1801 R0 XI.M11

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XI.M11 NICKEL-ALLOY NOZZLES AND PENETRATIONS

Program Description

The program includes (a) primary water stress corrosion cracking (PWSCC) susceptibility assessment to identify susceptible components, (b) monitoring and control of reactor coolant water chemistry to mitigate PWSCC, and (c) inservice inspection (ISI) of reactor vessel head penetrations in accordance with the American Society of Mechanical Engineers (ASME) Code, Section Xl, Subsection IWB, Table IWB 2500-1 (1995 edition through the 1996 addenda) to monitor PWSCC and its effect on the intended function of the component. For susceptible penetrations and locations, the program includes an industry-wide, integrated, long-term inspection program based on the industry responses to US NRC Generic Letter (GL) 97-01 contained in the NEI letter dated December 11, 1998, from Dave Modeen to Gus Lainas, "Responses to NRC Requests for Additional Information (RAIs) on GL 97-01" and individual plant responses. Primary water chemistry is monitored and maintained in accordance with the Electric Power Research Institute (EPRI) guidelines in TR-105714(Archived) (Rev. 3, or later revisions or update) to minimize the potential of crack initiation and growth.

Evaluation and Technical Basis

1. Scope of Program: The program is focused on managing the effects of crack initiation and growth due to primary water stress corrosion cracking (PWSCC) of nickel-base alloys. The program includes ISI in accordance with ASME Subsection IWB, Table IWB 2500-1. For susceptible components and locations, the program includes an industry wide, integrated, long-term inspection program based on the industry responses to US NRC GL 97-01 contained in the NEI letter dated December 11, 1998, from Dave Modeen to Gus Lainas, "Responses to NRC Requests for Additional Information (RAIs) on GL 97-01" and individual plant responses. Preventive measures are in accordance with EPRI guidelines in TR-105714(Archived) to mitigate PWSCC. An integrated cracking susceptibility assessment in accordance with industry susceptibility models and inspection results was performed in response to US NRC GL 97-01, to define the most susceptible plants and rank them in accordance with their susceptibility. The information is used by each plant to determine the proper timing of vessel head penetration examinations, either during the current license period or the period of license renewal, if necessary. The components and locations to be included in the long-term inspection program are those that currently have been identified as susceptible to PWSCC, and those that will become susceptible during the period of license renewal. Significant changes in the industry models, as future plants perform inspection, may require reassessment.

2. Preventive Actions: Preventive measures to mitigate PWSCC are in accordance with EPRI guidelines in TR-105714(Archived). The program description and the evaluation and technical basis of monitoring and maintaining reactor water chemistry are presented in Chapter XI.M2, "Water Chemistry."

3. Parameters Monitored/Inspected: The program monitors the effects of PWSCC on the intended function of the control rod drive (CRD) and other Alloy 600 head penetrations by detection and sizing of cracks and coolant leakage by ISI. In C-E-designed pressurized water reactors (PWRs), the CRD head penetration is called the control element drive (CED) head penetration.

4. Detection of Aging Effects: A review of the scope and schedule of the inspections, including the leakage detection system, based on US NRC GL 97-01, assures detection of cracks before the loss of intended function of the components.

The PWSCC susceptibility assessment was performed in response to US NRC GL 97-01 utilizing the most current industry susceptibility models that were based on material and operating parameters and inspection results to date, to rank plants in accordance with their susceptibility. This information is used to develop a plant-specific long-term inspection program, including schedule, scope and determination whether an augmented inspection program of nozzle penetration, including a combination of surface and volumetric examination, is necessary. Because the leakage through cracks in nozzles can be small, this aging management program (AMP), in accordance with US NRC GL 97-01, recommends implementation of an enhanced leakage detection method for detecting small leaks during plant operation.

5. Monitoring and Trending: An inspection schedule, in accordance with the integrated inspection program based on the US NRC GL 97-01 susceptibility assessment, provides timely detection of cracks. Inspection results are used to update the susceptibility models. The frequency of subsequent inspections is based on the finding of the initial inspections and flaw evaluations performed with staff-approved crack growth rate models for nickel-alloys.

6. Acceptance Criteria: Any indication detected is evaluated in accordance with ASME Section XI or other acceptable flaw evaluation criteria. To verify the adequacy of the long-term inspection program and acceptance criteria, if there have been significant changes since the applicants response to US NRC GL 97-01, the applicant either provides references to appropriate industry model revisions or provides updated information on crack initiation and crack growth data and models.

7. Corrective Actions: Repair and replacement procedures are equivalent to those requirements in ASME Section XI. Repair is in conformance with IWB-4000 and replacement is in accordance with IWB-7000. 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: Cracking of Alloy 600 has occurred in domestic and foreign PWRs (US NRC Information Notice (IN) 90-10). Furthermore, ingress of demineralizer resins has also occurred in operating plants (US NRC IN 96-11), the program relies upon monitoring and control of primary water chemistry to manage the effects of such excursions. An integrated susceptibility assessment and inspection program, based on the guidelines in US NRC GL 97-01, is effective in managing the effect of PWSCC on the intended function of reactor vessel head penetrations.

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.

EPRI TR-105714(Archived), PWR Primary Water Chemistry Guidelines-Revision 3, Electric Power Research Institute, Palo Alto, CA, November 1995.

Letter from David J. Modeen of Nuclear Energy Institute to Gus C. Lainais of Division of Engineering, Responses to NRC Requests for Additional Information on Generic Letter 97-01, December 11, 1998.

US NRC Generic Letter 97-01, Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations, U.S. Nuclear Regulatory Commission, April 1, 1997.

US NRC Information Notice 90-10, Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600, U.S. Nuclear Regulatory Commission, February 23, 1990.

US NRC Information Notice 96-11, Ingress of Demineralizer Resins Increase Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations, U.S. Nuclear Regulatory Commission, February 14, 1996.

Primary Water Stress Corrosion Cracking

American Society of Mechanical Engineers

United States of America

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

Nuclear Energy Institute

Electric Power Research Institute

Aging Management Program

Quality Assurance

Pressurized Water Reactor