XI.M3 (NUREG-1801 R0)
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XI.M3 REACTOR HEAD CLOSURE STUD BOLTING
Program Description
This program includes (a) inservice inspection (ISI) in conformance with the requirements of the American Society of Mechanical Engineers (ASME), Code, Section XI, Subsection IWB (1995 edition through the 1996 addenda), Table IWB 2500-1, and (b) preventive measures to mitigate cracking.
Evaluation and Technical Basis
- 1. Scope of Program: The program includes (a) ISI to detect crack initiation and growth due to stress corrosion cracking (SCC) or intergranular stress corrosion cracking (IGSCC); loss of material due to wear; and coolant leakage from reactor vessel closure stud bolting for both boiling water reactors (BWRs) and pressurized water reactors (PWRs), and (b) preventive measures of US NRC Regulatory Guide 1.65 to mitigate cracking. The program is applicable to closure studs and nuts constructed from materials with a maximum tensile strength limited to less than 1,172 MPa (170 ksi) (US Nuclear Regulatory Commission (US NRC) Regulatory Guide (RG) 1.65).
- 2. Preventive Actions: Preventive measures include avoiding the use of metal-plated stud bolting to prevent degradation due to corrosion or hydrogen embrittlement and to use manganese phosphate or other acceptable surface treatments and stable lubricants (RG 1.65). Implementation of these mitigation measures is an effective option for reducing SCC or IGSCC and for this program to be effective.
- 3. Parameters Monitored/Inspected: The ASME Section Xl 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 and ensure that aging effects will be discovered and repaired before the loss of intended function of the component. Inspection can reveal crack initiation and growth, 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, liquid penetration, or eddy current examinations to indicate the presence of surface discontinuities and flaws. Volumetric examination uses radiographic, ultrasonic, or eddy current 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 as specified in Table IWB-2500-1. Examination category B-G-1, for pressure-retaining bolting greater than 2 in. in diameter in reactor vessels specifies volumetric examination of studs in place, from the top of the nut to the bottom of the flange hole, and surface and volumetric examination of studs when removed. Also specified are volumetric examination of flange threads and visual VT-1 examination of surfaces of nuts, washers, and bushings. 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 and the system hydrostatic test.
- 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-31 00 by comparing ISI results with the acceptance standards of IWB-3400 and IWB-3500.
- 7. Corrective Actions: Repair and replacement are in conformance with the requirements of IWB-400 and IWB-7000, respectively, and the material and inspection guidance of RG 1.65. 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: The SCC has occurred in BWR pressure vessel head studs (Stoller 1991). The aging management program (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 will be adequately managed so that the intended functions of the reactor head closure studs and bolts will be maintained consistent with the current licensing basis (CLB) for the period of extended operation.
References
10 CFR 50.55a, Codes and Standards, Office of the Federal Register, National Archives and Records Administration, 2000.
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.
US NRC Regulatory Guide 1.65, Material and Inspection for Reactor Vessel Closure Studs, U.S. Nuclear Regulatory Commission, October 1973.
Stoller, S. M., Reactor Head Closure Stud Cracking, Material Toughness Outside FSAR -SCC in Thread Roots, Nuclear Power Experience, BWR-2, III, 58, p. 30, 1991.