XI.M3 (NUREG-2191 R0)
XI.M3 REACTOR HEAD CLOSURE STUD BOLTING
Program Description
This program includes (a) inservice inspection (ISI) in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)1 , Section XI, Subsection IWB, Table IWB 2500-1; and (b) preventive measures to mitigate cracking. The program also relies on recommendations delineated in the U.S. Nuclear Regulatory Commission (US NRC) Regulatory Guide (RG) 1.65, Revision 1.
Evaluation and Technical Basis
- 1. Scope of Program: The program manages the aging effects of cracking due to stress corrosion cracking (SCC) or intergranular stress corrosion cracking (IGSCC) and loss of material due to wear or corrosion for reactor vessel closure stud bolting (studs, washers, bushings, nuts, and threads in flange) for both boiling water reactors (BWRs) and pressurized water reactors.
- 2. Preventive Actions: Preventive measures may include:
- Avoiding the use of metal-plated stud bolting to prevent degradation due to corrosion or hydrogen embrittlement;
- Using manganese phosphate or other acceptable surface treatments;
- Using stable lubricants. Of particular note, use of molybdenum disulfide (MoS2) as a lubricant has been shown to be a potential contributor to SCC and should not be used;
- Using bolting material for closure studs that has an actual measured yield strength less than 150 kilo-pounds per square inch (ksi) [1,034 megapascals (MPa)] for newly installed studs, or 170 ksi (1,172 MPa) ultimate tensile strength for existing studs.
- Implementation of these mitigation measures can reduce potential for SCC or IGSCC, thus making this program effective.
- 3. Parameters Monitored/Inspected: The ASME Code Section XI ISI program detects and sizes cracks, detects loss of material, and detects coolant leakage by following the examination and inspection requirements specified in Table IWB-2500-1.
- 4. Detection of Aging Effects: The extent and schedule of the inspection and test techniques prescribed by the program are designed to maintain structural integrity, to detect aging effects and to repair or replace components before the loss of intended function of the component. Inspection can reveal cracking, loss of material due to corrosion or wear, and leakage of coolant.
- The program uses visual, surface, and volumetric examinations in accordance with the general requirements of Subsection IWA-2000. Surface examination uses magnetic particle or liquid penetrant examinations to indicate the presence of surface discontinuities and flaws. Volumetric examination uses radiographic or ultrasonic examinations to indicate the presence of discontinuities or flaws throughout the volume of material. Visual VT-2 examination detects evidence of leakage from pressure-retaining components, as required during the system pressure test.
- Components are examined and tested in accordance with ASME Code, Section XI, Table IWB-2500-1, Examination Category B-G-1, for pressure-retaining bolting greater than 2 inches in diameter. Examination Category B-P for all pressure-retaining components specifies visual VT-2 examination of all pressure-retaining boundary components during the system leakage test. Table IWB-2500-1 specifies the extent and frequency of the inspection and examination methods, and IWB-2400 specifies the schedule of the inspection.
- 5. Monitoring and Trending: The Inspection schedule of IWB-2400 and the extent and frequency of IWB-2500-1 provide timely detection of cracks, loss of material, and leakage.
- 6. Acceptance Criteria: Any indication or relevant condition of degradation in closure stud bolting is evaluated in accordance with IWB-3100 by comparing ISI results with the acceptance standards of IWB-3400 and IWB-3500.
- 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 Regulations (10 CFR) Part 50, Appendix B. Appendix A of the Generic Aging Lessons Learned for Subsequent License Renewal (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 aging management programs (AMP) for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.
- Repair and replacement are performed in accordance with the requirements of IWA-4000 and the material and inspection guidance of RG 1.65. The maximum yield strength of replacement material should be limited as recommended in RG 1.65
- 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: SCC has occurred in BWR pressure vessel head studs (Stoller, 1991). The AMP has provisions regarding inspection techniques and evaluation, material specifications, corrosion prevention, and other aspects of reactor pressure vessel head stud cracking. Implementation of the program provides reasonable assurance that the effects of cracking due to SCC or IGSCC and loss of material due to wear are adequately managed so that the intended functions of the reactor head closure studs and bolts are maintained consistent with the current licensing basis for the subsequent period of extended operation. Degradation of threaded bolting and fasteners in closures for the reactor coolant pressure boundary has occurred from boric acid corrosion, SCC, and fatigue loading (US NRC Inspection and Enforcement Bulletin 82-02, US NRC Generic Letter 91-17).
- 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 and Fuel Reprocessing 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.
US NRC. Generic Letter 91-17, “Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. October 1991.
_____. IE Bulletin 82-02, “Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. June 1982.
_____. Regulatory Guide 1.65, “Material and Inspection for Reactor Vessel Closure Studs.” Revision 1. Washington, DC: U.S. Nuclear Regulatory Commission. April 2010.
Stoller, S.M. “Reactor Head Closure Stud Cracking, Material Toughness Outside FSAR–SCC in Thread Roots.” BWR-2, Ill, 58. Nuclear Power Experience. 1991.