, such as sulfates or chlorides, accelerate the SCC processes.

The program delineated in NUREG-0313 and US NRC GL 88-01 and in the staff-approved BWRVIP-75(Archived) report includes recommendations regarding selection of materials that are resistant to sensitization, use of special processes that reduce residual tensile stresses, and monitoring and maintenance of coolant chemistry. The resistant materials are used for new and replacement components and include low-carbon grades of austenitic SS and weld metal, with a maximum carbon of 0.035 wt.% and a minimum ferrite of 7.5% in weld metal and cast austenitic stainless steel (CASS). Inconel 82 is the only commonly used nickelbase weld metal considered to be resistant to SCC; other nickel-alloys, such as Alloy 600 are evaluated on an individual basis. Special processes are used for existing, new, and replacement components. These processes include solution heat treatment, heat sink welding, induction heating, and mechanical stress improvement.

The program delineated in NUREG-0313 and US NRC GL 88-01 does not provide specific guidelines for controlling reactor water chemistry to mitigate IGSCC. Maintaining high water purity reduces susceptibility to SCC or IGSCC. Reactor coolant water chemistry is monitored and maintained in accordance with the guidelines in BWRVIP-29 (Electric Power Research Institute [EPRI] TR-103515)(Archived). The program description, and evaluation and technical basis of monitoring and maintaining reactor water chemistry are presented in Section XI.M2, 'Water Chemistry."

3. Parameters Monitored/Inspected: The program detects and sizes cracks and detects leakage by using the examination and inspection guidelines delineated in NUREG-0313, Rev. 2, and US NRC GL 88-01 or the referenced BWRVIP-75(Archived) guideline as approved by the US NRC staff.

4. Detection of Aging Effects: The extent, method, and schedule of the inspection and test techniques delineated in US NRC GL 88-01 or BWRVIP-75(Archived) are designed to maintain structural integrity and ensure that aging effects will be discovered and repaired before the loss of intended function of the component. The program uses volumetric examinations to detect IGSCC.

The US NRC GL 88-01 recommends that the detailed inspection procedure, equipment, and examination personnel be qualified by a formal program approved by the US NRC. These inspection guidelines, updated in the approved BWRVIP-75(Archived) document, provide the technical basis for revisions to US NRC GL 88-01 inspection schedules. Inspection can reveal crack initiation and growth and leakage of coolant. The extent and frequency of inspection recommended by the program are based on the condition of each weld (e.g., whether the weldments were made from IGSCC-resistant material, whether a stress improvement process was applied to a weldment to reduce residual stresses, and how the weld was repaired if it had been cracked). The inspection guidance in approved BWRVIP-75(Archived) replaces the extent and schedule of inspection in US NRC GL 88-01.

5. Monitoring and Trending: The extent and schedule for inspection, in accordance with the recommendations of US NRC GL 88-01 or approved BWRVIP-75(Archived) guidelines, provide timely detection of cracks and leakage of coolant. Based on inspection results, US NRC GL 88-01 or approved BWRVIP-75(Archived) guidelines provide guidelines for additional samples of welds to be inspected when one or more cracked welds are found in a weld category.

6. Acceptance Criteria: As recommended in US NRC GL 88-01, any indication detected is evaluated in accordance with the ASME Section XI, Subsection IWB-3640 (1995 edition through the 1996 addenda) and the guidelines of NUREG-0313.

Applicable and approved BWRVIP-14(Archived), BWRVIP-59(Archived), BWRVIP-60(Archived), and BWRVIP-62(Archived) documents provide guidelines for evaluation of crack growth in SSs, nickel alloys, and lowalloy steels. An applicant may use BWRVIP-61 guidelines for BWR vessel and internals induction heating stress improvement effectiveness on crack growth in operating plants.

7. Corrective Actions: The guidance for weld overlay repair and stress improvement or replacement is provided in US NRC GL 88-01; ASME Section XI, Subsections IWB-4000 and IWB-7000, IWC-4000 and IWC-7000, or IWD-4000 and IWD-7000, respectively for Class 1, 2, or 3 components; and ASME Code Case N 504-1. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable in addressing corrective actions.

8. Confirmation Process: Site quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable in addressing the confirmation process and administrative controls.

9. Administrative Controls: See Item 8, above.

10. Operating Experience: Intergranular stress corrosion cracking has occurred in small- and large-diameter BWR piping made of austenitic stainless steel and nickel-base alloys. Cracking has occurred in recirculation, core spray, residual heat removal (RHR), and reactor water cleanup (RWCU) system piping welds (US NRC GL 88-01, US NRC Information Notices [INs] 82-39 and 84-41). The comprehensive program outlined in US NRC GL 88-01 and NUREG-0313 and in the staff-approved BWRVIP-75(Archived) report addresses mitigating measures for SCC or IGSCC (e.g., susceptible material, significant tensile stress, and an aggressive environment). The GL 88-01 program has been effective in managing IGSCC in BWR reactor coolant pressure-retaining components and the revision to the GL 88-01 program, according to the staff-approved BWRVIP-75(Archived) report, will adequately manage IGSCC degradation.

References

10 CFR 50.55a, Codes and Standards, Office of the Federal Register, National Archives and Records Administration, 2000.

ASME Code Case N-504-1, Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping, Section XI, Division 1, 1995 Edition, ASME Boiler and Pressure Vessel Code -Code Cases -Nuclear Components, American Society of Mechanical Engineers, New York, NY.

ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, ASME Boiler and Pressure Vessel Code, 1995 edition through the 1996 addenda, American Society of Mechanical Engineers, New York, NY.

BWRVIP-14, Evaluation of Crack Growth in BWR Stainless Steel RPV Internals, (EPRI TR-105873, July 11, 2000)(Archived), Final Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-14, December 3, 1999.

BWRVIP-29 (EPRI TR-103515)(Archived), BWR Vessel and Internals Project, BWR Water Chemistry Guidelines-1993 Revision, Normal and Hydrogen Water Chemistry, Electric Power Research Institute, Palo Alto, CA, February 1994.

BWRVIP-59, Evaluation of Crack Growth in BWR Nickel-Base Austenitic Alloys in RPV Internals, (EPRI TR-108710)(Archived), BWRVIP and Electric Power Research Institute, Palo Alto, CA, March 24, 2000.

BWRVIP-60, BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Low Alloy Steel RPV Internals, (EPRI TR-108709, April 14, 2000)(Archived), Final Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-60, July 8, 1999.

BWRVIP-61, BWR Vessel and Internals Induction Heating Stress Improvement Effectiveness on Crack Growth in Operating Reactors, (EPRI TR-112076), BWRVIP and Electric Power Research Institute, Palo Alto, CA, January 29, 1999.

BWRVIP-62, BWR Vessel and Internals Project, Technical Basis for Inspection Relief for BWR Internal Components with Hydrogen Injection, (EPRI TR-108705)(Archived), BWRVIP and Electric Power Research Institute, Palo Alto, CA, March 7, 2000.

BWRVIP-75, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (NUREG-0313), (EPRI TR-113932, Feb. 29, 2000)(Archived), Initial Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-75, September 15, 2000.

US NRC Generic Letter 88-01, US NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, U.S. Nuclear Regulatory Commission, January 25, 1988; Supplement 1, February 4, 1992.

US NRC Information Notice 84-41, IGSCC in BWR Plants, U.S. Nuclear Regulatory Commission, June 1, 1984.

NUREG-0313, Rev. 2, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, W. S. Hazelton and W. H. Koo, U.S. Nuclear Regulatory Commission, 1988.

Boiling Water Reactor

United States of America

Nuclear Regulatory Commission - The US federal government agency that regulates nuclear power plants

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.

Stress Corrosion Cracking

Cast Austenitic Stainless Steel

Electric Power Research Institute

American Society of Mechanical Engineers

Quality Assurance

Reactor Pressure Vessel

US NRC regulatory guide