1801 R2 XI.M16A

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XI.M16A PWR VESSEL INTERNALS

Program Description

This program relies on implementation of the Electric Power Research Institute (EPRI) Report No. 1016596 (MRP-227)(Archived) and EPRI Report No. 1016609 (MRP-228)(Archived) to manage the aging effects on the reactor vessel internal (RVI) components.

This program is used to manage the effects of age-related degradation mechanisms that are applicable in general to the PWR RVI components at the facility. These aging effects include (a) various forms of cracking, including stress corrosion cracking (SCC), which also encompasses primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), or cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging or neutron irradiation embrittlement; (d) changes in dimension due to void swelling; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

The program applies the guidance in MRP-227(Archived) for inspecting, evaluating, and, if applicable, dispositioning non-conforming RVI components at the facility. The program conforms to the definition of a sampling-based condition monitoring program, as defined by the Branch Technical Position RSLB-1, with periodic examinations and other inspections of highly-affected internals locations. 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 effects exceeds the expected levels.

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

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 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 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. As a result, the program typically identifies 5 to 15% of the RVI locations as Primary Component locations for inspections, with another 7 to 10% of the RVI locations to be inspected as Expansion Components, as warranted by the evaluation of the inspection results. Another 5 to 15% of the internals locations are covered by Existing Programs, with the remainder requiring no additional measures. This process thus uses appropriate component functionality criteria, age-related degradation susceptibility criteria, and failure consequence criteria to identify the components that will be inspected under the program in a manner that conforms to the sampling criteria for sampling-based condition monitoring programs in Section A.1.2.3.4 of US NRC Branch Position RLSB-1. Consequently, the sample selection process is adequate to assure that the intended function(s) of the PWR reactor internal components are maintained during the period of extended operation.

The program’s use of visual examination methods in MRP-227(Archived) for detection of relevant conditions (and the absence of relevant conditions as a visual examination acceptance criterion) is consistent with the ASME Code, Section XI rules for visual examination. However, the program’s adoption of the MRP-227(Archived) guidance for visual examinations goes beyond the ASME Code, Section XI visual examination criteria because additional guidance is incorporated into MRP-227(Archived) to clarify how the particular visual examination methods will be used to detect relevant conditions and describes in more detail how the visual techniques relate to the specific RVI components and how to detect their applicable age-related degradation effects.

The technical basis for detecting relevant conditions using volumetric ultrasonic testing (UT) inspection techniques can be found in MRP-228(Archived), where the review of existing bolting UT examination technical justifications has demonstrated the indication detection capability of at least two vendors, and where vendor technical justification is a requirement prior to any additional bolting examinations. Specifically, the capability of program’s UT volumetric methods to detect loss of integrity of PWR internals bolts, pins, and fasteners, such as baffle-former bolting in B&W and Westinghouse units, has been well demonstrated by operating experience. In addition, the program’s adoption of the MRP-227(Archived) guidance and process incorporates the UT criteria in MRP-228(Archived), which calls for the technical justifications that are needed for volumetric examination method demonstrations, required by the ASME Code, Section V.

The program also includes future industry operating experience as incorporated in periodic revisions to MRP-227(Archived). The program thus provides reasonable assurance for the long-term integrity and safe operation of reactor internals in all commercial operating U.S. PWR nuclear power plants.

Age-related degradation in the reactor internals is managed through an integrated program. Specific features of the integrated program are listed in the following ten program elements. Degradation due to changes in material properties (e.g., loss of fracture toughness) was considered in the determination of inspection recommendations and is managed by the requirement to use appropriately degraded properties in the evaluation of identified defects. The integrated program is implemented by the applicant through an inspection plan that is submitted to the US NRC for review and approval with the application for license renewal.

Evaluation and Technical Basis

1. Scope of Program: The scope of the program includes all RVI components at the [as an administrative action item for the AMP, the applicant to fill in the name of the applicant’s nuclear facility, including applicable units], which [is/are] built to a [applicant to fill in Westinghouse, CE, or B&W, as applicable] NSSS design. The scope of the program applies the methodology and guidance in the most recently US NRC-endorsed version of MRP-227(Archived), which provides 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 B&W, CE, and Westinghouse. The scope of components considered for inspection under MRP-227(Archived) guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code, Section XI), those RVI components that serve an intended license renewal safety function pursuant to criteria in 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). The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review (AMR), as defined by the criteria set in 10 CFR 54.21(a)(1). 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 adequately managed in accordance with an applicant’s AMP that corresponds to GALL AMP XI.M1, “ASME Code, Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.”

The scope of the program includes the response bases to applicable license renewal applicant action items (LRAAIs) on the MRP-227(Archived) methodology, and any additional programs, actions, or activities that are discussed in these LRAAI responses and credited for aging management of the applicant’s RVI components. The LRAAIs are identified in the staff’s safety evaluation on MRP-227(Archived) and include applicable action items on meeting those assumptions that formed the basis of the MRP’s augmented inspection and flaw evaluation methodology (as discussed in Section 2.4 of MRP-227(Archived)), and NSSS vendor-specific or plant-specific LRAAIs as well. The responses to the LRAAIs on MRP-227(Archived) are provided in Appendix C of the LRA.

The guidance in MRP-227(Archived) specifies applicability limitations to base-loaded plants and the fuel loading management assumptions upon which the functionality analyses were based. These limitations and assumptions require a determination of applicability by the applicant for each reactor and are covered in Section 2.4 of MRP-227(Archived).

2. Preventive Actions: The guidance in MRP-227(Archived) 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. The program description, evaluation, and technical basis of water chemistry are presented in GALL AMP XI.M2, “Water Chemistry.”

3. Parameters Monitored/Inspected: The program manages the following age-related degradation effects and mechanisms that are applicable in general to the RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimension due to void swelling and irradiation growth, distortion, or deflection; and (e) loss of preload caused by 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-destruction examination (NDE) method, or for relevant flaw presentation signals if a volumetric 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, or by void swelling and irradiation growth; instead, the impact of loss of fracture toughness on component integrity is indirectly managed by using visual or volumetric examination techniques to monitor for cracking in the components and by applying applicable reduced fracture toughness properties in the flaw evaluations if cracking is detected in the components and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation under the MRP-227(Archived) guidance or ASME Code, Section XI requirements. The program uses physical measurements to monitor for any dimensional changes due to void swelling, irradiation growth, distortion, or deflection.

Specifically, the program implements the parameters monitored/inspected criteria for [as an administrative action item for the AMP, applicant is to select one of the following to finish the sentence, as applicable to its NSSS vendor for its internals: “for B&W designed Primary Components in Table 4-1 of MRP-227(Archived)”; “for CE designed Primary Components in Table 4-2 of MRP-227(Archived)”; and “for Westinghouse designed Primary Components in Table 4-3 of MRP- 227(Archived)”]. Additionally, the program implements the parameters monitored/inspected criteria for [as an administrative action item for the AMP, applicant is to select one of the following to finish the sentence, as applicable to its NSSS vendor for its internals: “for B&W designed Expansion Components in Table 4-4 of MRP-227(Archived)”; “for CE designed Expansion Components in Table 4-5 of MRP-227(Archived)”; and “for Westinghouse designed Expansion Components in Table 4-6 of MRP-227(Archived)”]. The parameters monitored/inspected for Existing Program Components follow the bases for referenced Existing Programs, such as the requirements for ASME Code Class RVI components in ASME Code, Section XI, Table IWB-2500-1, Examination Categories B-N-3, as implemented through the applicant’s ASME Code, Section XI program, or the recommended program for inspecting Westinghouse-designed flux thimble tubes in GALL AMP XI.M37, “Flux Thimble Tube Inspection.” No inspections, except for those specified in ASME Code, Section XI, are required for components that are identified as requiring “No Additional Measures,” in accordance with the analyses reported in MRP-227(Archived).

4. Detection of Aging Effects: The detection of aging effects is covered in two places: (a) the guidance in Section 4 of MRP-227(Archived) provides an introductory discussion and justification of the examination methods selected for detecting the aging effects of interest; and (b) standards for examination methods, procedures, and personnel are provided in a companion document, MRP-228(Archived). In all cases, well-established methods were selected. These methods include volumetric UT examination methods for detecting flaws in bolting, physical measurements for detecting changes in dimension, 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). The VT-3 visual methods may be applied for the detection of cracking only when the flaw tolerance of the component or affected assembly, as evaluated for reduced fracture toughness properties, is known and has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. 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.

In addition, the program adopts the recommended guidance in MRP-227(Archived) for defining the Expansion criteria that need to be applied to inspections of Primary Components and Existing Requirement Components and for expanding the examinations to include additional Expansion Components. As a result, inspections performed on the RVI components are performed consistent with the inspection frequency and sampling bases for Primary Components, Existing Requirement Components, and Expansion Components in MRP-227(Archived), which have been demonstrated to be in conformance with the inspection criteria, sampling basis criteria, and sample Expansion criteria in Section A.1.2.3.4 of US NRC Branch Position RLSB-1.

Specifically, the program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for [as an administrative action item for the AMP, applicant is to select one of the following to finish the sentence, as applicable to its NSSS vendor for its internals: “B&W designed Primary Components in Table 4-1 of MRP-227(Archived)”; “CE designed Primary Components in Table 4-2 of MRP-227(Archived)”; and “Westinghouse designed Primary Components in Table 4-3 of MRP- 227(Archived)”] and for [as an administrative action item for the AMP, applicant is to select one of the following to finish the sentence, as applicable to its NSSS vendor for its internals: “for B&W designed Primary Components in Table 4-4 of MRP-227(Archived)”; “for CE designed Primary Components in Table 4-5 of MRP-227(Archived)”; and “for Westinghouse designed Primary Components in Table 4-6 of MRP- 227(Archived)”]

The program is supplemented by the following plant-specific Primary Component and Expansion Component inspections for the program (as applicable):[As a relevant license renewal applicant action item, the applicant is to list (using criteria in MRP-227(Archived)) each additional RVI component that needs to be inspected as an additional plant-specific Primary Component for the applicant’s program and each additional RVI component that needs to be inspected as an additional plant-specific Expansion Component for the applicant’s program. For each plant specific component added as an additional primary or Expansion Component, the list should include the applicable aging effects that will be monitored for, the inspection method or methods used for monitoring, and the sample size and frequencies for the examinations].

In addition, in some cases (as defined in MRP-227(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 dimension due to void swelling, deflection or distortion. The physical measurements methods applied in accordance with this program include [Applicant to input physical measure methods identified by the MRP in response to US NRC RAI No. 11 in the US NRC’s Request for Additional Information to Mr. Christen B. Larson, EPRI MRP on Topical Report MRP-227(Archived) dated November 12, 2009].

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(Archived) and its subsections. The evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as well for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications. The examinations and reexaminations required by the MRP-227(Archived) guidance, together with the requirements specified in MRP-228(Archived) for inspection methodologies, inspection procedures, and inspection personnel, provide timely detection, reporting, and corrective actions with respect to the effects of the age-related degradation mechanisms within the scope of the program. The extent of the examinations, beginning with the sample of susceptible PWR internals component locations identified as Primary Component locations, with the potential for inclusion of Expansion Component locations if the effects are greater than anticipated, plus the continuation of the Existing Programs activities, such as the ASME Code, Section XI, Examination Category B-N-3 examinations for core support structures, provides a high degree of confidence in the total program.

6. Acceptance Criteria: Section 5 of MRP-227(Archived) provides specific examination acceptance criteria for the Primary and Expansion Component examinations. For components addressed by examinations referenced to ASME Code, Section XI, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Program reference document.

The guidance in MRP-227(Archived) contains three types of examination acceptance criteria:

• For visual examination (and surface examination as an alternative to visual examination), the examination acceptance criterion is the absence of any of the specific, descriptive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that are detected and sized for length by VT-1/EVT-1 examinations;

• For volumetric examination, the examination acceptance criterion is the capability for reliable detection of indications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for system-level assessment of bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and

• For physical measurements, the examination acceptance criterion for the acceptable tolerance in the measured differential height from the top of the plenum rib pads to the vessel seating surface in B&W plants are given in Table 5-1 of MRP-227(Archived). The acceptance criterion for physical measurements performed on the height limits of the Westinghouse-designed hold-down springs are [The incorporation of this sentence is a license renewal applicant action item for Westinghouse PWR applicants only – insert the applicable sentence incorporating the specified physical measurement criteria only if the applicant’s facility is based on a Westinghouse NSSS design: the Westinghouse applicant is to incorporate the applicable language and then specify the fit up limits on the hold down springs, as established on a plant-specific basis for the design of the holddown springs at the applicant’s Westinghouse-designed facility].

7. Corrective Actions: Corrective actions following the detection of unacceptable conditions are fundamentally provided for in each plant’s corrective action 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. Examples of methodologies that can be used to analytically disposition unacceptable conditions are found in the ASME Code, Section XI or in Section 6 of MRP-227(Archived). Section 6 of MRP-227(Archived) describes the options that are available for disposition of detected conditions that exceed the examination acceptance criteria of Section 5 of the report. These include engineering evaluation methods, as well as supplementary examinations to further characterize the detected condition, or the alternative of component repair and replacement procedures. The latter are subject to the requirements of the ASME Code, Section XI. The implementation of the guidance in MRP-227(Archived), 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 action bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the US NRC. Examples of previously US NRC-endorsed alternative corrective actions bases include those corrective actions bases for Westinghouse-design RVI components that are defined in Tables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of Westinghouse Report No. WCAP-14577-Rev. 1-A, or for B&W-designed RVI components in B&W Report No. BAW-2248. Westinghouse Report No. WCAP-14577-Rev. 1-A was endorsed for use in an US NRC SE to the Westinghouse Owners Group, dated February 10, 2001. B&W Report No. BAW-2248 was endorsed for use in an SE to Framatome Technologies on behalf of the B&W Owners Group, dated December 9, 1999. Alternative corrective action bases not approved or endorsed by the US NRC will be submitted for US NRC approval prior to their implementation.

8. Confirmation Process: Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable. It is expected that the implementation of the guidance in MRP-227(Archived) will provide an acceptable level of quality for inspection, flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with the 10 CFR Part 50, Appendix B, or their equivalent (as applicable), confirmation process, and administrative controls.

9. Administrative Controls: The administrative controls for such programs, including their implementing procedures and review and approval processes, are under existing site 10 CFR Part 50 Appendix B Quality Assurance Programs, or their equivalent, as applicable. Such a program is thus expected to be established with a sufficient level of documentation and administrative controls to ensure effective long-term implementation.

10. Operating Experience: Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. A summary of observations to date is provided in Appendix A of MRP-227-A(Archived). The applicant is expected to review subsequent operating experience for impact on its program or to participate in industry initiatives that perform this function.

The application of the MRP-227(Archived) guidance will establish a considerable amount of operating experience over the next few years. Section 7 of MRP-227(Archived) describes the reporting requirements for these applications, and the plan for evaluating the accumulated additional operating experience.

References

10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, Office of the Federal Register, National Archives and Records Administration, 2009.

10 CFR Part 50.55a, Codes and Standards, Office of the Federal Register, National Archives and Records Administration, 2009.

ASME Boiler & Pressure Vessel Code, Section V, Nondestructive Examination, 2004 Edition, American Society of Mechanical Engineers, New York, NY.

ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, The ASME Boiler and Pressure Vessel Code, 2004 edition as approved in 10 CFR 50.55a, The American Society of Mechanical Engineers, New York, NY.

B&W Report No. BAW-2248, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, Framatome Technologies (now AREVA Technologies), Lynchburg VA, July 1997. (US NRC Microfiche Accession Number A0076, Microfiche Pages 001 - 108).

EPRI 1014986(Archived), PWR Primary Water Chemistry Guidelines, Volume 1, Revision 6, Electric Power Research Institute, Palo Alto, CA, December 2007. (Non-publicly available ADAMS Accession Number ML081140278). The non-proprietary version of the report may accessed by members of the public at ADAMS Accession Number ML081230449

EPRI 1016596(Archived), Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), Electric Power Research Institute, Palo Alto, CA: 2008.

EPRI 1016609(Archived), Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228), Electric Power Research Institute, Palo Alto, CA, July 2009. (Non-publicly available ADAMS Accession Number ML092120574). The non-proprietary version of the report may accessed by members of the public at ADAMS Accession Number ML092750569.

US NRC RAI No. 11 in the NRC’s Request for Additional Information to the Mr. Christen B. Larson, EPRI MRP on Topical Report MRP-227(Archived) dated November 12, 2009.

US NRC Safety Evaluation from C. I. Grimes [US NRC] to R. A, Newton [Chairman, Westinghouse Owners Group], Acceptance for Referencing of Generic License Renewal Program Topical Report Entitled “License Renewal Evaluation: Aging Management for Reactor Internals,” WCAP-14577, Revision 1, February 10, 2001. (ADAMS Accession Number ML010430375).

US NRC Safety Evaluation from C. I. Grimes [US NRC] to W. R. Gray [Framatome Technologies], Acceptance for Referencing of Generic License Renewal Program Topical Report Entitled “Demonstration of the Management of Aging Effects for the Reactor Vessel Internals,” February 10, 2001. (ADAMS Accession Number ML993490288).

NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Appendix A.1, “Aging Management Review - Generic (Branch Technical Position RLSB-1),” U.S. Nuclear Regulatory Commission, Washington, DC, 2010.

Westinghouse Non-Proprietary Class 3 Report No. WCAP-14577-Rev. 1-A, License Renewal Evaluation: Aging Management for Reactor Internals, Westinghouse Electric Company, Pittsburgh, PA [March 2001]. Report was submitted to the US NRC Document Control Desk in a letter dated April 9, 2001. (ADAMS Accession Number ML011080790).

Pressurized Water Reactor

Electric Power Research Institute

Stress Corrosion Cracking

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

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

Aging Management Review

License Renewal Application - The document submitted by the plant licensee under 10 CFR 54 to extend the operating license of the plant from 40 to 60 years.

Non Destructive Examination

Enhanced Visual Test

Request for additional information

Safety Evaluation

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

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