XI.M8 (NUREG-1801 R1)
XI.M8 BWR PENETRATIONS
Program Description
The program includes (a) inspection and flaw evaluation in conformance with the guidelines of staff-approved boiling water reactor vessel and internals project BWRVIP-49(Archived) and BWRVIP-27(Archived) documents, and (b) monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-29 (Electric Power Research Institute (EPRI) TR-103515)(Archived) to ensure the long-term integrity and safe operation of boiling water reactor (BWR) vessel internal components. BWRVIP-49(Archived) provides guidelines for instrument penetrations, and BWRVIP-27(Archived) addresses the standby liquid control (SLC) system nozzle or housing.
Evaluation and Technical Basis
- 1. Scope of Program: The program is focused on managing the effects of cracking due to stress corrosion cracking (SCC) or intergranular stress corrosion cracking (IGSCC). The program contains preventive measures to mitigate SCC or IGSCC, inservice inspection (ISI) to monitor the effects of cracking on the intended function of the components, and repair and/or replacement as needed to maintain the ability to perform the intended function.
- The inspection and evaluation guidelines of BWRVIP-49(Archived) and BWRVIP-27(Archived) contain generic guidelines intended to present appropriate inspection recommendations to assure safety function integrity. The guidelines of BWRVIP-49(Archived) provide information on the type of instrument penetration, evaluate their susceptibility and consequences of failure, and define the inspection strategy to assure safe operation. The guidelines of BWRVIP-27(Archived) are applicable to plants in which the SLC system injects sodium pentaborate into the bottom head region of the vessel (in most plants, as a pipe within a pipe of the core plate DP monitoring system). The BWRVIP-27(Archived) guidelines address the region where the DP and SLC nozzle or housing penetrates the vessel bottom head and include the safe ends welded to the nozzle or housing. Guidelines for repair design criteria are provided in BWRVIP-57(Archived) for instrumentation penetrations, and BWRVIP-53(Archived) for SLC line.
- 2. Preventive Actions: 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 (EPRI TR-103515)(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 SCC/IGSCC on the intended function of the component by detection and sizing of cracks by ISI in accordance with the guidelines of approved BWRVIP-49(Archived) or BWRVIP-27(Archived) and the requirements of the American Society of Mechanical Engineers (ASME) Code, Section XI, Table IWB 2500-1 (2001 edition including the 2002 and 2003 Addenda). An applicant may use the guidelines of BWRVIP-62(Archived) for inspection relief for vessel internal components with hydrogen water chemistry, provided that such relief is submitted under the provisions of 10 CFR 50.55a and approved by the staff.
- 4. Detection of Aging Effects: The evaluation guidelines of BWRVIP-49(Archived) and BWRVIP-27(Archived) recommend that the inspection requirements currently in ASME Section XI continue to be followed. The extent and schedule of the inspection and test techniques prescribed by the ASME Section XI 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 cracking and leakage of coolant. The nondestructive examination (NDE) techniques appropriate for inspection of BWR vessel internals including the uncertainties inherent in delivering and executing NDE techniques in a BWR, are included in BWRVIP-03(Archived).
- Instrument penetrations and SLC system nozzles or housings are inspected in accordance with the requirements of ASME Section XI, Subsection IWB. Components are examined and tested as specified in Table IWB-2500-1, examination categories B-E for pressure-retaining partial penetration welds in vessel penetrations, B-D for full penetration nozzle-to-vessel welds, B-F for pressure-retaining dissimilar metal nozzle-to-safe end welds, or B-J for similar metal nozzle-to-safe end welds. In addition, these components are part of examination category B-P for pressure-retaining boundary. Further details for examination are described in Chapter XI.M1, “ASME Section XI, Inservice Inspection, Subsections IWB, IWC, and IWD,” of this report.
- 5. Monitoring and Trending: Inspections scheduled in accordance with IWB-2400 and approved BWRVIP-48(Archived) or BWRVIP-27(Archived) provide timely detection of cracks. The scope of examination and reinspection must be expanded beyond the baseline inspection if flaws are detected.
- 6. Acceptance Criteria: Any indication detected is evaluated in accordance with ASME Section XI or other acceptable flaw evaluation criteria, such as the staff-approved BWRVIP-49(Archived) or BWRVIP-27(Archived) guidelines. Applicable and approved BWRVIP-14(Archived), BWRVIP-59(Archived), and BWRVIP-60(Archived) documents provide guidelines for evaluation of crack growth in stainless steels (SSs), nickel alloys, and low-alloy steels, respectively.
- 7. Corrective Actions: Repair and replacement procedures in staff-approved BWRVIP-57(Archived) and BWRVIP-53(Archived) are equivalent to those required in the ASME Section XI. Guidelines for repair design criteria are provided in BWRVIP-57(Archived) for instrumentation penetrations and BWRVIP-53(Archived) for standby liquid control line. As discussed in the appendix to this report, the staff finds that licensee implementation of the guidelines in BWRVIP-48(Archived), as modified, will provide an acceptable level of quality for inspection and flaw evaluation of the safety-related components addressed in accordance with 10 CFR Part 50, Appendix B, 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 that licensee implementation of the guidelines in BWRVIP-48(Archived), as modified, will provide an acceptable level of quality for inspection and flaw evaluation of the safety-related components addressed in accordance with the 10 CFR Part 50, Appendix B, confirmation process and administrative controls.
- 9. Administrative Controls: See Item 8, above.
- 10. Operating Experience: Cracking due to SCC or IGSCC has occurred in BWR components made of austenitic stainless steels and nickel alloys. The program guidelines are based on an evaluation of available information, including BWR inspection data and information about the elements that cause IGSCC, to determine which locations may be susceptible to cracking. Implementation of the program provides reasonable assurance that cracking will be adequately managed so the intended functions of the instrument penetrations and SLC system nozzles or housings will be maintained consistent with the current licensing basis CLB) for the period of extended operation.
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.
10 CFR 50.55a, Codes and Standards, 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.
BWRVIP-03, BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines (EPRI TR-105696 R1, March 30, 1999)(Archived), Final Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-03, July 15, 1999.
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-27, BWR Vessel and Internals Project, BWR Standby Liquid Control System/Core Plate ?P Inspection and Flaw Evaluation Guidelines (EPRI TR-107286, April 1997)(Archived), Final License Renewal Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-27 for Compliance with the License Renewal Rule (10 CFR Part 54), December 20, 1999.
BWRVIP-29, BWR Vessel and Internals Project, BWR Water Chemistry Guidelines–1993 Revision, Normal and Hydrogen Water Chemistry (EPRI TR-103515)(Archived), Electric Power Research Institute, Palo Alto, CA, February 1994.
BWRVIP-48, BWR Vessel and Internals Project, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines (EPRI TR-108724, February 1998)(Archived), Final License Renewal Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-48 for Compliance with the License Renewal Rule (10 CFR Part 54), January 17, 2001.
BWRVIP-49, BWR Vessel and Internals Project, Instrument Penetration Inspection and Flaw Evaluation Guidelines (EPRI TR-108695, March 1998)(Archived), Final License Renewal Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-49 for Compliance with the License Renewal Rule (10 CFR Part 54), September 1, 1999.
BWRVIP-53, BWR Vessel and Internals Project, Standby Liquid Control Line Repair Design Criteria (EPRI TR-108716, March 24, 2000)(Archived), Initial Safety Evaluation Report by the Office of Nuclear Reactor Regulation for BWRVIP-53, October 26, 2000.
BWRVIP-57, BWR Vessel and Internals Project, Instrument Penetration Repair Design Criteria (EPRI TR-108721)(Archived), March 24, 2000.
BWRVIP-59, Evaluation of Crack Growth in BWR Nickel-Base Austenitic Alloys in RPV Internals (EPRI TR-108710)(Archived), 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-62, BWR Vessel and Internals Project, Technical Basis for Inspection Relief for BWR Internal Components with Hydrogen Injection (EPRI TR-108705)(Archived), March 7, 2000.