2191 R0 XI.M1

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XI.M1 ASME SECTION XI INSERVICE INSPECTION, SUBSECTIONS IWB, IWC, AND IWD

Program Description

Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, imposes the inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Rules for ISI of Nuclear Power Plant Components for Class 1, 2, and 3 pressure-retaining components and their integral attachments in light-water cooled power plants. The rules of Section XI require a mandatory program of examinations, testing and inspections to demonstrate adequate safety and to manage deterioration and aging effects. Inspection of these components is covered in Subsections IWB, IWC, and IWD, respectively, in accordance with the applicable plant ASME Code Section XI edition(s) and addenda as required by 10 CFR 50.55a(g)(4).1 The program generally includes periodic visual, surface, and/or volumetric examination and leakage test of Class 1, 2, and 3 pressure-retaining components and their integral attachments. Repair/replacement activities for these components are covered in Subsection IWA of the ASME Code.

The ASME Code Section XI ISI program, in accordance with Subsections IWA, IWB, IWC, and IWD, has been shown to be generally effective in managing aging effects in Class 1, 2, and 3 components and their integral attachments in light-water cooled power plants. https://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-0055a.html 10 CFR 50.55a] imposes additional conditions and augmentations of ISI requirements specified in the ASME Code, Section XI, and those conditions or augmentations described in 10 CFR 50.55a are included as part of this program. In certain cases, the ASME Code Section XI ISI program is augmented to manage effects of aging for license renewal and is so identified in the collections/nuregs/staff/sr2191/r0/index.html Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report.

Evaluation and Technical Basis

1. Scope of Program:The ASME Code Section XI program provides the requirements for ISI, repair, and replacement of Class 1, 2, and 3 pressure-retaining components and their integral attachments in light-water cooled nuclear power plants. The components within the scope of the program are specified in ASME Code, Section XI, Subsections IWB-1100, IWC-1100, and IWD-1100 for Class 1, 2, and 3 components, respectively. The components described in Subsections IWB-1220, IWC-1220, and IWD-1220 are exempt from the volumetric and surface examination requirements, but not exempt from VT-2 visual examination and pressure testing requirements of Subsections IWB-2500, IWC-2500, and IWD-2500.

2. Preventive Actions:This is a condition monitoring program; therefore, this program does not implement preventive actions.

3. Parameters Monitored/Inspected:The ASME Code, Section XI ISI program detects degradation of components by using the examination and inspection requirements specified in ASME Code, Section XI Tables IWB-2500-1, IWC-2500-1, and IWD-2500-1 for Class 1, 2, and 3 components, respectively.

The program uses three types of examination—visual, surface, and volumetric—in accordance with the requirements of Subsection IWA-2000. Visual VT-1 examination detects discontinuities and imperfections, such as cracks, corrosion, wear, or erosion, on the surface of components. Visual VT-2 examination detects evidence of leakage from pressure-retaining components, as required during the system pressure test. Visual VT-3 examination (a) determines the general mechanical and structural condition of components and their supports by verifying parameters such as clearances, settings, and physical displacements; (b) detects discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear, or erosion; and (c) observes conditions that could affect operability or functional adequacy of constant-load and spring-type components and supports.

Surface examination uses magnetic particle, liquid penetrant, 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 included in the inspection program.

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 to detect and repair or replace components before the loss of intended function of the component. Inspection can reveal cracking, loss of material due to corrosion, leakage of coolant, and indications of degradation due to wear or stress relaxation (such as changes in clearances, settings, physical displacements, loose or missing parts, debris, wear, erosion, or loss of integrity at bolted or welded connections).

Class 1, 2, and 3 components are examined and tested as specified in Tables IWB-2500-1, IWC-2500-1, and IWD-2500-1, respectively. The tables specify the extent and schedule of the inspection and examination methods for the components of the pressure-retaining boundaries.

5. Monitoring and Trending:For Class 1, 2, and 3 components, the inspection schedule of IWB-2400, IWC-2400, and IWD-2400, and the extent and frequency of IWB-2500-1, IWC-2500-1, and IWD-2500-1, respectively, provides for timely detection of degradation. The sequence of component examinations established during the first inspection interval is repeated during each successive inspection interval, to the extent practical. Volumetric and surface examination results are compared with recorded preservice examination and prior inservice examinations. Flaw conditions or relevant conditions of degradation are evaluated in accordance with IWB-3100, IWC-3100, or IWD-3100.

Examinations that reveal indications that exceed the acceptance standards described below are extended to include additional examinations in accordance with IWB-2430, IWC-2430, and IWD-2430 for Class 1, 2, and 3 components, respectively. Examination results that exceed the acceptance standards below are repaired/replaced or accepted by analytical evaluation in accordance with IWB-3600, IWC-3600 or IWD-3600, as applicable. Those items accepted by analytical evaluation are reexamined during the next three inspection periods of IWB-2410 for Class 1 components, IWC-2410 for Class 2 components, and IWD-2410 for Class 3 components.

6. Acceptance Criteria:Any indication or relevant conditions of degradation are evaluated in accordance with IWB-3000, IWC-3000, and IWD-3000 for Class 1, 2, and 3 components, respectively. Examination results are evaluated in accordance with IWB-3100, IWC-3100, or IWD-3100 by comparing the results with the acceptance standards of IWB-3400 and IWB-3500 for Class 1, IWC-3400 and IWC-3500 for Class 2, and IWD-3400 and IWD-3500 for Class 3 components. Flaws that exceed the size of allowable flaws, as defined in IWB-3500, IWC-3500 and IWD-3500 may be evaluated by using the analytical procedures of IWB-3600, IWC-3600 and IWD-3600 for Class 1, 2 and 3 components, respectively.

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 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 corrective actions element of this aging management program (AMP) for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

Repair and replacement activities are performed in conformance with IWA-4000.

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:Because the ASME Code is a consensus document that has been widely used over a long period, it has been shown to be generally effective in managing aging effects in Class 1, 2, and 3 components and their integral attachments in light-water cooled power plants (see Chapter I of the GALL-SLR Report).

Some specific examples of operating experience (OE) of component degradation are as follows:

Boiling Water Reactor (BWR): Cracking due to intergranular stress corrosion cracking (IGSCC) has occurred in small- and large-diameter BWR piping made of austenitic stainless steel (SS) and nickel alloys. IGSCC has also occurred in a number of vessel internal components, such as core shrouds, access hole covers, top guides, and core spray spargers [U.S. Nuclear Regulatory Commission (US NRC) Inspection and Enforcement Bulletin (IEB) 80-13, US NRC Information Notice (IN) 95-17, US NRC Generic Letter (GL) 94-03, and NUREG–1544]. Cracking due to thermal and mechanical loading has occurred in high-pressure coolant injection piping (US NRC IN 89-80) and instrument lines (Licensee Event Report (LER) 249/99-003-01). BWR jet pumps are designed with access holes in the shroud support plate at the bottom of the annulus between the core shroud and the reactor vessel wall. These holes are used for access during construction and are subsequently closed by welding a plate over the hole. Both circumferential (US NRC IN 88-03) and radial cracking (US NRC IN 92-57) have been observed in access hole covers. Failure of the isolation condenser tube bundles due to thermal fatigue and transgranular stress corrosion cracking caused by leaky valves has also occurred (NRC LER 219/98-014-00).

Pressurized Water Reactor (PWR) Primary System: Although the primary pressure boundary piping of PWRs has generally not been found to be affected by stress corrosion cracking (SCC) because of low dissolved oxygen levels and control of primary water chemistry, SCC has occurred in safety injection lines (US NRC IN 97-19 and 84-18), charging pump casing cladding (US NRC IN 80-38 and 94-63), instrument nozzles in safety injection tanks (US NRC IN 91-05), control rod drive seal housing (NRC Inspection Report 50-255/99012), and safety-related SS piping systems that contain oxygenated, stagnant, or essentially stagnant borated coolant (US NRC IN 97-19). Cracking has occurred in SS baffle former bolts in a number of foreign plants (US NRC IN 98-11) and has been observed in plants in the United States. Cracking due to thermal and mechanical loading has occurred in high-pressure injection and safety injection piping (US NRC IN 97-46 and US NRC Bulletin 88-08). Through-wall circumferential cracking has been found in reactor pressure vessel head control rod drive penetration nozzles (US NRC IN 2001-05). Evidence of reactor coolant leakage, together with crack-like indications, has been found in bottom-mounted instrumentation nozzles (US NRC IN 2003-11 and IN 2003-11, Supplement 1). Cracking in pressurizer safety and relief line nozzles and in surge line nozzles has been detected (US NRC IN 2004-11), and circumferential cracking in SS pressurizer heater sleeves has also been found (US NRC IN 2006-27). Also, primary water stress corrosion cracking has been observed in steam generator drain bowl welds inspected as part of a licensee’s Alloy 600/82/182 program (US NRC IN 2005-02).

PWR Secondary System:Steam generator tubes have experienced outside diameter stress corrosion cracking, intergranular attack, wastage, and pitting (US NRC IN 97-88). Carbon steel support plates in steam generators have experienced general corrosion. Steam generator shells have experienced pitting and SCC (US NRC INs 82-37, 85-65, and 90-04).

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry OE 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.” 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.

EPRI. BWRVIP-03, Revision 6 (EPRI 105696-R6)Archived, “BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines.” Palo Alto, California: Electric Power Research Institute. December 2003.

Licensee Event Report 219/98-014-00, “Failure of the Isolation Condenser Tube Bundles due to Thermal Stresses/Transgranular Stress Corrosion Cracking Caused by Leaky Valve.” LER Search. October 1998.

Licensee Event Report 249/99-003-01, “Supplement to Reactor Recirculation B Loop, High Pressure Flow Element Venturi Instrument Line Steam Leakage Results in Unit 3 Shutdown Due to Fatigue Failure of Socket Welded Pipe Joint.” LER Search. August 1999.

US NRC. Bulletin 88-08, “Thermal Stresses in Piping Connected to Reactor Coolant System.” Washington, DC: U.S. Nuclear Regulatory Commission. June 1988. Supplement 1, June 1988. Supplement 2, September 1988. Supplement 3, April 1989.

_____. Generic Letter 94-03, “Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors.” Agencywide Documents Access and Management System (ADAMS) Accession No. ML070600206. Washington, DC: U.S. Nuclear Regulatory Commission. July 1994.

_____. IE Bulletin 80-13, “Cracking in Core Spray Spargers.” Washington, DC: U.S. Nuclear Regulatory Commission. May 1980.

_____. Information Notice 80-38, “Cracking in Charging Pump Casing Cladding.” ADAMS Accession No. ML073550834. Washington, DC: U.S. Nuclear Regulatory Commission. October 1980.

_____. Information Notice 82-37, “Cracking in the Upper Shell to Transition Cone Girth Weld of a Steam Generator at an Operating PWR.” ADAMS Accession No. ML082970942. Washington, DC: U.S. Nuclear Regulatory Commission. September 1982.

_____. Information Notice 84-18, “Stress Corrosion Cracking in PWR Systems.” Washington, DC: U.S. Nuclear Regulatory Commission. March 1984.

_____. Information Notice 85-65, “Crack Growth in Steam Generator Girth Welds.” Washington, DC: U.S. Nuclear Regulatory Commission. July 1985.

_____. Information Notice 88-03, “Cracks in Shroud Support Access Hole Cover Welds.” Washington, DC: U.S. Nuclear Regulatory Commission. February 1988.

_____. Information Notice 89-80, “Potential for Water Hammer, Thermal Stratification, and Steam Binding in High-Pressure Coolant Injection Piping.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1989.

_____. Information Notice 90-04, “Cracking of the Upper Shell-to-Transition Cone Girth Welds in Steam Generators.” Washington, DC: U.S. Nuclear Regulatory Commission. January 1990.

_____. Information Notice 91-05, “Intergranular Stress Corrosion Cracking in Pressurized Water Reactor Safety Injection Accumulator Nozzles.” Washington, DC: U.S. Nuclear Regulatory Commission. January 1991.

_____. Information Notice 92-57, “Radial Cracking of Shroud Support Access Hole Cover Welds.” Washington, DC: U.S. Nuclear Regulatory Commission. August 1992.

_____. Information Notice 94-63, “Boric Acid Corrosion of Charging Pump Casing Caused by Cladding Cracks.” Washington, DC: U.S. Nuclear Regulatory Commission. August 1994.

_____. Information Notice 95-17, “Reactor Vessel Top Guide and Core Plate Cracking.” Washington, DC: U.S. Nuclear Regulatory Commission. March 1995.

_____. Information Notice 97-19, “Safety Injection System Weld Flaw at Sequoyah Nuclear Power Plant, Unit 2.” Washington, DC: U.S. Nuclear Regulatory Commission. April 18, 1997.

_____. Information Notice 97-46, “Unisolable Crack in High-Pressure Injection Piping.” Washington, DC: U.S. Nuclear Regulatory Commission. July 1997.

_____. Information Notice 97-88, “Experiences During Recent Steam Generator Inspections.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1997.

_____. Information Notice 98-11, “Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. March 1998.

_____. Information Notice 2001-05, “Through-Wall Circumferential Cracking of Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzles at Oconee Nuclear Station, Unit 3.” Washington, DC: U.S. Nuclear Regulatory Commission. April 2001.

_____. Information Notice 2003-11, “Leakage Found on Bottom-Mounted Instrumentation Nozzles.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2003.

_____. Information Notice 2003-11, Supplement 1, “Leakage Found on Bottom-Mounted Instrumentation Nozzles.” Washington, DC: U.S. Nuclear Regulatory Commission. January 2004.

_____. Information Notice 2004-11, “Cracking in Pressurizer Safety and Relief Nozzles and in Surge Line Nozzles.” Washington, DC: U.S. Nuclear Regulatory Commission. May 2004.

_____. Information Notice 2005-02, “Pressure Boundary Leakage Identified on Steam Generator Drain Bowl Welds.” Washington, DC: U.S. Nuclear Regulatory Commission. February 2005.

_____. Information Notice 2006-27, “Circumferential Cracking in the Stainless Steel Pressurizer Heater Sleeves of Pressurized Water Reactors.” Washington, DC: U.S. Nuclear Regulatory Commission. December 2006.

_____. Inspection Report 50-255/99012, “Palisades Inspection Report.” Item E8.2, Licensee Event Report 255/99-004, “Control Rod Drive Seal Housing Leaks and Crack Indications.” Washington, DC: U.S. Nuclear Regulatory Commission. January 2000.

_____. NUREG–1544, “Status Report: Intergranular Stress Corrosion Cracking of BWR Core Shrouds and Other Internal Components.” Washington, DC: U.S. Nuclear Regulatory Commission. March 1996.

American Society of Mechanical Engineers

Code of Federal Regulations - The US federal government regulations. 10 CFR 54 is the regulation for renewal of nuclear power plant operating licenses.

Generic Aging Lessons Learned - The US NRC GALL Report is NUREG-1801. The US NRC Subsequent License Renewal GALL is NUREG-2191

Subsequent License Renewal - The extension of the original license from 40 years to 60 years is known as License Renewal. Extensions of the license beyond 60 years are known as Subsequent License Renewal.

Quality Assurance

Aging Management Program

Operating Experience

Boiling Water Reactor

United States of America

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

US NRC regulatory guide

Pressurized Water Reactor

Stress Corrosion Cracking

Electric Power Research Institute

Boiling Water Reactor Vessel Internals Program - The BWRVIP is an industry program for General Electric boiling water reactors that has developed procedures for performing inspections of reactor vessel internals.

Agency wide Documents Access and Management System - The official recordkeeping system, through which the U.S. Nuclear Regulatory Commission (US NRC) provides access to the "libraries" or collections of publicly available documents