ML19178A339

Staff Assessment of Aging Management Program and Inspection Plan of Reactor Vessel Internals

ML19178A339
Person / Time
Site:	Wolf Creek
Issue date:	07/09/2019
From:	Robert Pascarelli Plant Licensing Branch IV
To:	Heflin A Wolf Creek
Pascarelli R
References
EPID L-2019-LRO-0013
Download: ML19178A339 (20)

Text

July 9, 2019 Mr. Adam C. Heflin President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION, UNIT 1 - STAFF ASSESSMENT OF AGING MANAGEMENT PROGRAM AND INSPECTION PLAN OF REACTOR VESSEL INTERNALS (EPID L-2019-LRO-0013)

Dear Mr. Heflin:

By letter dated October 25, 2017 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML17305A042), Wolf Creek Nuclear Operating Corporation (the licensee) submitted its Aging Management Program (AMP) and Inspection Plan for the Wolf Creek Generating Station, Unit 1 (Wolf Creek), Reactor Vessel Internals (RVIs). The Wolf Creek RVI AMP and Inspection Plan is based on the U.S. Nuclear Regulatory Commission (NRC)-approved topical report Material Reliability Program (MRP)-227-A Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), dated January 9, 2012 (ADAMS Accession No. ML120170453). The licensee submitted its RVI AMP and Inspection Plan to fulfill Commitment No. 19B in NUREG-1915, Safety Evaluation Report Related to the License Renewal of Wolf Creek Generating Station, dated October 2008 (ADAMS Accession No. ML083090483). The AMP includes inspection and evaluation guidelines for the RVI components at Wolf Creek, during the period of extended operation.

The NRC staff has completed its review of the Wolf Creek RVI AMP and Inspection Plan and concludes that it is consistent with the inspection and evaluation guidelines of MRP-227-A and is, therefore, acceptable. The NRC staff finds that the licensee has adequately addressed and resolved all the applicable applicant/licensee action items specified in MRP-227-A.

The NRC staffs approval of the Wolf Creek RVI AMP and Inspection Plan does not reduce, alter, or otherwise affect the current American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, inservice inspection requirements, or any Wolf Creek, specific licensing basis requirements related to inservice inspection.

A. Heflin The NRCs staff assessment of the Wolf Creek RVI AMP and Inspection Plan is enclosed. If you have any questions, please contact me at 301-415-3016 or via e-mail at Balwant.Singal@nrc.gov.

Sincerely,

/RA/

Robert J. Pascarelli, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure Staff Assessment cc: Listserv

ML19178A339 via email dated June 13, 2019 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DMLR/MVIB/BC*

NAME BSingal PBlechman DAlley DATE 07/08/19 07/08/19 06/13/19 OFFICE NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME RPascarelli BSingal DATE 07/09/19 07/09/19 STAFF ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION AGING MANAGEMENT PROGRAM AND INSPECTION PLAN OF REACTOR VESSEL INTERNALS WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION, UNIT 1 DOCKET NO. 50-482

1.0 INTRODUCTION

By letter dated October 25, 2017 (Reference 1), Wolf Creek Nuclear Operating Corporation (WCNOC or the licensee) submitted its Aging Management Program (AMP) and Inspection Plan for the Wolf Creek Generating Station, Unit 1 (Wolf Creek), Reactor Vessel Internals (RVIs).

The Wolf Creek RVI AMP and Inspection Plan is based on the U.S. Nuclear Regulatory Commission (NRC)-approved topical report Material Reliability Program (MRP)-227-A Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), dated January 9, 2012 (Reference 2). The licensee submitted its RVI AMP and Inspection Plan to fulfill Commitment No. 19B in NUREG-1915, Safety Evaluation Report Related to the License Renewal of Wolf Creek Generating Station, dated October 2008 (Reference 3). The AMP includes inspection and evaluation (I&E) guidelines for the RVI components at Wolf Creek, during the period of extended operation (PEO).

2.0 REGULATORY EVALUATION

2.1 Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, addresses the requirements for the plant license renewal process. The regulations at 10 CFR 54.21, Contents of application - technical information, require that each application for license renewal contain an integrated plant assessment and an evaluation of time-limited aging analyses. The plant-specific integrated plant assessment shall identify and list those structures and components subject to an aging management review, and demonstrate that the effects of aging (e.g., cracking, loss of material, loss of fracture toughness, dimensional changes, and loss of preload) will be adequately managed so that their intended functions will be maintained consistent with the current licensing basis during the PEO, as required by 10 CFR 54.29(a). In addition, 10 CFR 54.22, Contents of application technical specifications, requires a license renewal application to include any technical specification changes or additions necessary to manage the effects of aging during the PEO, as part of the license renewal application.

Enclosure

Structures and components subject to an AMP shall encompass those structures and components that are referred to as passive and long-lived. Passive structures and components perform an intended function, as described in 10 CFR 54.4, Scope, without moving parts or without a change in configuration or properties. Long-lived structures and components are not subject to replacement based on a qualified life or specified time period.

The scope of components considered for inspection under MRP-227-A includes core support structures typically denoted as Examination Category B-N-3 by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, and those RVI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1). The scope of the program does not include consumable components 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 AMP, as defined by the criteria set forth in 10 CFR 54.21(a)(1).

2.2 Licensee Renewal Commitment The NRC issued NUREG-1915 in October 2008. Appendix A of NUREG-1915 included license renewal Commitment No. 19B related to the Wolf Creek RVI AMP. Specifically, the licensee committed to the following activities related to RVIs:

(1) [p]articipate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, WCNOC will submit an inspection plan for reactor internals to the NRC for review and approval.

On January 12, 2009, Electric Power Research Institute (EPRI) submitted, for NRC staff review and approval, MRP-227, Revision 0, which was intended as guidance for applicants in developing their plant-specific AMPs for RVI components (Reference 4). MRP-227 contains a discussion of the technical basis for the development of plant-specific AMPs for RVI components in pressurized-water reactor (PWR) vessels and, provides I&E guidelines for PWR applicants to use in their plant-specific AMPs. The NRC staff issued Revision 1 to its Final Safety Evaluation (SE) for MRP-227 on December 16, 2011 (Reference 5), with seven topical report (TR) conditions and eight applicant/licensee action items. The TR conditions were specified to ensure that certain information was revised generically in the approved version of MRP-227-A and the applicant/licensee action items addressed plant-specific issues that could not be resolved generically in the SE dated December 16, 2011. On January 9, 2012, EPRI submitted the NRC-approved version of the TR designated as MRP-227-A (Reference 2). The licensee developed the Wolf Creek RVI AMP and Inspection Plan based on MRP-227-A. To fulfill license renewal Commitment No. 19B, the licensee needs to address the plant-specific issues specified in the eight applicant/licensee action items.

3.0 TECHNICAL EVALUATION

The NRC staff assessment of the Wolf Creek RVI AMP and Inspection Plan focused on determining whether the licensee adequately incorporated the I&E guidelines in MRP-227-A and

the licensees resolutions of the eight action items. Specifically, the NRC staff assessment focused on the following:

The licensees implementation of the MRP-227-A I&E guidelines for RVI components in the primary, expansion, and existing categories, as well as the appropriate acceptance criteria; Operating experience (OE) of RVI component degradation at Wolf Creek; Ten-program elements of the Wolf Creek RVI AMP; and The licensees resolutions of the eight licensee action items.

The NRC staff noted that the seven TR conditions identified in the NRC staffs SE are generic conditions imposed on the approval of MRP-227 and were incorporated when MRP-227-A was published; thus, these seven TR conditions are not addressed by the individual licensees. The following sections provide details of the NRC staff assessment.

3.1 Assessment Area 1 MRP-227-A I&E Guidelines for RVI Components in the Primary, Expansion, and Existing Categories, and Acceptance Criteria In the Wolf Creek RVI AMP and Inspection Plan (Enclosure to the letter dated October 25, 2017), the licensee implemented the MRP-227-A I&E guidelines in the primary, expansion, and existing categories in the following tables with plant-specific notes for certain RVI components:

Table 2-1, Westinghouse Plants Primary Components Table 2-2, Westinghouse Plants Expansion Component Table 2-3, Westinghouse Plants Existing Program Components These tables were taken directly from the following tables from MRP-227-A:

Table 4-3, Westinghouse plants Primary components Table 4-6, Westinghouse plants Expansion component Table 4-9, Westinghouse plants Existing Programs components The NRC staff reviewed the Wolf Creek RVI components in the primary, expansion, and existing categories and the corresponding MRP-227-A I&E guidelines. Based on its review, the staff finds that the Wolf Creek RVI AMP and Inspection Plan is consistent with the inspection tables in MRP-227-A. Additionally, the NRC staff reviewed the acceptance criteria in Table 2-4, Westinghouse Plants Examination Acceptance and Expansion Criteria, of the Wolf Creek RVI AMP and Inspection Plan, and determined that the acceptance criteria are consistent with Table 5-3, Westinghouse plants examination acceptance and expansion criteria, of MRP-227-A. Based on this review, the NRC staff has determined that the licensees RVI AMP and Inspection Plan has been adequately implemented in accordance with the MRP-227-A I&E guidelines.

3.2 Assessment Area 2 Operating Experience of RVI Component Degradation 3.2.1 Wear of Control Rod Guide Tube Cards Appendix A of MRP-227-A includes a generic discussion of wear of the control rod guide tube (CRGT) cards. Wolf Creek RVI AMP and Inspection Plan, Table 2-1, indicates that for general degradation visual inspections (VT-3) will be used for the CRGT plates (cards). Also, Table 2-1 indicates the inspections will be performed per the examination coverage and schedule recommended in WCAP-17451-P, Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections (Reference 6). The NRC staffs Final SE of WCAP-17096-NP, Revision 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements, dated May 3, 2016 (Reference 7), contains acceptance criteria for MRP-227-A primary and expansion components and determines that the examination methodology for CRGT plates (cards), based on WCAP-17451-P, to be acceptable. WCAP-17451-P considers OE and analyses that were not available during the development of MRP-227-A, and includes a more comprehensive inspection coverage for the CRGT plates (cards) than the coverage specified in MRP-227-A. The NRC staff finds that the licensees use of WCAP-17451-P adequately addresses operating experience of wear in CRGT plates (cards). Therefore, the staff finds there is reasonable assurance that the Wolf Creek CRGT plates (cards) will be adequately managed for wear during the PEO.

3.2.2 Baffle-to-Former Bolts Degradation Recent OE associated with baffle-to-former bolts degradation originated from a larger-than-expected number of degraded baffle-former bolts discovered in four-loop down-flow plants through MRP-227 inspections and voluntary inspections. The industry performed evaluations and issued EPRI Letter MRP 2016-022, dated July 27, 2016, Transmittal of

[Nuclear Energy Institute] NEI-03-08 Needed Interim Guidance Regarding Baffle Former Bolt inspections for Tier 1 plants as defined in Westinghouse [Nuclear Safety Advisory Letter]

NSAL 16-01 (Reference 8), specifically, recommending all Tier 1a plants perform ultrasonic inspection of the full population of baffle-to-former bolts.

Wolf Creek has continuously operated in an up-flow configuration and is considered a Tier 4 plant per NSAL 16-01. The NRC staff noted that up flow plant configurations could reduce the baffle jetting damage to fuel and reduce the bolt loads under normal operating and faulted conditions; therefore, they are the least affected by this OE and have the lowest susceptibility to baffle-former bolt degradation. As discussed in the staffs assessment of EPRI NEI 03-08, Revision 2, Needed Interim Guidance Regarding Baffle-Former Bolt Inspections in Westinghouse-Design Pressurized Water Reactors (Reference 9), the staff determined that Tier 4 plants, such as Wolf Creek, will continue to follow the current MRP-227-A guidelines and implement any future revisions to the MRP-227 recommendations. Therefore, the staff finds the Wolf Creek RVI AMP and Inspection Plan provides reasonable assurance that the licensee will adequately manage the effects of aging for the Wolf Creek baffle-to-former bolts during the PEO.

3.2.3 Clevis Insert Bolts The licensee stated that the clevis insert bolts, which are constructed of X-750, are addressed as an existing program component in the Wolf Creek RVI AMP and are included in the Wolf Creek ASME Code,Section XI Program as B-N-2 components. During ASME Code,Section XI, 10-year inservice inspection (ISI) exams, these components are inspected by VT-3

inspections using a submersible mini-submarine. The licensee confirmed that during the ASME Code,Section XI, 10-year ISI inspection performed in 2005 (Refueling Outage (RFO) 14), of the clevis inserts, bolts, pins, and welded lock bars, no indications of age-related degradation were observed.

The licensee stated that it evaluated industry OE concerning failure of X-750 clevis insert bolts at another Westinghouse nuclear steam supply system design plant in 2010 under its corrective action program. In response to this OE, the licensee performed an opportunistic VT-3 inspection of the clevis inserts, bolts, pins, and welded lock bars in 2011 (RFO18). The results of this inspection were similar to the 2005 (RFO 14) inspection such that no evidence of age-related degradation was observed. The licensee explained that its clevis insert design differs from the Westinghouse nuclear steam supply system design plant that experienced the clevis insert bolt failure in 2010.

The NRC staff reviewed the following documents related the reactor internals lower radial support system and clevis insert bolts:

NRC000219 - Westinghouse InfoGram IG-10-1, Reactor Internals Lower Radial Support Clevis Insert Cap Screw Degradation, dated March 31, 2010 (Reference 10).

Pressurized Water Reactor Owners Group (PWROG) Presentation Slides, Industry and NRC Coordination Meeting Materials Programs Technical Exchange Clevis Insert Bolt Update, dated June 2014 (Reference 11).

WAAP-8828-P, Revision 0, Lower Radial Support System (LRSS) Clevis Inserts and Attachment Bolts Design and Safety Function, dated March 2014 (Reference 12).

Entergy letter dated September 27, 2013, Reply to Request for Additional Information Regarding the License Renewal Application, Indian Point Nuclear Generating Unit Nos. 2 & 3 (Reference 13).

Based on its review, the NRC staff noted the following key points that are generically applicable to all Westinghouse RVI LRSS and clevis insert bolts:

The main design function of the LRSS that contains the clevis insert bolts is the prevention of tangential or rotational motion of the lower internals assembly while permitting axial displacement and differential radial expansion. These supports are also designed to limit displacements and misalignments in order to avoid overstressing the core barrel and to ensure that the control rods can be freely inserted.

The main aging effect of concern is wear due to flow-induced vibration. Failure of cap screws could result in increased wear, which would occur over several cycles (as well as during seismic events and loss-of-coolant accident conditions) and does not impact the function of the LRSS. This is based on the OE described in the InfoGram IG-10-1 for the plant that had experienced clevis insert bolt failures.

There is a high degree of redundancy in the LRSS. Because of the small clearances involved, it is unlikely that complete disengagement of the clevis inserts would occur.

If one clevis insert became nonfunctional, the other lower radial supports are capable of resisting all of the internal and external asymmetric loads.

Crack detection before bolt failure is not required because of inherent design redundancy.

Westinghouse performed an evaluation of the potential for creation of loose parts (and damage from loose parts) caused by clevis insert bolt degradation and concluded that no significant degradation of mechanical components is expected as a result of potential loose parts in the primary system. This is because separated cap screw heads will remain captured in the clevis insert counterbores. Although lock bars experienced wear related degradation at the plant with the bolt failures, the potential for damage from loose lock bars is minimal.

The visual inspections performed using video cameras during each 10-year interval under ASME Code,Section XI, can identify wear or dislodged components of the clevis insert cap screws or dowel pins at any location, if they exist.

Based upon the key points discussed above, and the absence of any observed age-related degradation of the clevis insert bolts at Wolf Creek, the NRC staff finds that the current inspection method and frequency of the ASME Code,Section XI, ISI Program are adequate to maintain continued functionality of the clevis inserts during the PEO.

3.3 Assessment Area 3 - Reactor Vessel Internals Aging Management Program Elements Section 3.0, Aging Management Program Attributes, of the Wolf Creek RVI AMP and Inspection Plan states the attributes of the AMP are in compliance with NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report, dated December 2010 (Reference 14), AMP Chapter Xl.M16A, PWR Vessel Internals. The licensee provided its evaluation of the ten program elements against the corresponding elements in GALL Report AMP Chapter XI.M16A.

The NRC staff reviewed the licensees RVI AMP against the ten program elements of the GALL Report AMP Chapter XI.M16A, as revised by Final License Renewal Interim Staff Guidance (ISG) LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors (Reference 15). Based on its review, the staff determined that the ten program elements of the Wolf Creek RVI AMP are consistent with the ten program elements described in LR-ISG-2011-04 and that the Wolf Creek RVI AMP includes additional guidance in WCAP 17451-P due to OE related to wear of CRGT plates (cards).

Therefore, the NRC staff finds the licensees implementation of the ten program elements acceptable for Wolf Creek.

3.4 Assessment Area 4 - Reponses to Licensee Action Items Applicant/Licensee Action Item 1 Section 4.2.1, Applicability of FMECA [Failure Modes, Effects, and Criticality Analyses] and Functionality Analysis Assumptions, of the SE for MRP-227-A (Reference 5), states:

As addressed in Section 3.2.5.1 of this SE, each applicant/licensee is responsible for assessing its plants design and operating history and demonstrating that the approved version of MRP-227 is applicable to the facility.

Each applicant/licensee shall refer, in particular, to the assumptions regarding plant design and operating history made in the FMECA and functionality analyses for reactors of their design (i.e., Westinghouse, CE [Combustion Engineering], or B&W [Babcock and Wilcox]) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-227. This is Applicant/Licensee Action Item 1.

The purpose of Licensee Action Item 1 is to determine the applicability of the I&E guidelines of MRP-227-A to the specific plant, in this case, to Wolf Creek. Following discussions between EPRI and the NRC staff, the basis for a plant to respond to the NRC staff questions to demonstrate compliance with MRP-227-A for originally licensed and uprated conditions was developed (References 16 and 17). It was determined that Licensee Action Item 1 would be addressed with plant-specific responses to the following two questions:

Question 1: Does the plant have non-weld or bolting austenitic stainless steel (SS) components with 20 percent cold work or greater, and, if so, do the affected components have operating stresses greater than 30 ksi [kilo-pounds per square inch]? (If both conditions are true, additional components may need to be screened in for stress corrosion cracking (SCC).

Question 2: Does the plant have an atypical fuel design or fuel management that could render the assumptions of MRP-227-A, regarding core loading/core design, non-representative for that plant?

To generically resolve Question 1, the PWROG developed PWROG-15105-NP, Revision 0, PA-MSC-1288 PWR RV Internals Cold-Work Assessment (Reference 18), which was submitted to the NRC for information by letter dated June 15, 2016 (Reference 19).

The NRC Staff Assessment of PWROG-15105-NP, Revision 0, dated April 21, 2017 (Reference 20), concluded the following:

The majority of austenitic SS materials were required to be solution annealed, which eliminates the possibility of effects from cold work on the SCC behavior of the materials.

Some of the material specifications stipulate limitations on the maximum allowed tensile strength and hardness values, which restricts the possible amount of cold work in the component.

No non-fastener RVI components were subject to cold work greater than twenty percent in PWR units, and these components are less susceptible to SCC.

Material specification and design with respect to the consideration of cold work in CE and Westinghouse non-fastener RVI components did not change over the years of construction of the PWR fleet. Since cold work on these RVI components was acceptably controlled during the construction period, it is concluded that non-fastener RVI components from unassessed Westinghouse and CE plants have low cold work and limited susceptibility to SCC.

Based on the above conclusions, the NRC staff determined in its SE dated January 29, 2018 (Reference 21), that a plant-specific response to Question 1 is no longer necessary. For Question 2, the staff determined in its SE dated January 29, 2018, that plant-specific information for core-design related parameters described in MRP Letter 2013-0251 should be addressed in the plant-specific RVI program plan.

In response to Question 2, the licensee provided the following confirmations for Wolf Creek:

The plant went to a low leakage core starting in cycle 4 and had a full low leakage core at the start of cycle 7.

The plant typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule.

The licensee confirmed that the reactor coolant system operates between Tcold (555.8 degrees Fahrenheit (°F)) and Thot (621.1 °F) and the design temperature for the vessel is 650 °F.

Based on these confirmations, the NRC staff finds the licensee satisfactorily addressed the assumptions in MRP-1912 and MRP-227-A regarding operational history, fluence and fuel management assumptions. The licensee explained that it also performed an evaluation for Wolf Creek of its fuel design and fuel management in accordance with the guidance provided in MRP Letter 2013-025. The results of that evaluation confirmed that Wolf Creek has not utilized atypical fuel designs or fuel management that could invalidate the assumptions in MRP-227-A regarding core loading/core design non-representative, including, power changes/uprates that have occurred over the operating lifetime of the unit. The staff noted that the licensees conclusion is based on comparisons of the Wolf Creek core geometry and operating characteristics with the MRP-227-A applicability guidelines for Westinghouse-designed reactors specified in MRP Letter 2013-025.

1 EPRI, MRP-227-A Applicability Template Guideline, MRP Letter 2013-025, October 14, 2013 (Reference 22).

2 MRP-191, Revision 0, Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs (Reference 23).

Based on the confirmation provided by the licensee and the evaluation performed consistent with MRP Letter 2013-025, the NRC staff concludes that the licensee adequately addressed Action Item 1; therefore, this action item is resolved.

Applicant/Licensee Action Item 2 Section 4.2.2, PWR Vessel Internal Components Within the Scope of License Renewal, of the SE for MRP-227-A, states:

As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressed in 10 CFR 54.4, each applicant/licensee is responsible for identifying which RVI components are within the scope of LR [license renewal] for its facility.

Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-189[3], Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the RVI components that are within the scope of LR for their facilities in accordance with 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for its facility, the applicant or licensee shall identify the missing component(s) and propose any necessary modifications to the program defined in MRP-227, as modified by this SE, when submitting its plant-specific AMP. The AMP shall provide assurance that the effects of aging on the missing component(s) will be managed for the period of extended operation. This is Applicant/Licensee Action Item 2.

The licensee stated that the Wolf Creek RVI components were compared to those components contained in Table 4-4 of MRP-191. Based on a comparison of the plant-specific components to Table 4-4 of MRP-191, the components required to be in the Wolf Creek RVI AMP are mostly consistent with those contained in MRP-191, with the exceptions of those described in the licensees submittal.

Specifically, the NRC staff noted that the upper instrumentation conduit and support parts, lower support column bolts, neutron panels lock caps, and the radial support bolts are fabricated from different grades of austenitic SS than what are listed in the generic evaluation of MRP-191. In addition, the upper support column bases are fabricated from 304 SS, rather than the CF8 material listed in MRP-191. The staff reviewed the material differences for the components that the licensee identified in its response to Applicant/Licensee Action Item 2. For plant-specific material differences within the wrought product family, the staff confirmed that the difference between plant-specific Type 302 SS, 304 SS, 304L SS, and 316 SS materials and generic Type 304 SS and 316 SS, does not have any impact on these components screening results for the eight age-related degradation mechanisms addressed in MRP-191. Therefore, the staff determined that these material differences will not affect the FMECA results that underlie the MRP-227-A I&E guidelines for these components. Accordingly, the staff determined that the FMECA grouping and MRP-227-A inspection criteria remain the same for the plant-specific components that are of a different wrought material type than the wrought material type analyzed in MRP-191.

In MRP-191, CF8 materials were screened-in for the aging mechanism of thermal embrittlement, along with other applicable degradation mechanisms. The licensee stated that 3

MRP-189, Revision 1, Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals (Reference 24).

the upper support column bases are 304 SS. The NRC staff noted that this material difference reduces the number of applicable degradation mechanisms for the upper support column bases at Wolf Creek. Specifically, the degradation mechanisms of thermal and irradiation embrittlement are not applicable for the Wolf Creek upper support column bases because of the fabrication material, consistent with MRP-227-A. Based on the above, the NRC staff concludes that the licensee adequately addressed Action Item 2, and no changes to the methodology of MRP-227-A are needed to account for the plant-specific material differences of the Wolf Creek RVIs. Based on the above, the NRC staff concludes that the licensee adequately addressed Action Item 2; therefore, this action item is resolved.

Applicant/Licensee Action Item 3 Section 4.2.3, Evaluation of the Adequacy of Plant-Specific Existing Programs, of the SE for MRP-227-A, states:

As addressed in Section 3.2.5.3 in this SE, applicants/licensees of CE and Westinghouse are required to perform plant-specific analysis either to justify the acceptability of an applicants/licensees existing programs, or to identify changes to the programs that should be implemented to manage the aging of these components for the period of extended operation. The results of this plant-specific analysis and a description of the plant-specific programs being relied on to manage aging of these components shall be submitted as part of the applicants/licensees AMP application. The CE and Westinghouse components identified for this type of plant-specific evaluation include: CE thermal shield positioning pins and CE in-core instrumentation thimble tubes (Section 4.3.2 in MRP-227, Revision 0), and Westinghouse guide tube support pins (split pins)

(Section 4.3.3 in MRP-227, Revision 0). This [issue] is Applicant/Licensee Action Item 3.

Water Chemistry Program The licensee stated that its Water Chemistry Program is credited for controlling the levels of corrosive contaminants in the Primary Water System, thereby preventing or mitigating cracking of RVI components by SCC and irradiation assisted stress corrosion cracking. The water chemistry program does not include any inspections of RVI components. The NRC staff noted that the program includes specifications for chemical species, impurities and additives, sampling and analysis frequencies, and corrective actions for control of reactor water chemistry. The methods for aging management relies on (1) limiting the concentration of chemical species known to cause corrosion and (2) adding chemical species known to inhibit degradation by their influence on pH (a measure of acidity or alkalinity of water soluble substances) and dissolved oxygen levels.

The NRC staff concluded in NUREG-1915 that the applicant has demonstrated that the Water Chemistry Program will adequately manage aging effects identified for components in contact with reactor coolant, such that there is reasonable assurance that their intended functions will be maintained consistent with the current licensing basis for the PEO.

Flux Thimble Tube Inspection Program The licensee stated that its Flux Thimble Tube Inspection Program performs wall thickness eddy current testing of all flux thimble tubes that form part of the reactor coolant system pressure

boundary. During each refueling outage, flux thimble tube wear is evaluated, and inspections are performed based upon evaluation results. Specifically, if the predicted wear for a given flux thimble tube is projected to exceed the established acceptance criteria prior to the next scheduled refueling outage, corrective actions are taken to reposition, cap or replace the tube.

The NRC staff concluded in NUREG-1915 that the applicant has demonstrated that the Flux Thimble Tube Inspection Program will adequately manage aging effects identified for the flux thimble tubes, such that there is reasonable assurance that their intended functions will be maintained consistent with the current licensing basis for the PEO.

ASME Code,Section XI, Inservice Inspection (ISI) Program ASME Code,Section XI, IWB-2500, provides inspection requirements for B-N-2 (Welded Core Support Structures and Interior Attachments to Reactor Vessels) and B-N-3 (Removable Core Support Structures). Visual inspections (VT-3) of the applicable components accessible surfaces are required one time per interval with relevant conditions requiring correction, as described in ASME Code,Section XI.

The NRC staff concluded in NUREG-1915 that the applicant has demonstrated that the ASME Code,Section XI, ISI program will adequately manage aging effects identified for B-N-2 components (Welded Core Support Structures and Interior Attachments to Reactor Vessels) and B-N-3 components (Removable Core Support Structures), such that there is reasonable assurance that their intended functions will be maintained consistent with the current licensing basis for the PEO.

Control Rod Guide Tube (CRGT) Support Pin (Split) Replacement The licensee stated that the original CRGT support pins were fabricated from X-750 (Rev. A heat treatment) that was later replaced prior to commercial operation, per recommendations from Westinghouse, with X-750 (Rev. B heat treatment). In 2002, Wolf Creek suffered a failure of an X-750 (Rev. B) support pin that was attributed to SCC. Westinghouse provided replacement support pins constructed of cold worked 316 SS that were installed in 2003. The licensee explained that the following routine activities are performed that can detect age-related degradation of the CRGT support pins.

During every refueling outage a full core offload is performed and a foreign object search and retrieval inspection of the reactor vessel is conducted prior to reloading the core.

A visual inspection is conducted inside the steam generators primary channel heads during outages when eddy current testing of the steam generator tubes is performed, which would detect the presence of support pin fragments.

The CRGTs and locking devices support pins are included in the Wolf Creek ISI Program (B-N-3 components) requiring a VT-3 inspection during the 10-year ISI exams.

Should additional recommendations for CRGT support pin replacement be issued by Westinghouse, Wolf Creek will take appropriate actions.

Table 3-3, Final disposition of Westinghouse internals, of MRP-227-A (Reference 2), identifies only Alloy X-750 CRGT support pins as requiring monitoring for aging during the PEO. Further, Subsection 4.4.3, Westinghouse Components, of MRP-227-A states that subsequent performance monitoring of the support pins should follow the recommendations of the original equipment manufacturer. The NRC staff finds that the licensee followed the recommendation of the original equipment manufacturer for plant-specific replacement with Type 316 SS CRGT support pins, consistent with the guidance in Subsection 4.4.3 of MRP-227-A. Furthermore, the staff noted that Type 316 SS CRGT support pins are in Category A in MRP-191 and binned into the No Additional Measures category (i.e., components determined to need no additional aging management). In addition, the staff finds the routine activities and inspections performed by the licensee during refueling outages are capable and adequate to detect age-related degradation of the CRGT support pins. Therefore, the staff determined there is reasonable assurance that the licensee will adequately manage the aging of the Wolf Creek CRGT support pins during the PEO.

Based on its review of the Water Chemistry Program, Flux Thimble Tube Inspection Program, ASME Code,Section XI ISI Inspection Program, and CRGT split pin replacements, as described above, the NRC staff concludes that the licensee adequately addressed Action Item 3; therefore, this action item is resolved.

Applicant/Licensee Action Item 4 and Applicant/Licensee Action Item 6 MRP-227-A Licensee Action Item 4 and Licensee Action Item 6 are only applicable to plants designed by B&W; therefore, Item 4 and Item 6 are not applicable to Wolf Creek.

Applicant/Licensee Action Item 5 Section 4.2.5, Application of Physical Measurements as part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components, of the SE for MRP-227-A, states:

As addressed in Section 3.3.5 in this SE, applicants/licensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRC-approved version of MRP-227 for loss of compressibility for Westinghouse hold down springs, and for distortion in the gap between the top and bottom core shroud segments in CE units with core barrel shrouds assembled in two vertical sections. The applicant/licensee shall include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are consistent with the plants licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 5.

The licensee stated in its response to Licensee Action Item 5 that the Wolf Creek hold-down spring is constructed of 403 SS, which is not susceptible to loss of preload due to stress relaxation. The NRC staff noted that loss of load is applicable to more susceptible material (i.e., Type 304 SS) than Type 403 SS, which has higher yield strength. Since Type 403 springs is used at Wolf Creek, no physical measurements of the hold-down spring are provided. Since Type 403 SS hold-down spring material has superior resistance to loss of load compared to Type 304 SS material, the staff determined that physical measurements for loss of compressibility for Type 403 SS hold-down springs in Wolf Creek are not necessary. Based on

the above, the NRC staff concludes that the licensee adequately addressed Action Item 5; therefore, this action item is resolved.

Applicant/Licensee Action Item 7 Section 4.2.7, Plant-Specific Evaluation of CASS [Cast Austenitic Stainless Steel] Materials, of the SE for MRP-227-A, states:

As discussed in Section 3.3.7 of this SE, the applicants/licensees of B&W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B&W IMI [incore monitoring instrumentation] guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operation or for additional RVI components that may be fabricated from CASS, martensitic stainless steel or precipitation hardened stainless steel materials. These analyses shall also consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. The requirement may not apply to components that were previously evaluated as not requiring aging management during development of MRP-227. That is, the requirement would apply to components fabricated from susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plants licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licensee shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 7.

The licensee confirmed in its response to Licensee Action Item 7 that only the lower internals assembly - bottom mounted instrumentation (BMI) column cruciform as being constructed of CASS, and that MRP-191 screened all CASS BMI cruciforms in for thermal embrittlement without concern for their ferrite content.

The NRC staff noted that Table 3-3 of MRP-227-A identified that CASS BMI column cruciforms are a No Additional Measures component. Regardless, the licensee stated that it conducted a search of the manufacturing records and located the certified material test reports for all 26 BMI column cruciforms. These certified material test reports indicate that the cruciform were statically cast and manufactured in accordance with SA-351, Grade CF8. Based on the information from the certified material test reports, the licensee determined that the calculated ferrite contents for all 26 BMI column cruciforms were below the 20 percent threshold for thermal embrittlement susceptibility of static castings as described in an NRC letter dated May 19, 2000, License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel Components (Reference 25).

The licensee stated that it identified only the hold-down spring as being constructed of martensitic 403 SS. The 403 SS hold-down spring was screened in for thermal embrittlement but was also categorized as a No Additional Measures component in MRP-191 and MRP-227-A. Finally, the licensee stated that it has no precipitation hardened SS RVI

components. The NRC staff reviewed Section 4.5.2, Reactor Internals Materials, and Table 5.2-4, Reactor Vessel Internals for Emergency Core Cooling Systems, of the Wolf Creek Updated Safety Analysis Report (UFSAR) (Reference 26) and confirmed that there is no precipitation hardened SS RVI components at Wolf Creek.

The CASS BMI cruciforms and Type 403 SS hold-down springs, as previously discussed, are categorized as No Additional Measures components in MRP-227-A. The NRC staff noted that Action Item 7 does not apply to components evaluated as No Additional Measures components and not requiring aging management as identified in MRP-227-A. Furthermore, based on the licensees confirmation and the staffs review of the UFSAR, it has been verified that Wolf Creek has no precipitation hardened SS RVI components. Based on the above, the NRC staff concludes that the licensee adequately addressed Action Item 7; therefore, this action item is resolved.

Applicant/Licensee Action Item 8 Section 4.2.8, Submittal of Information for Staff Review and Approval, of the SE for MRP-227-A, states:

As addressed in Section 3.5.1 of this SE, applicants/licensees shall make a submittal for NRC review and approval to credit their implementation of MRP-227, as amended by this SE, as an AMP for the RVI components at their facility. This submittal shall include the information identified in Section 3.5.1 of this SE. This is Applicant/Licensee Action Item 8.

The licensee stated that during the license renewal process, it committed to submit an inspection plan for reactor internals to the NRC for review and approval not less than 24 months prior to entering the PEO (March 12, 2025). The NRC staff noted that the licensees submittal serves as the Wolf Creek RVI Inspection Plan and provides a summary of the ten program elements of the RVI AMP in Sections 1.0 to 3.0, and the responses to the eight Licensee Action Items in Section 4.0.

The NRC staff noted that Wolf Creek falls under Category B per NRC Regulatory Issue Summary (RIS) 2011-07, License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management (Reference 27), which reminds license renewal holders and applicants that Category B plants must submit RVI AMPs/inspection plans in accordance with MRP-227-A. As described in Section 3.3 of this assessment, the NRC staff confirmed that the Wolf Creek RVI AMP and Inspection Plan addresses the ten program elements of the GALL Report and that the primary, expansion, and credited components identified in the Wolf Creek RVI AMP are consistent with MRP-227-A. Based on the above, the NRC staff concludes that the licensee adequately addressed Action Item 8; therefore, this action item is resolved.

4.0 CONCLUSION

As described above, the NRC staff has reviewed the Wolf Creek RVI AMP and Inspection Plan and concludes that it is consistent with the I&E guidelines of MRP-227-A and is therefore, acceptable. The NRC staff finds that the licensee has adequately addressed and resolved all the applicable applicant/licensee action items specified in MRP-227-A. The NRC staffs approval of the Wolf Creek RVI AMP and Inspection Plan does not reduce, alter, or otherwise affect current ASME Code,Section XI, ISI requirements, or any Wolf Creek, specific licensing basis requirements related to ISI.

5.0 REFERENCES

1. Hafenstine, C. R., Wolf Creek Nuclear Operating Corporation, letter to U.S. Nuclear Regulatory Commission, Docket No. 50-482: Wolf Creek Generating Station Reactor Vessel Internals Aging Management Plan, dated October 25, 2017 and

Enclosure:

Wolf Creek RVI Inspection Plan, Revision 0 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML17305A042).

2. Electric Power Research Institute, letter to U.S. Nuclear Regulatory Commission, Transmittal: PWR Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), dated January 9, 2012 (ADAMS Package Access No. ML120170453).

3. U.S. Nuclear Regulatory Commission, Safety Evaluation Report Related to the License Renewal of Wolf Creek Generating Station, Docket No. 50-482, Wolf Creek Nuclear Operating Corporation, NUREG-1915, dated October 2008 (ADAMS Accession No. ML083090483).

4. Electric Power Research Institute, letter to U.S. Nuclear Regulatory Commission, Report Transmittal; Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), dated January 12, 2009 (ADAMS Package Accession No. 090160212).

5. Nelson, R. A., U.S. Nuclear Regulatory Commission, letter to Neil Wilmshurst, Electric Power Research Institute, Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability [Program] (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (TAC No. ME0680), dated December 16, 2011 (ADAMS Accession No. ML11308A770).

6. Westinghouse Electric Company, LLC, Reactor Internals Guide Tube Wear -

Westinghouse Domestic Fleet Operational Projections, WCAP-17451-P, Revision 1, dated October 2013 (not publicly available, proprietary information).

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8. Electric Power Research Institute, letter to U.S. Nuclear Regulatory Commission, Transmittal of NEI-03-08 Needed Interim Guidance Regarding Baffle Former Bolt inspections for Tier 1 plants as Defined in Westinghouse NSAL 16-01, dated July 27, 2016 (ADAMS Accession No. ML16211A054).

9. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation Staff Assessment of Electric Power Research Institute NEI 03-08, Revision 2, Needed Interim Guidance Regarding Baffle-Former Bolt Inspections in Westinghouse-Design Pressurized Water Reactors, dated November 2017 (ADAMS Accession No. ML17310A861).

10. Waters, R. M., Entergy e-mail to Kimberly Green, U.S. Nuclear Regulatory Commission, NRC000219, the InfoGram on the Clevis Insert Cap Screws, dated August 10, 2015,

Attachment:

Westinghouse InfoGram IG-10-1, Reactor Internals Lower Radial Support Clevis Insert Cap Screw Degradation, dated March 31, 2010 (ADAMS Accession No. ML15223A367).

11. Pressurized Water Reactor Owners Group, Presentation Slides, Industry and NRC Coordination Meeting Materials Programs Technical Exchange Clevis Insert Bolt Update, dated June 2014 (ADAMS Accession No. ML15335A282).

12. Westinghouse Electric Company LLC, Lower Radial Support System (LRSS) Clevis Inserts and Attachment Bolts Design and Safety Function, WAAP-8828-P, Revision 0, dated March 2014 (not publicly available, proprietary information).

13. Dacimo, F., Entergy Nuclear Northeast, letter to U.S. Nuclear Regulatory Commission, Reply to Request for Additional Information Regarding the License Renewal Application Indian Point Nuclear Generating Unit Nos. 2 & 3, dated September 27, 2013 (ADAMS Accession No. ML13277A007).

14. U.S. Nuclear Regulatory Commission, Generic Aging Lessons Learned (GALL Report),

NUREG-1801, Revision 2, Final Report, dated December 2010 (ADAMS Accession No. ML103490041).

15. U.S. Nuclear Regulatory Commission, Final License Renewal Interim Staff Guidance LR-ISG-2011-04 Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors, dated May 2013 (ADAMS Accession No. ML12270A436).

16. Stuchell, S. D., U.S. Nuclear Regulatory Commission, memorandum to Anthony J.

Mendiola, U.S. Nuclear Regulatory Commission, Summary of May 21, 2013, Public Meeting Regarding Pressurized Water Reactor (PWR) Vessel Internals Inspections, dated June 24, 2013 (ADAMS Accession No. ML13164A126).

17 U.S. Nuclear Regulatory Commission, Presentation Slides, Status of Resolution of MRP-227-A Action Items 1 and 7, dated June 5, 2013 (ADAMS Accession no.

ML13154A152)

18. Pressurized Water Reactor Owners Group, PA-MSC-1288 PWR RV Internals Cold-Work Assessment, PWROG-15105-NP, Revision 0, dated April 2016 (ADAMS Accession No. ML17075A195).

19. Molkenthin, J. for Stringfellow, J., Pressurized Water Reactor Owners Group, letter to U.S. Nuclear Regulatory Commission, PWR Owners Group Submittal of PWROG-15105-NP, Revision 0, PWR RV Internals Cold-Work Assessment to the NRC for Information Only (PA-MSC-1288), dated June 15, 2016 (ADAMS Accession No. ML16222A299).

20. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation Summary Assessment of Pressurized-Water Reactor Owners Group - 15105-NP, Revision 0, PA-MSC-1288 Pressurized-Water Reactor Vessel Internals Cold Work Assessment Staff Assessment, dated April 2017 (ADAMS Accession No. ML17081A010).

21. Morey, D. C., U.S. Nuclear Regulatory Commission, letter to Brian Burgos, Electric Power Research Institute, Final Safety Evaluation of Action Items 1 and 7 from Topical Report MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline, (CAC No. MF7223 EPID: L-2016-TOP-0001),

dated January 29, 2018 (ADAMS Accession No. ML18016A008).

22. Electric Power Research Institute, MRP-227-A Applicability Template Guideline, MRP Letter 2013-025, dated October 14, 2013 (ADAMS Accession No. ML13322A454).

23. Electric Power Research Institute, Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191), dated November 2006 (ADAMS Accession No. ML091910130).

24. Electric Power Research Institute, Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals Component Items (MRP-189-Rev. 1), dated March 2009 (not publicly available, proprietary information).

25. Grimes, C. I., U.S. Nuclear Regulatory Commission, letter to Mr. Douglas J. Walters, Nuclear Energy Institute, License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components, dated May 19, 2000 (ADAMS Accession No. ML003717179).

26. Wolf Creek Nuclear Operating Corporation, Wolf Creek Generating Station Revision 32 to Updated Final Safety Analysis Report, Chapters 4 and 5 (ADAMS Accession Nos.

ML19092A071 and ML19092A073, respectively).

27. U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2011-07, License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management, dated July 21, 2011 (ADAMS Accession No. ML111990086).

Principal Contributor: O. Yee Date: July 9, 2019