ML15037A458

Staff Assessment of the Aging Management Program for Reactor Vessel Internal Components

ML15037A458
Person / Time
Site:	Prairie Island
Issue date:	03/06/2015
From:	David Pelton Plant Licensing Branch III
To:	Davison K Northern States Power Co, Xcel Energy
Beltz T
References
TAC MF0052, TAC MF0053
Download: ML15037A458 (13)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 6, 2015 Mr. Kevin K. Davison Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East Welch, MN 55089

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 - STAFF ASSESSMENT REGARDING PROGRAM PLAN FOR AGING MANAGEMENT FOR REACTOR VESSEL INTERNALS (TAC NOS. MF0052 AND MF0053)

Dear Mr. Davison:

By letter dated October 1, 2012, as supplemented by letters dated March 7, March 22, June 24, and October 8, 2013, and March 28, October 20, 2014. and December 11, 2014, Northern States Power Company, a Minnesota corporation (the licensee), doing business as Xcel Energy, submitted the Prairie Island Nuclear Generating Plant (PINGP) Pressurized Water Reactor Vessel Internals Program for U.S. Nuclear Regulatory Commission (NRC) review. The PINGP aging management plan (AMP) was developed based on the NRC staff approved topical report MRP-227-A, "Material Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The submittal of the AMP was to fulfill a regulatory commitment that originated from license renewal activities as documented in NUREG-1960, "Safety Evaluation Report Related to the License Renewal of Prairie Island Nuclear Generating Plant Units 1 and 2, Docket Nos. 50-282 and 50-306."

The NRC staff has completed its review of the PINGP reactor vessel internals AMP and concludes that it is acceptable because it is consistent with the inspection and evaluation guidelines of MRP-227-A, and the licensee has appropriately addressed all eight action items specified in MRP-227-A.

Regulatory Commitment 25, as documented in Appendix A of NUREG-1960, is considered fulfilled. The NRC staff's approval of the PINGP reactor vessel internals AMP does not reduce, alter, or otherwise affect current American Society of Mechanical Engineers code,Section XI, inservice inspection requirements, or any specific PINGP licensing requirements related to inservice inspection. The staff notes that Section 7.0, "Implementation Requirements," of MRP-227-A requires that the NRC be notified of any deviations from the "Needed" requirements.

The NRC staff's safety assessment of the PINGP AMP for reactor vessel internals components is enclosed.

If you have any questions concerning this matter, please contact the Senior Project Manager, Terry Beltz at (301) 415-3049 or via e-mail at Terry.Beltz@nrc.gov.

i~

Davia L. Pe ton, C ief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosure:

Staff Assessment cc w/encls: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 STAFF ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AGING MANAGEMENT PROGRAM PLAN FOR REACTOR VESSEL INTERNALS NORTHER STATES POWER COMPANY - MINNESOTA PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12276A041), as supplemented by letters dated March 7, March 22, June 24, and October 8, 2013, and March 28, October 20, and December 11, 2014 (ADAMS Accession Nos. ML13067A284, ML13084A378, ML13175A333, ML13284A081, ML14087A219, ML14293A582, and ML14349A658, respectively), Northern States Power Company, a Minnesota corporation (NSPM, the licensee), doing business as Xcel Energy, submitted an aging management program (AMP) for the reactor vessel internals (RVI) at the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. The PINGP AMP was developed based on the NRC staff approved Electric Power Research Institute topical report, "Material Reliability Program [MRP]: Pressurized Water Reactor [PWR] Internals Inspection and Evaluation [l&E]

Guidelines (MRP-227-A)," and its supporting reports were used as technical bases for developing the PINGP AMP (ADAMS Accession Nos. ML12017A193 - ML12017A199).

The licensee submitted its AMP with the intent of meeting license renewal (LR) Commitment 25 as stated in Appendix A of NUREG-1960, "Safety Evaluation Report Related to the License Renewal of Prairie Island Nuclear Generating Plant Units 1 and 2, Docket Nos. 50-282 and 50-306," published in August 2011 (ADAMS Accession No. ML11236A175).

Commitment 25 in NUREG-1960 states, in part, the following:

A. A PWR Vessel Internals Program will be implemented. Program features will be as described in LRA [license renewal application] Section B2.1.32.

B. An inspection plan for reactor internals will be submitted for NRC review and approval at least twenty-four months prior to the period of extended operation. In addition, the submittal will include any necessary revisions to the PINGP PWR Vessel Internals Program, as well as any related changes to the PINGP scoping, screening and aging management review results for Enclosure

reactor internals, to conform to the NRC--approved inspection and Evaluation Guidelines.

The NRC staff reviewed the AMP inspection plan for the RVI components at PINGP to determine whether the licensee provided an inspection plan consistent with the l&E guidelines in MRP-227-A. The staffs assessment is provided below.

2.0 REGULATORY EVALUATION

Part 54 to Title 1O of the Code of Federal Regulations (1 O CFR 54) addresses the requirements for renewal of operating licenses for nuclear power plants. The regulation at 10 CFR 54.21 requires that each application for LR contain an integrated plant assessment (IPA) and an evaluation of time limited aging analyses (TLAAs). The plant-specific IPA shall identify and list those structures and components subject to an aging management review (AMR) 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 the intended functions will be maintained consistent with the current licensing basis (CLB) for the period of extended operation (PEO) as required by 10 CFR 54.29(a). In addition, 10 CFR 54.22 requires that a LR application include any technical specification (TS) changes or additions necessary to manage the effects of aging during the PEO as part of the LR application.

Structures and components subject to an AMP shall encompass those structures and components that (1) perform an intended function, as described in 10 CFR 54.4, without moving parts or without a change in configuration or properties, and (2) are not subject to replacement based on a qualified life or specified time period. These structures and components are referred to as "passive" and "long-lived" structures and components, respectively.

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 (ASME) Code,Section XI) and those RVI components that serve an intended LR 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 in 10 CFR 54.21(a)(1).

The NRC staff's evaluation of the LR application for PINGP, Units 1 and 2, was documented in Appendix A in NUREG-1960, which contains Commitment 25 related to RVI components. The PINGP inspection plan was developed by the licensee based on MRP-227-A with the intent to meet this LR commitment. MRP-227-A summarized the most recent industry recommended l&E guidelines for PWR RVI components. The safety evaluation (SE) in MRP-227-A was issued on December 16, 2011, 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 published MRP-227-A, and the applicant/licensee action items were specified for applicants/licensees to address plant-specific issues which could not be resolved generically in the December 16, 2011, SE on MRP-227-A. In fact, almost all actions related to the above-mentioned LR commitment have been accomplished as a result of issuance of the December 16, 2011, SE in MRP-227-A, or are being addressed by the licensee in its plant-

specific inspection plan based on MRP-227-A in response to the eight applicant/licensee action items.

3.0 TECHNICAL EVALUATION

In its October 1, 2012, submittal, the licensee included inspection requirements of RVI components that are consistent with MRP-227-A. In addition, by letter dated March 7, 2013, the licensee submitted its response to the NRC staff's TR Conditions 1 through 7, which were included in the staff's SE for MRP-227-A. A supplemental letter dated March 22, 2013, included the staff's Action Items 3 through 8, which were addressed in the SE for MRP-227-A. The staff reviewed the AMP for the RVI components addressed in these submittals and determined that a major portion of the licensee's AMP contained no specific technical information that would affect the review and approval of the PINGP inspection plan. Therefore, the focus of the NRC staff's evaluation is related to the following issues: (1) TR Condition 7 in the staff's SE for MRP-227-A requires the licensee address its operating experience at PINGP units; (2) the staff's Action Items 1 through 8 that are addressed in the staff's SE for MRP-227-A; (3) the staff's Conditions as addressed in the SE for MRP-227-A; and (4) the staff's evaluation of the licensee's AMP for the RVI components under the current ASME Code,Section XI, inservice inspection (ISi) program.

The following section provides the NRC staff's evaluation of the four issues described above.

3.1 Evaluation of the Licensee's AMP for Some RVI Components - Operating Experience Appendix A in MRP-227-A addresses the AMP for some RVI components that are susceptible to various aging degradation mechanisms. Appendix A also includes operating experience related to the identification of various aging effects in some RVI components that were manifested during the plant operation. In this context, in a letter dated March 24, 2013, the NRC staff requested in request for additional information (RAl)-1 that the licensee address the AMPs associated with RVI components that were fabricated from (1) nickel based alloys, (2) precipitation hardened stainless steel materials, (3) alloys A-286 and A453, and (4) type 431 stainless steel materials. Additionally, the staff requested the following items: (1) AMP for the clevis insert assembly and (2) AMP for the control rod guide tube (CRGT) cards. The following paragraphs address these issues.

Materials Susceptible to Degradation In its June 24, 2013, submittal, the licensee provided a response to RAl-1. The licensee stated that precipitation hardened stainless steel materials, type 431 stainless steel materials, and Alloy A-286 and A453 were not used at PINGP. Clevis inserts and their bolts and locking bars were made from nickel alloys, and type 347 stainless steel materials in baffle-former bolts were used at PINGP. Nickel alloys are susceptible to primary water stress corrosion cracking (PWSCC). Since clevis inserts are being inspected under the ASME Code,Section XI, program, the NRC staff determined that aging degradation due to PWSCC in clevis insert assemblies is adequately managed by the licensee. Type 347 stainless steel baffle-former bolts which were binned under "Primary" inspection category are being inspected every 1O years using an ultrasonic testing (UT) technique. Based on the licensee's response, the staff concluded that aging degradation of the aforementioned materials at PINGP is being adequately

managing by the licensee during the PEO. Therefore, the staff concludes that the concern expressed in RAl-1 is resolved.

RVI Components in the ASME Code,Section XI. ISi Program In the June 24, 2013, submittal, in response to the NRC's RAl-2, the licensee provided a list of all the RVI components that were inspected thus far, under the ASME Code,Section XI, ISi program. The licensee stated that these materials were inspected under the B-N-3 category of the ASME Code,Section XI, and so far, no aging degradation was identified in the RVI components. Based on the response, the staff determined that its concern with respect to RAl-2 is resolved.

Operating Experience on PWSCC in Alloy X-750 In the June 24, 2013, submittal, in response to the NRC's RAl-3, the licensee provided information regarding the industry's operating experience on PWSCC in Alloy X-750 (Nickel Alloy) used in clevis insert assembly. The licensee acknowledged that Alloy X-750 was used in the clevis insert assembly and that the licensee could not retrieve the necessary document that would confirm the type of high temperature heat treatment (HTH) that was performed on the bolt during fabrication. HTH on Alloy X-750 would make the bolts more resistant to PWSCC, and since the licensee did not confirm HTH was performed on its clevis insert assembly, the licensee screened in the bolts for identifying the aging effects due to PWSCC. So far, no PWSCC was identified in clevis insert assembly at PINGP units. However, the license stated that it would continue to inspect the clevis insert assembly under its ASME Code,Section XI program. Based on this response the staff determined that its concern with respect to RAl-3 is resolved.

AMP for the CRGT Guide Cards In response to RAl-4 related to the AMP for the CRGT guide cards, in a letter dated June 24, 2013, the licensee stated that it intends to comply with the generic guidelines addressed in WCAP-17451, "Reactor Internals Guide Tube Wear-Westinghouse Fleet Operational Projections." The licensee stated that it would comply with the inspection criteria addressed in Table 4-3 in MRP-227-A. The inspection criteria include inspections of the CRGT guide cards for identifying wear using VT-3 technique. In Attachment 1 of the March 22, 2013, submittal, the licensee stated that it would begin the initial examination of the cards no later than two refueling outages from the beginning of the PEO with a subsequent inspection frequency at a 10-year interval. This is consistent with the inspection criteria addressed in MRP-227-A, and therefore, the NRC staff determined that its concern with respect to RAl-4 is resolved.

3.2 Evaluation of the Licensee's AMP for the RVI Components under the ASME Code,Section XI. ISi Program As discussed under Section 3.1 of this SE, the licensee provided a list of all the RVI components that were inspected thus far, under the ASME Code,Section XI inspection program. The licensee stated that these materials were inspected under the B-N-3 category of the ASME Code,Section XI, using VT-3 inspection technique, and so far, no aging degradation was identified in these RVI components. Based on this, the NRC staff concludes

that continuing the ASME Code,Section XI inspections provide adequate assurance that the licensee is adequately managing the aging degradation in the RVI components at the PINGP units.

3.3 Demonstration of Applicant/Licensee Action Item 1 Compliance to SE for MRP-227-A Section 4.2.1 of the SE for MRP-227-A states that "Each applicant/licensee should refer, in particular, to the assumptions regarding plant design and operating history made in the failure modes, effects and criticality analysis (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. 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 (Al) 1."

To resolve the generic issue of the information needed from licensees to resolve Al 1, a series of proprietary and public meetings were conducted. These meetings were attended by the NRC staff, Westinghouse, the Electric Power Research Institute (EPRI), and utility representatives.

The representatives discussed regulatory concerns and to determine a path for a comprehensive and consistent utility response to demonstrate applicability of MRP-227-A, specifically for Westinghouse and CE-design PWR RVI. A summary of the proprietary meeting presentations and supporting proprietary generic design basis information is contained in Westinghouse proprietary report WCAP-17780-P, "Reactor Internals Aging Management MRP-227-A Applicability for Combustion Engineering and Westinghouse Pressurized Water Reactor Designs," and it provides background proprietary design information regarding variances in stress, fluence, and temperature in the plants designed by Westinghouse and CE to support NRC reviews of utility submittals to demonstrate plant-specific applicability of MRP-227-A.

As a result of the technical discussions with the NRC staff, a technical basis was developed for the response to Al 1. For a plant to demonstrate compliance with the MRP-227-A for originally licensed and uprated conditions, it would need to provide a satisfactory response to the following two questions by each Westinghouse and CE unit.

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 [kilopounds per square inch]? (If both conditions are true, additional components may need to be screened in for SCC).

Question 2: Does the plant have 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?

The MRP reviewed these questions and concluded to perform a generic evaluation of the RVI components designed by Westinghouse and CE. By MRP Letter No. 2013-025 dated October 14, 2013, EPRI provided to licensees "MRP-227-A Applicability Guidelines for Combustion Engineering and Westinghouse Pressurized Water Reactor Designs (MRP-227-A Applicability Guidelines)," a non-proprietary document containing guidance for responding to the two questions above. Regarding Question 1, EPRI provided guidance for licensees to assess

whether RVI components at their plant, other than those identified in the generic evaluation, have the potential for cold work greater than 20 percent.

The NRC staff issued RAI 2-1 in a letter dated September 18, 2013 (ADAMS Accession No. ML13253A122), requesting the licensee to provide the following plant-specific information (further information discussed in References 1 and 2) related to verification of the applicability of MRP-227-A to PINGP, Units 1 and 2:

a) Do the PINGP, Units 1 and 2, RVI have non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater and, if so, do the affected components have operating stresses greater than 30 ksi? If so, perform a plant-specific evaluation to determine the aging management requirements for the affected components.

b) Have PINGP, Units 1 and 2, ever utilized atypical fuel design or fuel management that could make the assumptions of MRP-227-A regarding core loading/core design non-representative for that plant, including power changes/uprates? If so, describe how the differences were reconciled with the assumptions of MRP-227-A, or provide a plant-specific aging management program for the affected components, as appropriate.

The NRC staff considers the MRP-227-A Applicability Guidelines acceptable for the following reasons, as described below:

The licensee provided a response to Question a) in a letter dated October 20, 2014. The licensee stated that at PINGP, Units 1 and 2 have no RVI components that were exposed to cold work of 20 percent or greater. The licensee further stated that all the RVI component's welds were screened for monitoring SCC. The licensee evaluated the effect of cold work on SCC in RVI components by using technical bases as addressed in Westinghouse's generic aging evaluation of RVI components-MRP-191 Revision 0, "Materials Reliability Program:

Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs," dated November 2006 (ADAMS Accession No. ML091910130).

Since MRP-191 was used as a basis in establishing l&E guidelines in MRP-227-A, the licensee concluded that no additional stress analyses are required for RVI components. The licensee confirmed that it does not have any RVI components with greater than 20 percent cold work.

Based on these technical bases, the licensee concluded that no further review is required with respect to the impact of cold work on SCC in RVI components at PINGP units.

The NRC staff reviewed the licensee's response and, based on the information provided, concludes that the licensee complied with the guidelines provided in MRP-227-A Applicability Guidelines for RVI components that were binned under Categories 1 through 4 as described in these guidelines. The licensee stated that, for both PINGP units, no cold work 20 percent or more was used in its RVI components. Therefore, the staff concludes that the licensee complied with MRP-227-A Applicability Guidelines with respect to the extent of cold work in its RVI components at PINGP units.

With respect to Question b), MRP-227-A Applicability Guidelines provide quantitative criteria to allow a licensee to assess whether a particular plant has atypical fuel design or fuel management. For Westinghouse design plants, such as both PINGP units, the criteria are as follows:

(1) The heat generation rate must bes 68 Watts/cm 3 (Watts per cubic centimeter);

(2) The maximum average core power density must be less than 124 Watts/cm 3 ;

(3) The active fuel to upper core plate (UCP) distance must be greater than 12.2 inches.

The licensee provided a response to Question 2 in a letter dated October 1, 2014. The licensee stated that it complied with the criteria stated above for the essential attributes, i.e., maximum average core power density, and active fuel distance to UCP. For the heat generation figure of merit assessment, the licensee submitted an acceptance criterion value for figure of merit as addressed in MRP 2013-025, Attachment 1. Since the licensee did not submit a plant-specific heat generation figure of merit value, the NRC staff requested that an actual plant-specific heat generation figure of merit value for each PINGP unit be submitted for review. The staff issued this request in RAl-2-1, part b-1, in an e-mail dated November 24, 2014.

In its letter dated December 11, 2014, the licensee provided plant-specific heat generation figure of merit values for PINGP units. Based on the submitted response, the staff concludes that the licensee complied with the guidelines related to fuel management issue addressed in MRP-227-A Applicability Guidelines. Implementation of l&E guidelines addressed in MRP-227-A would be valid during the PEO, if the licensee would continue to comply with the fuel management criteria as addressed in MRP-227-A Applicability Guidelines.

3.4 Evaluation of the Licensee's Resolution of Action Item 2 in the SE for MRP-227-A In the March 22, 2013, submittal, the licensee stated that it performed the scoping and screening of the RVI components per the requirements of the LR process. The RVI materials used at the PINGP units are consistent with the materials specified in MRP-191 which was used as a technical basis document for the development of l&E guidelines that are addressed in MRP-227-A for CE and Westinghouse units. Based on this evaluation, the licensee concluded that no revisions are required to the AMP for the RVI components at PINGP units.

The staff reviewed the licensee's evaluation and concludes that: (1) the licensee's AMP for the RVI components is consistent with MRP-227-A l&E guidelines; (2) no additional RVI components at the PINGP units were screened in due to the usage of different type of materials that were not prescribed in MRP-191/MRP-227-A; and (3) the licensee complied with the guidelines addressed in Action Item 1 (the staff's evaluation of Action Item 1 was discussed in Section 3.3 of this SE).

Based on this assessment, the staff considers that the licensee adequately addressed Action Item 2 and that this action item is resolved.

3.5 Evaluation of the Licensee's Resolution of Action Items 3 through 8 in the SE for MRP-227-A In its letter dated March 7, 2013, the licensee submitted resolution of Action Items 3 through 8 as addressed in the staff's SE for MRP-227-A. The following paragraphs address the licensee's evaluation of the Action Items and the staff's review.

Action Item 3 Action Item 3 in the staff's SE for MRP-227-A stated that the licensee is required to perform a plant-specific evaluation of its existing program on CRGT support pins at the PINGP Units 1 and

2. The licensee stated that its original CRGT support pins were Alloy X-750 which is a nickel base alloy and it is susceptible to PWSCC. The licensee replaced these pins with cold worked Type 316 stainless steel with a larger shank diameter. According to Table 3-3 in MRP-227-A, CRGT pins were screened in for managing the aging degradation due to wear and fatigue. The licensee stated that Type 316 material has a fatigue life that could be compared to Alloy X-750 material. In its June 24, 2013, submittal, in response to RAl-2, the licensee stated that CRGT split pins are being inspected under the ASME Code,Section XI, ISi program. The licensee would continue to inspect these pins under its ASME Code,Section XI, ISi program which would identify any aging degradation due to wear and fatigue in a timely manner. Therefore, the staff concludes that the AMP for CRGT split pins would be adequately managed by the licensee under its ASME Code,Section XI, ISi program during the PEO.

Based on the above, the NRC staff concludes that the licensee is in compliance with Action Item 3 and that this action item is resolved.

Action Items 4 and 6 Action Items 4 and 6 of the staff's SE for MRP-227-A are applicable to the RVI components designed by B&W, and therefore, they are not applicable to PINGP units.

Action Item 5 Action Item 5 requires that the Westinghouse units identify the plant-specific acceptance criteria to be applied when performing physical measurements for loss of compressibility due to loss of load for the hold-down springs. Loss of load is applicable to more susceptible material (i.e.,

Type 304 stainless steel) than Type 403 stainless steel which has higher yield strength. Since Type 403 springs is used at PINGP units, no physical measurements of the hold-down spring are provided. Since Type 403 stainless steel hold-down spring material has superior resistance to loss of load than Type 304 stainless steel material, the staff determined that physical measurements for loss of compressibility for Type 403 stainless steel hold-down springs in PINGP units are not necessary.

Based on the above, the NRC staff concludes that the licensee is in compliance with Action Item 5 and, therefore, this action item is resolved.

Action Item 7 Action Item 7 requires that a plant-specific analysis demonstrating that lower support columns made of cast austenitic stainless steel (CASS) materials would maintain their functionality during the LR period. The relevant aging degradation mechanisms for the CASS lower support columns are thermal embrittlement and irradiation embrittlement. The staff noted that the lower support columns at the PINGP units are made of wrought (non-cast) stainless steel and, therefore, they are not susceptible to thermal embrittlement. With respect to the irradiation embrittlement in the wrought stainless steel lower support columns, the staff noted that in

Table 4-6 of the MRP-227-A report, the wrought stainless steel lower support columns were binned under "Expansion" inspection category with upper core barrel flange weld as a Primary link. Cracking is the relevant aging effect that occurs due to irradiation embrittlement and the extent of cracking in any RVI component due to irradiation embrittlement would depend on the intensity of the applied stress. The upper core barrel flange weld would be exposed to higher stresses than the lower support column bodies. Hence, cracking would be identified in the upper core barrel flange weld (Primary link) much sooner than in the lower support columns.

Since the lower support columns are part of the licensee's inspection program that is consistent with the MRP-227-A guidelines, the staff concludes that effective AMP for the wrought lower support columns is adequately implemented at PINGP units.

Based on the above, the NRG staff concludes that the licensee is in compliance with Action Item 7 and that this action item is resolved.

Action Item 8 Action Item 8 of the NRG staff's SE for MRP-227-A requires that the licensee submit the AMP for the RVI components that is consistent with l&E guidelines addressed in MRP-227-A. The licensee's AMP program elements were reviewed by the staff and the evaluation was addressed in its SE - NUREG-1960, "Safety Evaluation Report Related to License Renewal of the Prairie Island Nuclear Generation Plants, Units 1 and 2." The staff's acceptance of these AMP program elements were addressed in Section 3.0.3.3.2 in NUREG-1960.

Based on the above, the NRG staff concludes that the licensee is in compliance with Action Item 8 and that this action item is resolved.

3.6 TR Conditions in the Staff's SE for the MRP-227-A With respect to the seven conditions in the staff's SE for the MRP-227-A, the NRG staff reviewed the licensee submittals dated October 1, 2012, and May 7, 2013, and its evaluation is provided below.

Condition 1: The licensee, in its inspection program addressed in Enclosure 1 of the October 1, 2012, submittal, has added the upper core plate and lower support forging or casting to its RVI inspection program. This addition is consistent with the guidelines addressed in Table 4-6 of the MRP-227-A report, and, therefore, the staff considers that this issue closed.

Condition 2: Consistent with the l&E guidelines addressed in Table 4-3 of the MRP-227-A report, the licensee included the upper and lower core barrel welds and lower core barrel flange in its inspection program that was addressed in Enclosure 1 of the October 1, 2012, submittal.

Therefore, the licensee satisfied this condition.

Condition 3: This condition is not applicable to Westinghouse designed RVI components and, therefore, the staff is not addressing this issue in this SE.

Condition 4: A criterion for a minimum area of inspection coverage is addressed in this condition. This criterion states that a minimum of 75 percent coverage of the entire examination volume (i.e., including both accessible and inaccessible regions) of the RVI components and

their welds, and a minimum sample size of 75 percent of the total population of like components i.e.; bolts should be inspected. The licensee included this guideline in its Enclosure 1 of the October 1, 2012, submittal, and, therefore, the staff considers that this issue closed.

Condition 5: This condition states that a 10-year inspection frequency for baffle-former bolts in Westinghouse-designed reactors should be implemented following the initial or baseline inspection is required. The licensee satisfied this condition by including this criterion in its of the October 1, 2012, submittal, and, therefore, the staff considers that this issue closed.

Condition 6: Subsequent re-examination for all "Expansion" inspection category components should be at a 10-year interval once degradation is identified in the associated "Primary" inspection category component. The licensee included this criterion in its Enclosure 1 of the October 1, 2012, submittal, and, therefore, the NRC staff considers that this issue closed.

Condition 7: The licensee in its submittal dated May 7, 2013, stated that this condition is satisfied because the Appendix A in MRP-227-A report was updated to include the operating experience related to the aging degradation of the RVI components in the PWR fleet. The NRC staff accepts this response and considers that this issue is closed.

Based on the above, the NRC staff concludes that the licensee, in its AMP for RVI components, included all the conditions that are addressed in the staff's SE for MRP-227-A. Therefore, the staff accepts the licensee's inspection program for the RVI components.

4.0 CONCLUSION

The NRC staff has reviewed the inspection plan for PINGP, Units 1 and 2. The staff concludes that the PINGP inspection plan is acceptable because it is consistent with l&E guidelines of MRP-227-A, and the licensee appropriately addressed all eight applicant/licensee action items specified in MRP-227-A. Consequently, the staff determines that licensee has fulfilled Regulatory Commitment 25 as documented in Appendix A of NUREG-1960 for the RVI components at PINGP, Units 1 and 2.

5.0 REFERENCES

1. Meeting Summary Electric Power Research Institute (EPRI) - Westinghouse January 22-23, 2013, dated February 21, 2013 (ADAMS Accession No. ML13042A048).

2. February 25, 2013 Summary of Teleconference with EPRI and Westinghouse Electric Company, dated March 15, 2013 (ADAMS Accession No. ML13067A262).

Principal Contributor: Ganesh Cheruvenki Date: March 6, 2015

If you have any questions concerning this matter, please contact the Senior Project Manager, Terry Beltz at (301) 415-3049 or via e-mail at Terry.Beltz@nrc.gov.

Sincerely, IRA/

David L. Pelton, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosure:

Staff Assessment cc w/encls: Distribution via ListServ DISTRIBUTION:

PUBLIC RidsNrrDeEvib Resource LPL3-1 r/f RidsNrrDlrRarb Resource RidsAcrsAcnw_MailCTR Resource RidsRgn3MailCenter Resource RidsNrrDorllpl3-1 Resource KMorganButler, EDO Region Ill RidsNrrPMPrairielsland Resource GCheruvenki, NRR/EVIB RidsNrrLAMHenderson Resource ADAMS Access1on No. ML15037A458 *via

. ema1*1 d ae t d J anuary 30 2015 OFFICE LPL3-1/PM LPL3-1/LA DE/EVIB DLR/RARB/BC LPL3-1/BC NAME TBeltz MHenderson SRosenburg

• DMorey DPelton DATE 02/10/2015 02/11/2015 01/30/2015 03/02/2015 03/06/2015 OFFICIAL RECORD COPY