2191 R0 XI.M16A

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XI.M16A PWR VESSEL INTERNALS

Program Description

This program is used to manage the effects of age-related degradation mechanisms that are applicable to the pressurized water reactor (PWR) reactor vessel internal (RVI) components. These aging effects include: (a) cracking, including stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), and cracking due to fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to thermal aging and neutron irradiation embrittlement; (d) changes in dimensions due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

In the absence of an acceptable generic methodology such as an approved revision of Materials Reliability Program (MRP)-227 that considers an operating period of 80 years, this program may be based on an existing plant program that is consistent with Electric Power Research Institute (EPRI) Technical Report No. 1022863(Archived), “Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines,” (MRP-227-A), which is implemented in accordance with Nuclear Energy Institute (NEI) 03-08, “Guideline for the Management of Materials Issues.” The staff approved the augmented inspection and evaluation (I&E) criteria for PWR RVI components in NRC Safety Evaluation (SE), Revision 1, on MRP-227 by letter dated December 16, 2011.

Because the guidelines of MRP-227-A(Archived) are based on an analysis of the RVI that considers the operating conditions up to a 60-year operating period, these guidelines are supplemented through a gap analysis that identifies enhancements to the program that are needed to address an 80-year operating period. In this program, the term “MRP-227-A (as supplemented)” is used to describe either MRP-227-A (as supplemented) by this gap analysis, or an acceptable generic methodology such as an approved revision of MRP-227 that considers an operating period of 80 years.

The program applies the guidance in MRP-227-A (as supplemented) for inspecting, evaluating, and, if applicable, dispositioning non-conforming RVI components at the facility. These examinations provide reasonable assurance that the effects of age-related degradation mechanisms will be managed during the period of extended operation. The program includes expanding periodic examinations and other inspections, if the extent of the degradation identified exceeds the expected levels.

MRP-227-A(Archived) guidance for selecting RVI components for inclusion in the inspection sample is based on a four-step ranking process. Through this process, the RVIs for all three PWR designs were assigned to one of the following four groups: “Primary,” “Expansion,” “Existing Programs,” and “No Additional Measures.” Definitions of each group are provided in MRP-227-A(Archived).

In the absence of an acceptable generic methodology such as an approved revision of MRP-227 that considers an operating period of 80 years, the gap analysis described below is used to provide reasonable assurance that the aging management for the RVI components identified in the four groups is appropriate for 80 years of operation.

The result of this four-step sample selection process is a set of “Primary” internals component locations for each of the three plant designs that are inspected because they are expected to show the leading indications of the degradation effects, with another set of “Expansion” internals component locations that are specified to expand the sample should the indications be more severe than anticipated.

The degradation effects in a third set of internals locations are deemed to be adequately managed by “Existing Programs,” such as American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Examination Category B-N-3, examinations of core support structures. A fourth set of internals locations are deemed to require “No Additional Measures.”

If the program is based on MRP-227-A(Archived) with a gap analysis, the inspection categories, inspection criteria, and other program characteristics required by MRP-227-A(Archived) are identified and justified for each component in the applicable program elements. The justification should focus on the aging management of the additional aging considerations (i.e., new aging effect/mechanism) during the subsequent period of extended operation. The acceptance criteria in the Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), Section 3.1.2.2.9 and the review procedures in Section 3.1.3.2.9 provide additional information.

Evaluation and Technical Basis

1. Scope of Program: The scope of the program includes all RVI components based on the plant’s applicable nuclear steam supply system design. The scope of the program applies the methodology and guidance in MRP-227-A (as supplemented), which provides an augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants designed by Babcock & Wilcox (B&W), Combustion Engineering (CE), and Westinghouse. The scope of components includes core support structures, those RVI components that serve an intended license renewal safety function pursuant to criteria in Title 10 of the Code of Federal Regulations (10 CFR) 54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii). In addition, ASME Code, Section XI includes inspection requirements for PWR removable core support structures in Table IWB-2500-1, Examination Category B-N-3, which are in addition to any inspections that are implemented in accordance with MRP-227-A (as supplemented).

The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation. The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are managed in accordance with an applicant’s AMP that corresponds to GALL-SLR Report AMP XI.M1, “ASME Code, Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.”

This program element specifies if the program is based on an existing program that is consistent with MRP-227-A(Archived) with a gap analysis or if it is based on an acceptable generic methodology such as an approved revision of MRP-227 that considers an operating period of 80 years. If based on MRP-227-A(Archived) with a gap analysis, the scope of the program focuses on identification and justification of the following:

1. Components that screen in for additional aging effects or mechanisms when assessed for the 60-80 year operating period.
2. Components that previously screened in for an aging effect or mechanism and the severity of that aging effect or mechanism could significantly increase for the 60-80 year operating period.
3. Changes to the existing MRP-227-A(Archived) program characteristics or criteria, including but not limited to changes in inspection categories, inspection criteria, or primary-to-expansion component criteria and relationships.

2. Preventive Actions: The program relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms [e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms (SCC, PWSCC, or IASCC)]. Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program, as described in GALL-SLR Report AMP XI.M2, “Water Chemistry.”

3. Parameters Monitored or Inspected: The program manages the following age-related degradation effects and mechanisms that are applicable in general to RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss of fracture toughness induced by thermal aging and neutron irradiation embrittlement; (d) changes in dimensions due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

For the management of cracking, the program monitors for evidence of surface-breaking linear discontinuities if a visual inspection technique is used as the non-destructive examination (NDE) method or for relevant flaw presentation signals if a volumetric ultrasonic testing (UT) method is used as the NDE method. For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components. For the management of loss of preload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement. Instead, the impact of loss of fracture toughness on component integrity is indirectly managed by: (1) using visual or volumetric examination techniques to monitor for cracking in the components, and (2) applying applicable reduced fracture toughness properties in the flaw evaluations, in cases where cracking is detected in the components and is extensive enough to necessitate a supplemental flaw growth or flaw tolerance evaluation. The program uses physical measurements to monitor for any dimensional changes due to void swelling or distortion.

Specifically, the program implements the parameters monitored/inspected criteria consistent with the applicable tables in Section 4, “Aging Management Requirements,” in MRP-227-A (as supplemented).

4. Detection of Aging Effects: The inspection methods are defined and established in Section 4 of MRP-227-A (as supplemented). Standards for implementing the inspection methods are defined and established in MRP-228(revision referenced is archived). In all cases, well-established inspection methods are selected. These methods include volumetric UT examination methods for detecting flaws in bolting and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). VT-3 visual methods may be applied for the detection of cracking in non-redundant RVI components only when the flaw tolerance of the component, as evaluated for reduced fracture toughness properties, is known and the component has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. VT-3 visual methods are acceptable for the detection of cracking in redundant RVI components (e.g., redundant bolts or pins used to secure a fastened RVI assembly).

In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss of preload caused by thermal and irradiation- enhanced stress relaxation and creep.

The program adopts the guidance in MRP-227-A (as supplemented) for defining the “Expansion Criteria” that need to be applied to the inspection findings of “Primary” components and for expanding the examinations to include additional “Expansion” components. RVI component inspections are performed consistent with the inspection frequency and sampling bases for “Primary” components, “Existing Programs” components, and “Expansion” components in MRP-227-A (as supplemented).

In some cases (as defined in MRP-227-A(Archived)), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for changes in dimensions due to void swelling or distortion.

Inspection coverages for “Primary” and “Expansion” RVI components are implemented consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227-A(Archived), or as modified by a gap analysis.

This program element should justify the appropriateness of the inspection methods, sample size criteria, and inspection frequency criteria for managing the effects of degradation during the subsequent period of extended operation, including any changes to these criteria from their prior assessment in MRP-227-A(Archived).

5. Monitoring and Trending: The methods for monitoring, recording, evaluating, and trending the data that result from the program’s inspections are given in Section 6 of MRP-227-A (as supplemented) and its subsections. Component reinspection frequencies for “Primary” and “Expansion” category components are defined in specific tables in Section 4 of the MRP-227-A report (as supplemented). The examination and re- examinations that are implemented in accordance with MRP-227-A (as supplemented), together with the criteria specified in MRP-228(revision referenced is archived) for inspection methodologies, inspection procedures, and inspection personnel, provide for timely detection, reporting, and implementation of corrective actions for the aging effects and mechanisms managed by the program.

The program applies applicable fracture toughness properties, including reductions for thermal aging or neutron embrittlement, in the flaw evaluations of the components in cases where cracking is detected in a RVI component and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program includes criteria to evaluate the aging effects in the inaccessible portions of the components and the resulting impact on the intended function(s) of the components. For redundant components (such as redundant bolts, screws, pins, keys, or fasteners, some of which are accessible to inspection and some of which are not accessible to inspection), the program includes criteria to evaluate the aging effects in the population of components that are inaccessible by the applicable inspection technique and the resulting impact on the intended function(s) of the assembly containing the components.

Flaw evaluation methods, including recommendations for flaw depth sizing and for crack growth determinations as well as for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications, are defined in MRP-227-A (as supplemented).

6. Acceptance Criteria: Section 5 of MRP-227-A (as supplemented), which includes Table 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-designed RVIs, provides the specific examination and flaw evaluation acceptance criteria for the “Primary” and “Expansion” RVI component examination methods. For RVI components addressed by examinations performed in accordance with the ASME Code, Section XI, the acceptance criteria in IWB-3500 are applicable. For RVI components covered by other “Existing Programs,” the acceptance criteria are described within the applicable reference document. As applicable, the program establishes acceptance criteria for any physical measurement monitoring methods that are credited for aging management of particular RVI components.

This program element should justify the appropriateness of the acceptance criteria for managing the effects of degradation during the subsequent period of extended operation, including any changes to acceptance criteria based on the gap analysis.

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 Generic Aging Lessons Learned for Subsequent License Renewal (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 AMP for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. The implementation of the guidance in MRP-227-A (as supplemented), plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Other alternative corrective actions bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the US NRC. Alternative corrective actions not approved or endorsed by the NRC will be submitted for US NRC approval prior to their implementation.

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.

Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the recommendations of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable. The implementation of the guidance in MRP-227-A (as supplemented), in conjunction with NEI 03-08 and other guidance documents, reports, or methodologies referenced in this AMP, provides an acceptable level of quality and an acceptable basis for confirming the quality of inspections, flaw evaluations, and corrective actions.

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.

The administrative controls for these types of programs, including their implementing procedures and review and approval processes, are implemented in accordance with the recommended industry guidelines and criteria in NEI 03-08 and existing site 10 CFR Part 50, Appendix B, Quality Assurance Programs, or their equivalent, as applicable. The evaluation in Section 3.5 of the NRC’s SE, Revision 1, on MRP-227-A(Archived) provides the basis for endorsing NEI 03-08. This includes endorsement of the criteria in NEI 03-08 for notifying the NRC of any deviation from the I&E methodology in MRP-227-A(Archived) and justifying the deviation no later than 45 days after its approval by a licensee executive.

10. Operating Experience: The review and assessment of relevant operating experience (OE) for its impacts on the program, including implementing procedures, are governed by NEI 03-08 and Appendix A of MRP-227-A(Archived). Consistent with MRP-227-A(Archived), the reporting of inspection results and OE is treated as a “Needed” category item under the implementation of NEI 03-08.

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 and Fuel Reprocessing Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR Part 50.55a, “Codes and Standards.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

ASME. ASME Code, Section V, “Nondestructive Examination.” 2004 Edition. New York, New York: American Society of Mechanical Engineers.

_____. ASME Code, Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” New York, New York: American Society of Mechanical Engineers. 2008.

EPRI. EPRI Technical Report No.1022863(Archived), “Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A).” Agencywide Documents Access and Management System (ADAMS) Accession No. ML12017A193 (Transmittal letter from the EPRI-MRP) and ADAMS Accession Nos. ML12017A194, ML12017A196, ML12017A197, ML12017A191, ML12017A192, ML12017A195 and ML12017A199 (Final Report). Palo Alto, California: Electric Power Research Institute. December 2011.

_____. EPRI 1016609(Archived), “Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228).” (Non-publicly available ADAMS Accession No. ML092120574). The non-proprietary version of the report may be accessed by members of the public at ADAMS Accession No. ML092750569. Palo Alto, California: Electric Power Research Institute. July 2009.

NEI. NEI 03-08, Revision 2, “Guideline for the Management of Materials Issues.” ADAMS Accession No. ML101050337. Washington, DC: Nuclear Energy Institute. January 2010.

US NRC. License Renewal Interim Staff Guidance LR-ISG-2011-04, “Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors.” ADAMS Accession No. ML12270A436. Washington, DC: U.S. Nuclear Regulatory Commission. June 3, 2013.

_____. License Renewal Interim Staff Guidance LR-ISG-2011-05, “Ongoing Review Of Operating Experience.” ADAMS Accession No. ML12044A215. Washington, DC: U.S. Nuclear Regulatory Commission. March 16, 2012.

_____. Safety Evaluation from Robert A. Nelson (NRC) to Neil Wilmshurst (EPRI), “Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596(Archived) (MRP-227), Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines.” ADAMS Accession No. ML11308A770. Washington, DC: U.S. Nuclear Regulatory Commission. December 16, 2011.

Pressurized Water Reactor

Stress Corrosion Cracking

Primary Water Stress Corrosion Cracking

Electric Power Research Institute

Nuclear Energy Institute

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

Safety Evaluation

American Society of Mechanical Engineers

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.

Non Destructive Examination

Ultrasonic Test - A nondestructive test method using ultrasonic waves to determine the thickness of a component and to detect cracks or other degradation.

Enhanced Visual Test

Quality Assurance

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

United States of America

Operating Experience

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

License Renewal Interim Staff Guidance