ML16281A510

Staff Assessment of the Reactor Vessel Internals Aging Management Program Plans

ML16281A510
Person / Time
Site:	Robinson
Issue date:	12/15/2016
From:	Dion J Plant Licensing Branch II
To:	Glover R Duke Energy Progress
Galvin D, NRR-DORL 415-6256
References
CAC ME9633
Download: ML16281A510 (23)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 15, 2016 Mr. Richard Michael Glover Site Vice President H. B. Robinson Steam Electric Plant Duke Energy Progress, LLC.

3581 West Entrance Road, RNPA01 Hartsville, SC 29550

SUBJECT:

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 - STAFF ASSESSMENT OF THE REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM PLANS (CAC NO. ME9633)

Dear Mr. Glover:

By letter dated September 26, 2012, as supplemented by letters dated May 23, July 25, 2013; January 9, February 19; September 5, October 1, October 15, 2014; February 18, and October 5, 2016; Duke Energy Progress, LLC (the licensee) (previously Duke Energy Progress, Inc. or Progress Energy Carolinas), submitted reactor vessel internals (RVI) aging management program plan (AMP) for the H. B. Robinson Steam Electric Plant Unit No. 2 (RNP). The RVI AMP was submitted to fulfill License Renewal Commitment No. 33 for RNP, as documented in Appendix A of NUREG-1785, "Safety Evaluation Report Related to the License Renewal of H. B. Robinson Steam Electric Plant, Unit 2." The RVI AMP is based on "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A)." License Renewal Commitment No. 33 for RNP was fulfilled upon submittal of the RVI AMPs on September 26, 2012.

The U.S. Nuclear Regulatory Commission (NRC) staff's review of the licensee's RVI AMP is provided in the enclosed staff assessment. The NRC staff concludes that the licensee's RVI AMP is acceptable because it is consistent with the inspection and evaluation guidelines of MRP-227-A, and the licensee has adequately addressed all eight specified licensee action items.

The NRC staff's approval of the RNP RVI AMP does not reduce, alter, or otherwise affect current American Society of Mechanical Engineers Code,Section XI, lnservice Inspection (ISi) requirements, or any RNP specific licensing requirements related to ISi. The licensee must follow the implementation requirements as defined in Section 7.0 of MRP-227-A, which require that the NRC be notified of any deviations from the "Needed" requirements.

R. Glover If you have any questions concerning this matter, please contact Project Manager, Dennis Galvin at (301) 415-6256 or Dennis.Galvin@nrc.gov.

Sincerely,

~~~ief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-261

Enclosure:

Staff Assessment cc w/enclosures: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 STAFF ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE AGING MANAGEMENT PROGRAM PLANS FOR REACTOR VESSEL INTERNALS DUKE ENERGY PROGRESS, LLC H. B. ROBINSON STEAM ELECTRIC PLANT UNIT NO. 2 DOCKET NO. 50-261

1.0 INTRODUCTION AND BACKGROUND

By letter dated September 26, 2012 (Reference 1), as supplemented by letters dated May 23, July 25, 2013; January 9, February 19, September 5, October 1, October 15, 2014; February 18, and October 5, 2016 (References 2 to 10, respectively); Duke Energy Progress, LLC (the licensee) (previously Duke Energy Progress, Inc. or Progress Energy Carolinas),

submitted reactor vessel internals (RVI) aging management program plan (AMP)

(Reference 11) for the H. B. Robinson Steam Electric Plant Unit No. 2 (RNP or Robinson). The RVI AMP is based on the U. S. Nuclear Regulatory Commission (NRC)-approved Electric Power Research Institute (EPRI) topical report (TR), MRP-227-A, "Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" (Reference 12).

The licensee submitted the RVI AMP to fulfill License Renewal Commitment No. 33 for RNP, as documented in Appendix A of NUREG-1785, "Safety Evaluation Report Related to the License Renewal of H. B. Robinson Steam Electric Plant, Unit 2" (Reference 13). The AMP included Inspection and Evaluation (l&E) guidelines for the RVI components at RNP.

2.0 REGULATORY EVALUATION

Title 1O of the Code of Federal Regulations (1 O CFR) Part 54, "Requirements for renewal of operating licenses for nuclear power plants," addresses the requirements for plant license renewal process. The regulation at 10 CFR Section 54.21, "Contents of application-technical information, requires that each application for license renewal (LR) 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 for the period of extended operation (PEO) as required by 10 CFR 54.29(a). In addition, 10 CFR 54.22, "Contents of application-technical specifications, requires that a license renewal application include any technical specification changes or additions necessary to manage the effects of aging during the PEO as part of the LR application.

Enclosure

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, "Scope,"

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 ).

On January 12, 2009, the EPRI submitted, for NRC staff review and approval, MRP-227, Revision 0 (Reference 14), which was intended as guidance for applicants in developing their plant-specific AMPs for RVI components. 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 also provides l&E guidelines for PWR applicants to use in their plant-specific AMPs. The final NRC safety evaluation (SE) regarding MRP-227 was issued on December 16, 2011 (Reference 15), with seven 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 applicant/licensees to address plant-specific issues that could not be resolved generically in the December 16, 2011, SE. On January 9, 2012, EPRI submitted the NRC-approved version of the TR designated MRP-227-A (Reference 16).

3.0 TECHNICAL EVALUATION

The licensee's AMP is organized as follows:

1. Purpose

2. Background

3. Program Owner

4. Description of RNP RVI Aging Management Program and Industry Program

5. RNP's RVI Aging Management Program Attributes

6. Demonstration of Applicant/Licensee Action Items and SE Conditions Compliance to SE on MRP-227-A

7. Program Enhancement and Implementation Schedule

8. Implementing Documents

9. References Appendix A Illustrations Appendix B RNP's License Renewal Aging Management Review Summary Tables Appendix C MRP-227-A Augmented Inspections Sections 1-4, 7-9 and Appendices A and C contain no specific technical information that would affect the review and approval of the RN P's inspection plan. Therefore, the focus of the NRC

staff's evaluation is related to Section 5, which includes operating experience, Appendix 8, and Section 6.

3.1 Robinson Reactor Vessel Internals Aging Management Program Attributes: Section 5.0 Licensee Evaluation In its September 26, 2012, submittal (Reference 1), the licensee stated that the AMP for the RVI components at RNP is in compliance with the 10 attributes addressed in NUREG-1801, "Generic Aging Lessons Learned (GALL) report - Final Report," Revision 2 (Reference 17),

Section Xl.M16A, "PWR Vessel Internals." The licensee further stated that: (1) RNP fully utilized the GALL process contained in NUREG-1801 in performing an aging management review of the RVI components in the license renewal process, (2) the AMP for the RVI components at RNP includes consideration of the augmented inspections identified in MRP-227-A and fully meets the requirements of Section Xl.M16A in the GALL Report, and (3) the augmented inspections, based on program enhancements resulting from industry programs, will become part of the ASME Code,Section XI, program.

NRC Staff Evaluation

The NRC staff reviewed the 10 program elements for RN P's AMP and compared them with the program elements Xl.M16A in GALL report, Revision 2, and concluded the following: (1) all 10 program elements described in Section 5.0 of RN P's September 26, 2012, submittal are consistent with the program elements addressed in Section Xl.M16A, in GALL report, Revision 2, (2) all the program elements included extensive discussions on the implementation of MRP-227-A, l&E guidelines and ASME Code,Section XI, ISi criteria, (3) the licensee's discussions included applicability of its Action Items addressed in the NRC staff's SE for MRP-227-A, and (4) the licensee complied with the provisions addressed in MRP-227-A, specifically with respect to the following program elements: (a) parameters monitored or inspected, (b) detection of aging effects, and (c) monitoring and trending. Therefore, the staff finds the licensee's implementation of the 1O AMP elements acceptable for RNP.

3.2 Operating Experience: Section 5.10 Section 5.1 O of the September 26, 2012, submittal addressed operating experience related to the aging degradation reported to date in RVI components at RNP. In this context, the licensee identified aging degradation mechanisms for two RVI components: (1) control rod guide tube (CRGT) split pins, which are susceptible to primary water stress corrosion cracking (PWSCC);

and, (2) control rod guide cards that exhibited wear during normal operation.

3.2.1 Control Rod Guide Tube Split Pins Licensee Evaluation CRGT split pins at RNP were fabricated from Alloy X-750 which did not receive high temperature heat (HTH) treatment. Alloy X-750 material without undergoing HTH treatment is more susceptible to PWSCC. The licensee, in Section 6.0 of its September 26, 2012, submittal, stated that as part of its corrective action it replaced Alloy X-750 with type 316 stainless steel material, which has superior resistance to PWSCC.

NRC Staff Evaluation

In a letter dated July 25, 2013 (Reference 3), in response to NRC Requests for Additional Information (RAI) 2-1 and 2-2 (Reference 18), the licensee stated that CRGT split pins were categorized under the ASME Code,Section XI, Examination Category B-N-3. The licensee stated it will continue to inspect the CRGT spilt pins under its ASME Code,Section XI, lnservice Inspection (ISi) program.

The NRC staff accepts this response because: (1) the licensee replaced the CRGT split pins with a type 316 stainless steel material that is more resistant to PWSCC and, (2) the licensee will continue to perform ASME Code,Section XI, inspections of these pins during the PEO. The staff determined that routine inspections of CRGT split pins per the ASME Code,Section XI, provide reasonable assurance that the aging degradation in this component is adequately monitored by the licensee during the PEO. Therefore, the staff considers that the license has addressed operating experience for the CRGT split pins and the issue addressed in RAI 2-1 and RAI 2-2 is closed.

3.2.2 Control Rod Guide Cards Licensee Evaluation The licensee stated that control rod guide cards were binned under "Primary" inspection category in MRP-227-A and that they would be inspected for aging degradation due to wear every 10 years. In Appendix C, Table C-1, of the licensee's submittal dated September 26, 2012 (Reference 1), the licensee indicated that the control rod guide cards are to be inspected no later than two refueling outages from the beginning of the PEO. This submittal also stated that during the spring 2010 refueling outage, the licensee performed inspections on control rod guide cards to assess the wear in these cards.

NRC Staff Evaluation

Based on its review of the licensee's submittal, by letter dated March 27, 2013 (Reference 19), in RAI 4, the NRC staff requested that the licensee provide the following information related to the control rod guide cards: (1) the number of cards inspected, (2) the inspection results, (3) how the criteria for maximum allowed wear was established, (4) the licensee's corrective actions, if any, and (5) the licensee's plan for subsequent inspections of this component during the PEO.

In its response to RAI 4, in a letter dated May 23, 2013 (Reference 2), the licensee stated that all nine guide cards within nine of the guide tubes (20 percent 1 compliant with MRP-227-A requirements) were inspected and no aggressive degradation due to wear was observed at RNP to date. The licensee stated that it used the acceptance criteria developed by Westinghouse for monitoring aging effects in control rod guide cards. The licensee did not observe aggressive guide card inner wear or backside wear in the inspected guide tube locations nor did it observe any general trend in the ligament wear pattern of the inspected guide tubes. Since the magnitude of the ligament wear was small, the licensee concluded that it did not need to take any corrective action.

1 20 percent refers to 20 percent examination of the number of CRGT assemblies. with all guide cards within each selected CRGT assembly examined.

The Westinghouse-developed criteria for RNP included maximum allowed guide tube-guide card wear, and the acceptance criteria and related tests (as stated by the licensee) are:

Guide card drag test - The purpose of this test was to determine if the contact between a worn Reactor Cluster Control Assembly (RCCA) rodlet and a worn guide card could result in hanging up or catching of the rodlet during a controlled rod insertion or a gravity free-fall, Determination of collapse strength of a typical rodlet assembly - The purpose of this calculation note was to determine, by finite element (FE) simulation, the collapse strength of a typical rodlet assembly (rodlet host tube and coaxial internal absorber) due to axial (longitudinal) compressive forces that could occur during insertion or withdrawal of the ROCCA, Stress evaluation of the worn guide card holes - The purpose of the evaluation was to determine the limitation of the wear on guide card holes based on stresses remaining acceptable in the guide tube assembly, Evaluation of RCCA rodlet wedging or sticking during or after a faulted event - The purpose of this evaluation was to determine a ligament thickness wear acceptance criterion for a 15x15 style guide tube that ensures that a rodlet will not wedge or stick during or after a faulted event, In order to apply the acceptance criteria, it was determined that at the selected guide tubes, the holes with the smallest ligaments and expected aggressive wear should be visually examined and measured at every guide card span.

The licensee further stated that there is a concern for rodlet breakout; therefore, the slot width on the guide card is measured, and depending on the amount of wear, the guide card is compared against different acceptance criteria. The acceptance criteria establishes the zone in which the guide tube would fall based on the measured wear at the most worn guide card hole in that guide tube. These zones are defined as Green, Yellow, and Red Zones and have associated recommendations to follow when a guide tube falls within a certain zone.

In Section 5.6 of the September 26, 2012 submittal, the licensee referenced Westinghouse report WCAP-17096-NP, Revision 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," dated December 2009 (Reference 20). WCAP-17096-NP, Revision 2, provides inspection guidelines and acceptance criteria for monitoring wear in guide cards. In Section 5.10 of the September 26, 2012, submittal, the licensee indicated that it is participating in a generic program developed by Westinghouse addressing inspection and acceptance criteria forthe guide cards. This generic program resulted in part in WCAP-17451-P, Revision 1, "Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections," dated October 2013 (Reference 21). WCAP-17451-P, Revision 1 provides Westinghouse-design PWR domestic plants with guidance on managing guide card wear. The NRC staff reviewed WCAP-17096-NP, Revision 2 and WCAP-17451-P, Revision 1, and issued an SE on May 3, 2016 (Reference 22). Regarding WCAP-17096-NP, Revision 2, the NRC staff found the evaluation methodology and acceptance criteria for the guide cards acceptable because it provides a methodology for measuring wear that is based on ensuring functionality of the RCCAs, and the acceptance criteria provide margin for future wear. Regarding

WCAP-17451-P, the NRC staff found the report provides a rigorous and comprehensive basis for the methods and criteria for guide card wear evaluation. The SE has been incorporated into the approved version WCAP-17096-NP-A Revision 2, dated August 2016 (Reference 23).

Therefore, with respect to RNP, the NRC staff considers an AMP for guide cards based on WCAP-17096-NP-A, Revision 2, an acceptable alternative to MRP-227-A. The existing Robinson guide card inspection criteria continue to be valid and are conservative in comparison to the wear criteria in WCAP-17 451-P, Revision 1.

Based on the licensee's response, the NRC staff concludes that the AMP for guide cards is adequately managed at RNP. The staff's conclusion is based on the following: (1) no aggressive degradation to date was observed at RNP, (2) guide cards are inspected every 10-year interval in accordance with l&E guidelines addressed in MRP-227-A, (3) the licensee would use the fleet-wide inspection results in establishing the subsequent inspection frequency and, (4) if any unacceptable wear in guide cards were to be observed during the future inspections, the licensee, as part of its corrective action, would determine subsequent inspection frequency for the guide cards based on the extent of wear observed in these cards.

Based on the evaluation stated above, the staff concludes that the licensee adequately demonstrated that its AMP is effective in identifying wear in guide cards in a timely manner.

Therefore, the staff considers that the issue addressed in RAI 4 is closed.

3.2.3 Clevis Insert Bolts The NRC staff noted that page A-2 of Appendix A to MRP-227-A stated that failures of Alloy X-750 clevis insert bolts were reported by one licensee. Alloy X-750 clevis insert bolts that did not receive HTH during the original fabrication are susceptible to PWSCC.

Licensee Evaluation In its response to the NRC staff's RAI 3 (Reference 19), in a letter dated May 23, 2013 (Reference 2), the licensee stated that the clevis insert bolts did not receive HTH treatment, and therefore, they are susceptible to PWSCC. The clevis inserts are being inspected for wear using visual testing (VT-3), and this component was binned under "Existing" inspection category in MRP-227-A.

In RAI 3-3, sent by e-mail dated September 23, 2013 (Reference 24), the NRC staff expressed concern that VT-3 may not be adequate to identify cracking in clevis insert bolts, and, therefore, it requested the licensee to provide justification for using VT-3 to detect PWSCC. In its initial response to the staff's RAI 3-3, in a letter dated January 9, 2014 (Reference 4), the licensee stated that industry efforts to investigate recent operating experience on cracking of the clevis insert bolts was ongoing and that a subsequent detailed response would address the concern.

In its subsequent response to the staff's RAI 3-3, in a letter dated September 5, 2014 (Reference 6), the licensee stated that during the removal of core barrel, motion of the lower end of the core barrel causes wear on the clevis insert. The failures of the bolts do not result in the loss of intended function of the clevis insert. The intended function can be monitored by VT-3 inspections of the radial key/clevis insert interfacing surfaces where any wear would be detected.

Any detection of wear between these surfaces is very important to assess the performance of the clevis insert. Westinghouse addressed these issues in Technical Bulletin (TB) 14-5, "Reactor

Internals Lower Radial Support Clevis Cap Screw Degradation," 2 (Reference 25) in which emphasis was made on the absence of any impact on the safety function of the clevis insert due to failed bolts. TB 14-5 also stated that even though failure of bolts does not result in loss of safety function of the clevis insert, continued operation with degraded bolts would pose difficulty to removing the core barrel.

Westinghouse's evaluation of the clevis insert bolts was based on the following factors:

(1) degradation of the bolts does not affect the safety function of the clevis insert because the clevis insert is essentially held in place during normal operation by the interfacing lower internal components; (2) the clevis insert could only be dislodged from the radial key when the internals are removed and the unit was operating with bolts in a degraded condition for a long period of time; (3) the loss of function of the clevis inserts is attributed to an increased motion between the lower end of the core barrel caused by wear on the clevis insert; (4) the existing ASME Code,Section XI, VT-3 inspections would monitor the wear between the core barrel and the clevis insert adequately, thereby, ensuring safety function of the clevis insert assembly; and, (5) Westinghouse, in its TB 14-5, proposed that the inspection methods should also focus on monitoring the aging degradation of the bolts.

NRC Staff Evaluation

Based on the information provided, the NRC staff concludes that the AMP of the clevis inserts is adequately managed at RNP because: (1) no relevant indications were found to date in the clevis insert bolts; (2) no wear was observed between the clevis insert keyways and the clevis insert; (3) no minor scratches and gouges were observed in the clevis insert keyways and the insert; (4) continued inspections of the clevis insert bolts in accordance with ASME Code,Section XI, would ensure the functionality of the clevis inserts; and (5) since the inspected area of coverage was 100 percent, the staff believes that the functionality of the clevis insert was not compromised due to disassembly of the core barrel. Therefore, the staff considers that the issue addressed in RAI 3 and RAI 3-3 is closed. Based on the information provided, the staff considers that the licensee's AMP for the clevis inserts provides reasonable assurance that the clevis insert's safety function is maintained during the PEO.

3.2.4 Materials Susceptible to Degradation Operating experience in the PWR fleet to date identified that nickel base and stainless steel alloys are susceptible to some of the aging degradation mechanisms addressed in MRP-227-A.

In this context, by a letter dated March 27, 2013 (Reference 1), in RAI 1 NRC staff requested that the licensee confirm that the following materials are not currently used in the RVI components at RNP: nickel base alloys (i.e., inconel 600), weld metals (i.e., Alloy 82 and 182),

Alloy X-750 (excluding control rod guide tube split pins), Alloy A-286, ASTM [American Society for Testing and Materials] A 453 (Grade 660, Condition A or B), type stainless steel 347 material (excluding baffle-former bolts), precipitation hardened (PH) stainless steel materials (i.e., 17-4 and 15-5) and type 431 stainless steel material.

2 Westinghouse TB 14-5 was provided to the NRC as enclosure 3 to the licensee's RAI response dated September 5, 2014 (Reference 6).

Licensee's Evaluation In its response to RAI 1, dated May 23, 2013 (Reference 2), the licensee stated that the following materials are not used at RNP: Alloy A-286, ASTM A 453 (Grade 660, Condition A or B), PH stainless steel materials (17-4 and 15-5), and type 431 stainless steel material.

However, the licensee identified the following materials which were used in RVI components at RNP: (1) type 347 stainless steel material is used in baffle-former bolts and this material is susceptible to irradiation assisted stress corrosion cracking, (2) lnconel 600 is used in clevis inserts and flux thimble assembly and this material is prone to PWSCC, (3) Alloy X-750 is used in clevis insert bolts and this material is prone to PWSCC, and (4) Alloy 82 weld is used for tack welds in clevis insert lock keys and this material is prone to PWSCC.

NRC Staff Evaluation

The NRC staff accepted the response because: (1) baffle-former bolts are binned under "Primary" inspection category and the aging degradation is monitored using l&E guidelines of MRP-227-A, (2) clevis inserts are routinely inspected in accordance with the requirements of ASME Code,Section XI, (3) flux thimble tubes are monitored under the licensee's thimble tube inspection program in accordance with NRC Bulletin 88-09, and (4) clevis inserts are inspected in accordance with the requirements of ASME Code,Section XI, and, therefore, their aging degradation is routinely monitored. Therefore, the staff considers that the issue addressed in RAI 1 is closed. Based on the review, the staff concludes that aging degradation in RVI components described above are adequately monitored using routine inspections as required by the ASME Code,Section XI criteria, and by the criteria addressed in l&E guidelines of MRP-227-A.

3.3 Appendix B, "RNP License Renewal Aging Management Review Summary Tables," and ASME Code,Section XI, RVI Components Consistent with AMP Xl.M16A in the GALL report, the licensee included a partial list of RVI components that are part of its ASME Code,Section XI, ISi program at RNP. The NRC staff reviewed Appendix B of the licensee's submittal dated September 26, 2012 (Reference 1), and in RAI 2-1 (Reference 18), requested the licensee to submit a complete list of all ASME Code,Section XI, B-N-3, core support structure RVI components at RNP. In its response dated July 25, 2013 (Reference 3), the licensee provided a complete list of all ASME Code,Section XI RVI components at RNP. The licensee stated that these components were inspected in accordance with requirements of the ASME Code,Section XI, criteria. Based on the licensee's response, the staff considers that the issue addressed in RAI 2-1 is closed because RVI components categorized under ASME Code,Section XI, are being inspected in accordance with ASME Code,Section XI, criteria.

3.4 Applicant/Licensee Action Items of Safety Evaluation for MRP-227-A 3.4.1 Evaluation of the Licensee's Resolution of Action Item 1 Section 4.2.1 of the SE for MRP-227-A states that "Each applicant/licensee shall 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, Combustion Engineering (CE), or Babcock and Wilcox (B&W)),

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."

To resolve the generic issue of the information needed from licensees to resolve Action Item 1, a series of proprietary and public meetings were conducted, at which the NRC, Westinghouse, EPRI, and utility representatives, discussed regulatory concerns and determined 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 bases information are 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" (Reference 26). This report provides background on the proprietary design information regarding variances in stress, fluence, and temperature in the RVI components that were 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, the basis for a plant to respond to the NRC staff RAI questions regarding Action Item 1, to demonstrate compliance with MRP-227-A for originally licensed and uprated conditions, was determined to be satisfied with plant-specific responses to the following two questions:

Question 1: Does the plant 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 both conditions are true, additional components may need to be screened in for stress corrosion cracking, 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?

In an enclosure to MRP Letter 2013-025, "MRP-227-A Applicability Template Guideline, dated October 14, 2013 (Reference 27), EPRI provided to licensees "MRP-227-A Applicability Guidelines for Combustion Engineering and Westinghouse Pressurized Water Reactor Designs" (MRP-227-A Applicability Guidelines), which provides guidance for responding to the two questions above. An NRC staff evaluation (Reference 28), assessed MRP Letter 2013-025 and the technical basis contained in WCAP-17780-P, and concluded that if an applicant or licensee demonstrates that its plant(s) comply with the guidance in MRP Letter 2013-025, there is reasonable assurance that the l&E guidance of MRP-227-A will be applicable to the specific plant(s). The staff evaluation also concluded that the guidance in MRP Letter 2013-025 provides an acceptable basis for licensees to respond to the generic Questions 1 and 2 addressed above.

By correspondence dated September 23, 2013 (Reference 24), the NRC staff requested in RAI 3-1 that the licensee provide information similar to that identified in Questions 1 and 2 above related to verification of the applicability of MRP-227-A to RNP.

With respect to Question 1, the applicability guidelines addressed in MRP Letter 2013-025 provides guidance for the 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. In its initial response to the staff's RAI 3-1, in a letter dated January 9, 2014 (Reference 4), the licensee stated that it would provide subsequent detailed response that aligned with MRP Letter 2013-025. In its subsequent response to Question 1, by letter dated September 5, 2014 (Reference 6), the licensee stated that it considered plant modifications and operating history of RNP and binned the RVI components under the following categories:

Category 1: Cast austenitic stainless steel (CASS);

Category 2: Hot-formed stainless steel; Category 3: Annealed austenitic stainless steels; Category 4: Austenitic stainless steels fasteners; and, Category 5: Cold formed austenitic stainless steels without subsequent solution annealing.

Materials binned under Categories 1, 2 and 3 contain no greater than 20 percent of cold work due to controlled fabrication and compliance with material specifications. Therefore, the RVI components binned under Categories 1, 2 and 3 are consistent with MRP Letter 2013-025 guidelines. Only materials that fall under Categories 4 and 5 were treated as cold worked and they were evaluated as such. For components binned under Category 4, cold work greater than 20 percent was already considered in the Westinghouse's generic aging evaluation of the RVI components. This generic evaluation is addressed in MRP-191, Revision 0, "Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs" (Reference 29). Since MRP-191 was used as a basis for establishing l&E guidelines in MRP-227-A, the licensee concluded that no additional stress analyses are required for Category 4 RVI components. The licensee confirmed that it does not have any Category 5 RVI components that were subject to cold work greater than 20 percent at RNP. Based on these technical bases, the licensee concluded that no additional plant-specific evaluation related to the effect of cold work on stress corrosion cracking (SCC) in RVI components binned under Categories 1 through 5 is required.

The NRC staff reviewed the licensee's response and based on the information provided, the staff concluded that the licensee complied with the guidelines provided in MRP Letter 2013-025 for RVI components that were binned under Categories 1 through 5. The staff also noted that the RVI components binned under Categories 1 through 5 were not exposed to cold work greater than 20 percent. Therefore, the staff concludes that the licensee has demonstrated that it adequately evaluated the effects cold work on sec in RVI components binned under Categories 1 through 5 at RNP. Therefore, the staff concludes that the licensee complied with the guidelines (with respect to the evaluation of cold work on SCC issue) addressed in MRP Letter 2013-025.

With respect to Question 2, the MRP Letter 2013-025 provides quantitative criteria to allow a licensee to assess whether a particular plant has atypical fuel design or fuel management. For a Westinghouse design plant such as RNP, these criteria are:

(1) The heat generation rate must be :5 [less than or equal to] 68 Watts/cm 3 .

(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.

In its response to Question 2 of RAI 3-1, by letters dated October 1, 2014 and October 5, 2016 (References 7 and 10)3 , the licensee stated that to date, it complied with the criteria stated above for the essential attributes (i.e., heat generation rate, maximum average core power density, and active fuel distance to UCP). The licensee provided plant-specific values related to heat generation rate, maximum average core power density, and active fuel distance to UCP.

The NRC staff reviewed the submitted values and concludes that the licensee complied with the guidelines related to fuel management issue addressed in MRP Letter 2013-025. Based on this review, the staff considers that the licensee addressed Action Item 1 satisfactorily.

3.4.2 Evaluation of the Licensee's Resolution of Action Item 2 Section 4.2.2 of the SE for MRP-227-A states that "each applicant/licensee is responsible for identifying which RVI components are within the scope of LR for its facility. Applicants/licensees shall review the information in 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 such that the effects of aging on the missing component(s) will be managed for the PEO."

The licensee, in its September 26, 2012, submittal (Reference 1), stated that it performed the scoping and screening of the RVI components as per the requirements of the LR process. The licensee also stated that RVI materials used at RNP are consistent with the materials specified in MRP-191, which was used as a technical basis document for the development of l&E guidelines in MRP-227-A. Furthermore, the licensee stated that it did not propose any modifications to the AMP addressed in MRP-227-A. Based on this evaluation, the licensee concluded that no revisions are required to the AMP for the RVI components at RNP. The NRC 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 RNP 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. Details of the staff's evaluation of the Action Item 1 are addressed in Section 3.4.1 of this Staff Assessment. Based on this assessment, the staff considers that the licensee addressed Action Item 2 satisfactorily.

3.4.3 Evaluation of the Licensee's Resolution of Action Item 3 Section 4.2.3 of the SE for MRP-227-A states that "applicants/licensees of Westinghouse are required to perform plant-specific analysis either to justify the acceptability of an 3

The enclosures to References 7 and 10 provide the response to RAI 3-1. The enclosures are the same with the enclosure to Reference 10 being the redacted version of the enclosure to Reference 7.

applicant's/licensee's 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 analyses and a description of the plant-specific programs being relied on to manage aging of these components shall be submitted as part of the applicant's/licensee's AMP application. The Westinghouse components identified for this type of plant-specific evaluation include: Westinghouse guide tube support pins (split pins)."

Action Item 3 in the NRC staff's SE for the MRP-227-A report stated that the licensee is required to perform a plant-specific evaluation of its existing program on CRGT support split pins at RNP. In Section 3.2.1 of this Staff Assessment, the staff determined that routine inspections of CRGT split pins per the ASME Code,Section XI, provide reasonable assurance that the aging degradation in this component is adequately monitored by the licensee during the PEO. Based on this review, the staff considers that the licensee addressed Action Item 3 satisfactorily.

3.4.4 Evaluation of the Licensee's Resolution of Action Items 4 and 6 Action Items 4 and 6 of the NRC staff's SE for the MRP-227-A report are applicable to the RVI components designed by B&W, and, therefore, they are not applicable to RNP.

3.4.5 Evaluation of the Licensee's Resolution of Action Item 5 Section 4.2.5 of the SE for MRP-227-A states that "applicants/licensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRG-approved version of MRP-227 for loss of compressibility for Westinghouse hold down springs. 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 as part of their submittal to apply the approved version of MRP-227."

Action Item 5 in the NRC staff's SE for MRP-227-A states that the licensee should identify a plant-specific acceptance criterion to be applied while performing the physical measurement of loss of compressibility for hold-down springs. The hold-down spring at RNP was fabricated with 304 austenitic stainless steel material that is susceptible to loss of preload due to irradiated assisted creep/stress relaxation.

The NRC staff noted that some of the owners of Westinghouse-designed reactors have replaced 304 stainless steel hold-down springs with 403 stainless steel material that is more resistant to creep/stress relaxation than 304 stainless steel material. In this context, by letter dated May 29, 2013 (Reference 18), in RAI 2-3, the staff requested that the licensee address its plans to replace the hold-down spring with 403 stainless steel material.

Licensee's Evaluation By letter dated July 25, 2013 (Reference 3), in response to the NRC staff's RAI 2-3, the licensee stated that it would replace the existing 304 stainless steel hold-down spring with a 403 stainless steel spring. The licensee further stated that the rate of stress relaxation in 403 stainless steel material is lower than 304 stainless steel material even when exposed to

a neutron fluence that exceeds the threshold limit for irradiated assisted creep/stress relaxation. Therefore, the licensee concluded that the replacement of the hold-down spring with 403 stainless steel material provides reasonable assurance that the AMP for this item is adequately managed during the PEO. By letter dated February 18, 2016 (Reference 9), in response to the staff's RAI 6 (Reference 30), the licensee stated that it replaced the 304 stainless steel hold-down spring with 403 stainless steel material in spring 2015 refueling outage 29. The guidelines in MRP-227-A states that the replacement should occur within the three cycles of the beginning of the PEO, and the licensee complied with this criterion. Furthermore, the licensee stated that the replacement schedule did not adversely impact the integrity of the core support structure because it occurred in accordance with the schedule requirements of MRP-227-A.

NRC Staff Evaluation

The NRC staff reviewed the response and concludes that replacement of the 304 hold-down spring with 403 stainless steel alleviates the loss of preload due to irradiated assisted creep/stress relaxation. This is due to the fact that stiffness value of 403 stainless steel material is higher than that of 304 stainless steel material. Since the replacement of the hold-down spring occurred within the three cycles of the beginning of the PEO, the staff concludes the licensee complied with the guidelines of MRP-227-A.

Based on the licensee's response, the staff considers that the licensee addressed Action Item 5 satisfactorily.

3.4.6 Evaluation of the Licensee's Resolution of Action Item 7 Action Item 7 was discussed in Section 3.3.7 of the NRC staff's SE for the MRP-227-A, and it specifies that the licensees of Westinghouse reactors develop plant-specific analyses to be applied for their facilities to demonstrate that CASS lower support columns (LSCs) will maintain their function during the PEO. These components are subject to irradiation embrittlement (IE) and thermal embrittlement (TE). CASS materials with delta ferrite greater than 20 percent would be susceptible to loss of fracture toughness due to TE. CASS materials that are exposed to neutron fluence value greater than 1x10 17 n/cm 2 (E>1 MeV) are susceptible to IE.

Occurrence of TE depends on the ferrite content, which in turn depends on the chemical composition of CASS material, Molybdenum content and, casting process. By letter dated January 8, 2014, in RAI 5(b)-1 (Reference 31), the NRC staff requested that the licensee provide the ferrite content and the casting method for LSC using certified material test reports.

By letter dated February 19, 2014 (Reference 5), the licensee stated that the maximum ferrite content in all 68 LSCs is 18.4 percent. Since the ferrite content is less than 20 percent, the staff determined that the LSCs at RNP are not susceptible to TE. However, the staff noted that the IE is still an active aging degradation in the LSCs. Functionality of the LSCs would be affected if the structural integrity of these columns is compromised due to IE. To assist licensees in addressing the functionality of the LSCs, the industry developed and submitted to staff for information a generic functionality report, Pressurized Water Reactor Owners Group (PWROG)-14048-P, Revision 0, "Functionality Analysis: Lower Support Columns" (Reference 32). In the staff assessment of PWROG-14048-P (Reference 33), the staff identified site-specific analyses necessary to demonstrate the applicability of PWROG-14048-P to a

particular site. Based on this assessment, by letter dated February 8, 2016 (Reference 30), the staff, issued the RAI S(b)-1-2, which states:

The NRC staff has determined that the flaw tolerance analysis contained in report PWROG-14048-P utilized conservative assumptions to demonstrate that the likelihood of failure of the LSCs is low during the period of extended operation (PEO). It is reasonable to infer that the functionality of the LSCs will be maintained during the PEO if the likelihood of failure of the LSCs is shown to be low. Therefore, the staff requests the licensee to demonstrate how the flaw tolerance analysis in PWROG-14048-P is applicable to the Robinson LSCs using plant-specific parameters (such as LSC geometry and number of LSCs) and conditions (such as loading conditions and LSC stresses). If the licensee determines that PWROG-14048-P is not applicable to the Robinson LSCs or chooses not to apply it, the staff requests that the licensee identify its approach to demonstrating that the functionality of the LSCs will be maintained during the PEO.

By a letter dated February 18, 2016 (Reference 9), the licensee stated that the industry will use the PWROG-14048-P report and provide justification of plant-specific applicability for the participating plants. The licensee stated that industry will revise PWROG-14048-P report to consider the staff's conclusions of its assessment of the report. Furthermore, the licensee stated that industry will consider RNP's specific parameters, operating conditions and, its applicability justification in the revised version of the PWROG-14048-P report. The licensee made a commitment to submit a revised version of the PWROG-14048-P report by June 30, 2017.

The NRC staff reviewed the licensee's response to RAI 5(b )-1-2 and the associated commitment related to LSC's functionality. Based on its review, the staff determined that the licensee's commitment is consistent with the guidelines addressed in Action Item 7 of the staff SE for MRP-227-A. The staff considers that the licensee's response is acceptable because of the following reasons:

( 1) It is established that the licensee is an active participant in the PWROG project for demonstrating LSC's functionality for the PEO, taking into consideration potential aging degradation of LSCs due to IE, and, (2) The June 30, 2017, completion date for this commitment includes a submittal of the revised version of the PWROG-14048-P report. This revised report will include RNP's specific parameters, operating conditions and, its applicability justification for using this report for the evaluation of the functionality of RNP's LSCs.

The NRC staff finds that reasonable control for the implementation of the above regulatory commitment is best provided through the licensee's administrative processes, including its commitment management program. The above regulatory commitment does not warrant the creation of a regulatory requirement, and is not relied upon for the approval of the RNP RVI AMP.

3.4. 7 Evaluation of the Licensee's Resolution of Action Item 8 Action Item 8 was discussed in Section 3.5.1 of the staff's SE for MRP-227-A and specifies that the licensee submit an AMP for the RVI components that is consistent with l&E guidelines addressed in MRP-227-A. In its submittal dated September 26, 2012 (Reference 1), the licensee included its AMP and in Section 5 addressed the 10 AMP program elements of the MRP-227-A guidelines. As discussed in Section 3.1 of this Staff Assessment, the staff finds the licensee's implementation of the 10 AMP elements in conjunction with l&E guidelines in MRP-227-A and thus is acceptable for RNP. Therefore, the staff considers that the licensee addressed Action Item 8 satisfactorily.

3.5 Topical Report Conditions in the Staff's Safety Evaluation for MRP-227-A Section 4.1 of the NRC staff SE for MRP-227 contains seven conditions that the licensee must follow to receive credit for MRP-227-A implementation. The NRC staff reviewed the licensee's submittal against these seven conditions.

Condition 1: The licensee, in its inspection program addressed in Section 6.1 of its September 26, 2012, submittal (Reference 1), 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 MRP-227-A; therefore, the NRC staff finds Condition 1 is met.

Condition 2: Consistent with the l&E guidelines addressed in Table 4-3 of MRP-227-A, the licensee included the upper and lower core barrel welds and lower core barrel flange in its AMP. Therefore, the NRC staff finds Condition 2 is met.

Condition 3: This condition is not applicable to Westinghouse designed RVI components and, therefore, the NRC staff the staff finds Condition 3 is met.

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 (e.g., bolts) should be inspected. The licensee included this guideline in its AMP; therefore, the staff finds Condition 4 is met.

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. The licensee satisfied this condition by including this criterion in Section 6.1 of the September 26, 2012, submittal; therefore, the staff finds Condition 5 is met.

Condition 6: This condition states that 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 guideline in the AMP; therefore, the NRC staff finds Condition 6 is met.

Condition 7: In Section 5.0 of the September 26, 2012, submittal, the licensee stated that the operating experience related to the aging degradation of the RVI components in the PWR fleet would be periodically documented.

Furthermore, the licensee included the operating experience related to the aging degradation of some of the RVI components at RNP. The NRC staff's review of the operating experience at RNP is addressed in Section 3.2 of this Staff Assessment. Based on the review, the staff found that the licensee provided the necessary information required by MRP-227-A. The staff finds Condition 7 is met.

Based on the review of the licensee's responses to the seven conditions, the NRC staff concludes that the licensee had adequately addressed all the conditions stated in the staff's SE for MRP-227-A.

4.0 CONCLUSION

The NRC staff has reviewed the AMP for the RNP's RVI components and concludes that the RNP AMP is acceptable because it is consistent with l&E guidelines of MRP-227-A. The licensee addressed all eight applicant/licensee action items and seven conditions specified in MRP-227-A appropriately.

The NRC staff's approval of the RNP RVI AMP does not reduce, alter, or otherwise affect current ASME Code,Section XI, ISi requirements, or any RNP specific licensing requirements related to ISi. The licensee must follow the implementation requirements as defined in Section 7.0 of MRP-227-A, which require that the NRC be notified of any deviations from the "Needed" requirements.

5.0 REFERENCES

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ME4200)," May 3, 2016 (ADAMS Accession Nos. ML16061A243 (Public Version) and ML16061A194 (Proprietary Version: not publicly available, Package4 )).

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24. Lingam, S., U. S. Nuclear Regulatory Commission, E-mail to Connelly, S., Duke Energy Progress, Inc., "RE: Robinson, Unit 2 PWR Vessel Internal Program Plan for Aging Management - Revised Requests for Additional Information (RAls) (TAC No. ME9633),"

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ML13183A372 (Transmittal letter dated June 28, 2013) and ML13183A373 (Proprietary Report: not publicly available)).

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Screening, Categorization and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Designs" November 2006 (ADAMS Accession No. ML091910129 (Transmittal letter dated July 2, 2009) and ADAMS Accession No. ML091910130 (Report))

30. Barillas, M., U. S. Nuclear Regulatory Commission, letter to Glover, R. M., Duke Energy Progress, Inc., "H. B. Robinson Steam Electric Plant, Unit 2 - Request for Additional 4

While WCAP-17096-NP, Revision 2, is publicly available, the SE used information from the non-publicly available report, WCAP-17 451-P, Revision 1, "Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projects," October 2013, which EPRI referred to in RAI responses associated with the review of WCAP-17096-NP, Revision 2.

Information Related to the Pressurized Water Reactors Internals Program Plan for Aging Management of Reactor Internals (TAC ME9633)," February 8, 2016 (ADAMS Accession No. ML16019A053)

31. Lingam, S., U.S. Nuclear Regulatory Commission, E-mail to Hightower, R., Duke Energy Progress, Inc., "RE: Robinson, Unit 2 PWR Vessel Internal Program Plan for Aging Management - Request for Additional Information (RAls) (TAC No. ME9633),"

January 8, 2014 (ADAMS Accession No. ML14016A279).

32. Pressurized Water Reactor Owners Group (PWROG)-14048-P, Revision 0, "Functionality Analysis: Lower Support Columns," December 2014 (ADAMS Accession No. ML15077A113 (Transmittal letter dated March 13, 2015) and ADAMS Accession No. ML15077A114 (Proprietary Report: not publicly available))

33. McHale, J. J., U. S. Nuclear Regulatory Commission, memorandum to Hsueh, K., U. S.

Nuclear Regulatory Commission, "Summary Assessment of Report PWROG-14048-P

'Functionality Analysis: Lower Support Columns,"' December 17, 2015 (ADAMS Accession No. ML15334A462)

Principal Contributor: G. Cheruvenki, NRR Date of issuance: December 15, 2016

ML16281A510 *by memo (ML16074A232 (non-public)

OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA NRR/DE/EVIB*

NAME DGalvin BClayton JMcHale DATE 12/8/2016 12/8/2016 6/23/2016 OFFICE NRR/DLR/RARB* NRR/DORL/LPL2-2/BC(A) DORL/LPL2-2/PM NAME DMorey JDion DGalvin DATE 5/10/2016 12/9/2016 12/15/2016