XI.M5 (NUREG-1801 R1)
XI.M5 BWR FEEDWATER NOZZLE
Program Description
This program includes (a) enhanced inservice inspection (ISI) in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Code, Section XI, Subsection IWB, Table IWB 2500-1 (2001 edition4 including the 2002 and 2003 Addenda) and the recommendation of General Electric (GE) NE-523-A71-0594, and (b) system modifications to mitigate cracking. The program specifies periodic ultrasonic inspection of critical regions of the boiling water reactor (BWR) feedwater nozzle.
Evaluation and Technical Basis
- 1. Scope of Program: The program includes enhanced ISI to monitor the effects of cracking on the intended function of the component, and systems modifications to mitigate cracking.
- 2. Preventive Actions: Mitigation occurs by systems modifications, such as removal of stainless steel cladding and installation of improved spargers. Mitigation is also accomplished by changes to plant-operating procedures, such as improved feedwater control and rerouting of the reactor water cleanup system, to decrease the magnitude and frequency of temperature fluctuations.
- 3. Parameters Monitored/Inspected: The aging management program (AMP) monitors the effects of cracking on the intended function of the component by detection and sizing of cracks by ISI in accordance with ASME Section XI, Subsection IWB and the recommendation of GE NE-523-A71-0594.
- 4. Detection of Aging Effects: The extent and schedule of the inspection prescribed by the program are designed to ensure that aging effects will be discovered and repaired before the loss of intended function of the component. Inspection can reveal cracking. GE NE-523-A71-0594 specifies ultrasonic testing (UT) of specific regions of the blend radius and bore. The UT examination techniques and personnel qualifications are in accordance with the guidelines of GE NE-523-A71-0594. Based on the inspection method and techniques and plant-specific fracture mechanics assessments, the inspection schedule is in accordance with Table 6-1 of GE NE-523-A71-0594. Leakage monitoring may be used to modify the inspection interval.
- 5. Monitoring and Trending: Inspections scheduled in accordance with GE NE-523-A71-0594 provide timely detection of cracks.
- 6. Acceptance Criteria: Any cracking is evaluated in accordance with IWB-3100 by comparing inspection results with the acceptance standards of IWB-3400 and IWB-3500.
- 7. Corrective Actions: Repair is performed in conformance with IWB-4000 and replacement 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 to address the 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 to address the confirmation process and administrative controls.
- 9. Administrative Controls: See Item 8, above.
- 10. Operating Experience: Cracking has occurred in several BWR plants (NUREG-0619, US NRC Generic Letter 81-11). This AMP has been implemented for nearly 25 years and has been found to be effective in managing the effect of cracking on the intended function of feedwater nozzles.
References
10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, Office of the Federal Register, National Archives and Records Administration, 2005.
ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, ASME Boiler and Pressure Vessel Code, 2001 edition including the 2002 and 2003 Addenda, American Society of Mechanical Engineers, New York, NY.
GE-NE-523-A71-0594, Rev. 1, Alternate BWR Feedwater Nozzle Inspection Requirements, BWR Owner’s Group, August 1999.
US NRC Generic Letter 81-11, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking (NUREG-0619), U.S. Nuclear Regulatory Commission, February 29, 1981.
NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, U.S. Nuclear Regulatory Commission, November 1980.