ML14014A205

Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Relief Request Regarding Risk-Informed Inservice Inspection Program for the Fifth 10-Year Inservice Inspection Interval (TAC Nos. MF1150 and MF1151)

ML14014A205
Person / Time
Site:	Point Beach
Issue date:	01/30/2014
From:	Carlson R D Plant Licensing Branch III
To:	McCartney E Point Beach
Beltz T A
References
TAC MF1150, TAC MF1151
Download: ML14014A205 (16)

Text

• • uNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Eric McCartney Site Vice President NextEra Energy Point Beach, LLC Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, WI 54241-9516 January 30, 2014 SUBJECT: POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 -ISSUANCE OF RELIEF REQUEST REGARDING RISK-INFORMED INSERVICE INSPECTION PROGRAM FOR THE FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL (TAC NOS. MF1150 AND MF1151) Dear Mr. McCartney: By letter dated March 19, 2013, as supplemented by letters dated August 9, 2013, and September 5, 2013, NextEra Energy Point Beach, LLC (NextEra or the licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC) for authorization to implement use of a risk-informed I safety-based (RIS_B). inservice inspection (lSI) program for piping at the Point Beach Nuclear Plant (PBNP), Units 1 and 2. NextEra proposed the use of the RIS_B process for the lSI of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Class 1 and Class 2 piping, Examination Categories B-F, B-J, C-F-1, and C-F-2 piping welds. Specifically, pursuant to Title 10 of the Code olFederal Regulations (1 0 CFR) 50.55a(3)(i), the licensee requested to use the proposed alternative on the basis that it provides an acceptable level of quality and safety. The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that NextEra has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a, and is in compliance with the ASME Code's requirements. Therefore, the NRC staff authorizes the proposed RIS_B program in accordance with 10 CFR 50.55a (a)(3)(i) for the fifth 1 0-year inservice inspection interval at PBNP, Units 1 and 2. The NRC staff's approval of the licensee's RIS_B program does not constitute approval. of Code Case N-716.

\ E. McCartney -2 -If you have any questions, please contact Ter;ry Beltz, Senior Project Manager, at (301) 415-3049. Docket Nos. 50-266 and 50-301

• Enclosure: Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST FOR APPROVAL OF A RISK-INFORMED I SAFETY-BASED INSERVICE INSPECTION PROGRAM NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, 1 AND 2 DOCKET NOS. 50-266 AND 50-301 (TAC NOS. MF1150AND MF1151) 1.0 INTRODUCTION By letter dated March 19, 20t3 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13079A092), as supplemented by letters August 9, 2013, and September 5, 2013 (ADAMS Accession Nos. ML 13221A526 and ML 13249A233, respectively), NextEra Energy Point Beach,.LLC (NextEra, licensee) submitted a request to the U.S. Nuclear Regulatory Commission (NRC)"for authorization to implement a risk-informed inservice inspection (RI-ISI) p*rogram plan for the Point Beach Nuclear Plant (PBNP), Units 1 and 2, for the fifth 1 0-year lSI interval. NextEra proposed the use of the risk-informed I safety-based inservice inspection (RIS_B) for the lSI of American Society of Mechanical Engineers Boiler and Pressure Vessel Cs)de (ASME Code), Class 1 and Class 2 piping, Examination Categories B-F, B-J, C-F-1, aod C-F-2 piping welds. Specifically, pursuant to Title-".1 0 of the Code of Federal Regulations (1 0 CFR) 50.55a(3)(i), the licensee requested to use the proposed alternative on the basis that the alternative provides an acceptable level of quality and *safety. NextEra requests to implement a RIS_B program based, in part, on ASME Code Case N-716, "Alternative Piping Classification and Examination Requirements,Section XI Division 1" (Reference 1 ), hereafter referred to as N-716. The ASME Code developed the provisions of N-716 to be used in lieu of the requirements of IWB-2420, IWB-2430, Table IWB-2500-1 (Examination Categories 8-F and B-J), IW.C-2420, IWC-2430, and Table IWC-2500-1 (Examination Categories C-F-1 and C-F-2) for 1$1 of Class 1 or 2 piping, and IWB-2200 and IWC-2200 for pre-service inspection of Class 1 or 2.piping, or as additional requirements for Class 3 piping or Non-Class piping for plants issued an initial prior to December 31, 2000. *The N-716 requirements are expeGted to reduce the number of inspections required, but may also define additional requirements for Class 3 or non:-Ciass piping. Enclosure

-2-N-716 has not been endorsed for generic use by the NRC; however the NRC staff's review of the licensee's application of the code case, as described in this safety evaluation, indicates that it fully complies with the regulatory requirements and safety goals set forth in NRC Regulatory Guides (RGs) for risk-informed inservice inspection programs. The licensee's relief request refers to the methodology described in N-716 instead of describing the details of the methodology. The licensee, however, modified the methodology described in N-716 while developing its proposed RIS_8 program. When the methodology used by the licensee is accurately described in N-716, this safety evaluation (SE) refers to the details found in N-716. When the methodology used by the licensee deviates or expands upon the methodology

• described in f\J-716, this SE refers to the licensee submittals cited above. 2.0 REGULATORY EVALUATION Pursuant to 10 CFR 50.55a(g), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, "except design and access provisions and preservice examination requirements" set forth in the Code to the extent practical within the limitations of design, geometry, and materials of construction of the components. Section 50.55a(g) also states that lSI of the ASME Code, Class 1 ,, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable addenda, except where specific relief has been granted by the NRC. The regulations also require, during the first 1 0-year lSI interval and during subsequent intervals, that a licensee's lSI program comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference into 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the conditions listed therein. The P8NP is currently in its fifth 1 0-year lSI interval which began in August 2012. The ASME Section XI code of record for the fifth lSI interval is the 2007 Edition through the 2008 Addenda. Pursuant to 10 CFR 50.55a(g), a certain percentage of ASME Code Category 8-F, 8-J, C-F-1 and C-F-'2 pressure retaining piping welds must receive lSI during each 1 0-year lSI interval. The ASME Code requires 100 percent of all 8-F welds and 25 percent of all 8-J welds greater than 1-inch nominal pipe size be selected for volumetric or surface examination, or both, on the basis of existing stress analyses. For Categories C-F-1 and C-F-2 piping welds, 7.5 percent of non-exempt welds are selected for volumetric or surface examination, or both. In accordance with 10 CFR 50.55a(a)(3), the NRC may authorize alternatives to the requirements of 10 CFR 50.55a(g) if an applicant demonstrates that the proposed alternatives would provide an acceptable level of quality and safety, or that compliance with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The objective of the lSI program is to identify conditions (i.e., flaw indications) that are precursors to leaks and ruptures in the pressure boundary .of these components that may impact plant safety. It states in 10 CFR 50.55a(a)(3) that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety; or (ii) compliance with .the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

-3 -The licensee has proposed to use a RIS_B program for ASME Code Class 1 and Class 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2 piping welds) as an alternative to the ASME Code,Section XI requirements, on the basis that it provides an acceptable level of quality and safety. The NRC staff finds that there is regulatory basis for the licensee to request and the NRC to authorize this alternative, pursuant to the technical evaluation that follows. The information provided by the licensee in support of the request has been evaluated by the NRC staff and the bases for disposition are documented below. 3:0 TECHNICAL EVALUATION 3.1 Licensee's Request for Alternative The ASME Code components affected by the licensee's proposed alternative are as follows: System: Class: Components Affected: . Various Code Class 1 and 2 Systems Quality Groups A and B (ASME Code Class 1, and 2) . All Class 1 and 2 Piping Welds-Examination Categories B-F, B-J, C-F-1, and C-F-2 The licensee's proposed alternative is as follows: Pursuant to 10 CFR 50.55a(a)(3)(i), Next Era requests an alternative to the requirements of the ASME B&PVC,Section XI, 2007 Edition through the 2008 Addenda, Division 1, Tables IWB-2500-1 and IWC-2500-1, for Examination Categories B-F, B-J, C-F-1 and C-F-2. The alternative would implement a RIS_B inservice inspection program for piping at PBNP. The proposed program is based, in part, on the ASME,Section XI, Code Case N-716. The alternative is proposed for the fifth 1 0-year lSI interval. 3.2 NRC Staff Evaluation Background N-716 is founded, in large part, on the RI-ISI process as described in Electric Power Research Institute TR-112657 Revision B-A, "Revised Risk-Informed lnservice Inspection Evaluation Procedure" (Reference 2, EPRI TR) (ADAMS Accession Number ML0134701 02), which was previously reviewed and approved by the NRC. The NRC staff reviewed the development of the proposed RIS_B RI-ISI program using the following documents: *

• NUREG-0800, Chapter 3.9.8, "Standard Review Plan (SRP) for the Review of Informed lnservice Inspection of Piping," September 2003 (ADAMS Accession No. ML03251 0135) '

• RG 1.17 4, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant':"Specific Changes to the Licensing Basis" (ADAMS Accession No. ML023240437)

-4-* RG 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmaking -lnservice Inspection of Piping" (ADAMS Accession No. ML032510128)

• RG 1.200, Revision 1, "An Approach for the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (ADAMS Accession No. ML070240001) RG 1.17 4 provides guidance on the. use of probabilistic risk analysis (PRA) findings and risk insights in support of licensee requests for changes to a plant's licensing basis. RG 1.178 describes a RI-ISI program as one that incorporates risk insights that can focus inspections on more important locations while at the same time maintaining or improving public health and safety. RG 1.200 describes one acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making. As described above N-716 is founded, in large part, on the RI-ISI process as described in the EPRI TR, which was previously reviewed and approved by the NRC. In general, the licensee simplified the EPRI TR method by using generically identified system parts as high safety-significant (HSS), and used the plant-specific PRA to evaluate in detail only system parts that could not be screened out as low safety-significant (LSS). An acceptable RI-ISI program replaces the number and locations of nondestructive examination (NDE) inspections required by ASME Code,Section XI requirements with the number and locations of these inspections based on the RI-ISI guidelines as described in RG 1.178. The proposed RIS_8 program permits alternatives, in accordance with N-716, to the requirements of IW8-2420, IW8-2430, and IW8-2500 (Examination Categories 8-F and 8-J) and IWC-2420, IWC-2430, and IWC-2500 (Examination Categories C-F-1 and C-F-2) for lSI of Class 1 and 2 piping and IW8-2200 and IWC-2200 for preservice inspection of Class 1 or 2 piping, or as additional requirements for Subsection IWD, and may require lSI and preservice inspection of Class 3, or Non-Class piping. All piping components, regardless of risk classification, will continue to receive ASME Code-required pressure and leak testing, as part of the current ASME Code,Section XI program. *

• The EPRI TR RI-ISI process includes the following steps which, when successfully applied, satisfy the guidance provided in RGs 1.17 4 and 1.178:

• Scope Definition

• Consequence Evaluation

• Degradation Mechanism Evaluation

• Piping Segment Definition

• Risk Categorization o lnspection/NDE Selection

• Risk Impact Assessment

• Implementation Monitoring and Feedback

-5-These processes result in a program consistent with the concept that, by focusing inspections on the most safety-significant welds, the number of inspections can be reduced while at the same time maintaining protection of public health and safety. In general, the methodology in N-716 replaces a detailed evaluation of the safety significance of each pipe segment with a generic population of high safety-significant segments, followed by a screening flooding analysis to identify any plant-specific high safety-significant segments. The screening flooding analysis is performed in accordance with the flooding PRA approach that is consistent with ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications" (Reference 3) as endorsed in RG 1.200. As described below, the acceptability of the licensee's proposed RIS_B program is evaluated by comparing the processes it has applied to develop its program with the steps from the EPRI-TR. 3.2.1 Scope Definition The scope of evaluation to support RIS_B program development and of the proposed changes includes ASME Code Class 1, 2, 3, and Non-Class piping welds. SRP 3.9.8 and RG 1.178 address scope issues. The primary acceptance guideline in the SRP is that the selected scope needs to support the demonstration that any proposed increase in core damage frequency (CDF) and risk are small. The scope of the licensee's evaluation included all Class 1 and 2 piping where ASME inspections could be discontinued, providing assurance that the change-in-risk estimate would, as a minimum, capture the risk increase associated with implementing the RIS_B program in lieu of the ASME program. Also, as noted in section 3.4.2 of the March 19, 2013, submittal, one Class 3 service water pipe in the cable spreading room was identified for inspection in accordance with N-716 for inclusion in the RIS_B program. RG 1.178 identifies different groupings of plant piping that should be included in a RIS_B program, and also clarifies that a "full-scope" risk-informed evaluation is acceptable. The scope of the RIS_B program, as defined in N-716 and as evaluated by NextEra, is consistent with the definition of full-scope in RG 1.178. The NRC staff concludes that the "full-scope" extent of the piping included in the RIS_B program changes, satisfies the guidelines described in SRP 3.9.8 and RG 1.178 and, therefore, is acceptable. 3.2.2 Consequence Evaluation The methodology described in RG 1.178 and EPRI TR-112657 divides all piping within the scope of the proposed EPRI RI-ISI program into piping segments. The consequence of each segment failure must be estimated as a conditional core damage probability (CCDP) and conditional large early release probability (CLERP) or by using a set of tables in the EPRI TR-112657 that yield equivalent results. The consequences are used to determine the safety-significance of the segments. In contrast to the EPRI TR-112657 methodology, N-716 does not require that the consequence of each segment failure be estimated to determine the safety-significance of piping segments. Instead, N-716 identifies portions of systems that should be generically classified as HSS at all plants. A consequence analysis is not required for system parts generically classified as HSS because there is no higher safety significance category to which the system part can be assigned and degrpdation mechanisms, not consequence, are used to select inspection locations in the HSS weld population. The licensee's PRA is subsequently used to search for

-6-any additional, plant-specific HSS segments that are not included in the generic HSS population. Sections 2(a)(1) through 2(a)(4) in N-716 provide guidance that identifies the portions of systems that should be generically classified as HSS based on a review of almost 50 RI-ISI programs. These previous RI-ISI programs were all developed by considering both direct and indirect effects of piping pressure boundary failures and the different failure modes of piping. This is consistent with the guidelines for evaluating pipe failures with PRA described in RG 1.178, the EPRI TR-112657, and SRP 3.9.8. Therefore, the generic results are derived from acceptable analyses. Section 2(a)(5) in N-716 provides guidance that defines additional, plant-specific HSS segments that should be identified using a plant-specific PRA of pressure boundary failures. The licensee stated that it used its PRA of pressure boundary failures (flooding analysis) to identify additional plant-specific HSS segments and that the flooding analysis considered both the direct and indirect effect of pressure boundary failures and the different failure modes of piping. This is consistent with the guidelines for evaluating pipe failures with PRA described in RG 1.178, EPRI TR-112657, and SRP 3.9.8. Each of the licensee's consequence evaluations (the generic and the plant-specific flooding analysis) considers both direct and indirect effects of piping pressure boundary failures and the different piping failure modes to systematically use risk insights and PRA results to characterize the consequences of piping failure. This is consistent with the guidelines for evaluating pipe failures with PRA described in RG 1.178 and EPRI TR-112657 and, therefore, is acceptable. 3.2.3 Degradation Mechanism Evaluation The EPRI TR and Code case differ in the number of pipe segments which are evaluated. The EPRI TR requires the evaluation of each pipe segment to determine all applicable degradation mechanisms. This is then used to determine the safety significance ofthe segment. Alternatively, the code case identifies a generic population of piping segments to be assigned to the HSS category without evaluation, followed by a search for plant-specific HSS welds. In lieu ofconducting a degradation mechanism evaluation for all the LSS piping, all locations were conservatively assigned to the medium-failure potential for the purpose of assigning a failure frequency to be used to calculate the change in risk. This results in an equal or greater estimated increase in risk from discontinued inspections because the failure frequencies would always be equal to or less than those used in the licensee's analysis, if the susceptibility of all LSS welds to all degradation mechanism was determined. The NRC finds this approach of N-716 and the licensee's alternative acceptable, because the assumed degradation mechanism will always result in the assignment of a failure probability at least as high as the complete analysis required by the EPRI TR methodology. The relief request and the EPRI TR differ on the number of pipe segments evaluated for Flow Accelerated Corrosion (FAC) and water hammer. The EPRI TR states that all pipe segments are to be evaluated for FAC and water hammer as the presence of these degradation mechanisms may affect the failure potential for the piping segment. N-716 requires evaluation of all piping segments not specified as HSS by the co,de case to determine whether water ' .

-7 -hammer is present. If water hammer is present in a piping segment then, that segment is assigned a high failure potential in accordance with Table 3 of N-716 because, as stated above, LSS segments are assumed to have a medium failure potential initially. The NRC stafffinds the licensee's use of the N-716 approach acceptable, as it is consistent with the EPRI TR for those segments considered and it is at least as conservative as the EPRI TR for those segments not fully evaluated as these segments were assumed to be of high safety significance. 3.2.4 Piping Segment Definition Previous guidance on RI-ISI, including RG 1.178 and EPRI TR-112657, are centered on defining and using piping. segments. RG 1.178 states, for example, that the analysis and definition of a piping segment must be consistent and technically sound. The primary purpose of segments is to group welds so that co11sequence analyses can be done for the smaller number of segments instead of for each weld. Sections 2(a)(1) to 2(a)(4) in N-716 identify system parts (segments and groups of segments) that are generically assigned HSS without requiring a plant-specific consequence determination and any subdivision of these system parts is unnecessary. Section 2(a)(5) in N-716 uses a PRAto identify plant-specific piping that might be assigned HSS. A flooding PRA consistent with ASME PRA Standard searches for specific HSS piping by first identifying zones that may be sensitive to flooding, and theri evaluating the failure potential of piping in these zones. Lengths of piping whose failure impacts the same plant equipment within each zone are equivalent to piping segments. Therefore, piping segments are either not needed to reduce the number of consequence analyses required (for the generic HSS piping) or, when needed during the plant-specific analysis, the length of pipe included in the analysis is consistent with the definition of a segment in RG 1.178 and SRP 3.9.8. An additional purpose of piping segments in EPRI TR-112657 is as an accounting/tracking tool. In the EPRI methodology, all parts of all systems within the selected scope of the RI-ISI program are placed in segments and the safety significance of each segment is developed. For each safety significant classification, a fixed percentage of welds within all the segments of that class are selected. Additional selection guidelines ensure that this fixed percentage of inspections is distributed throughout the segments to ensure that all damage mechanisms are targeted and all piping systems continue to be inspected: N-716 generically defines a large population of welds as HSS. An additional population of welds may be added based on the informed search for plant-specific HSS segments. When complete, the N-716 process yields a well-defined population of HSS welds accomplishing the same objective as accounting for each weld throughout the analysis by using segments. N-716 provides additional guidelines to ensure that this fixed percentage is appropriately distributed throughout the population of welds subject to inspection, all damage mechanisms are targeted, and all piping systems continue to be inspected. The NRC staff concludes that the segment identification in RG 1.178, .as used as an accounting tool, is not needed within the generic population of HSS welds. The risk-informed search for HSS segments based on a flooding PRA divides up piping systems into segments based on which is consistent with the segment definition in RG 1.178. Since the licensee's proposed method accomplishes the same objective as the approved methods without requiring that segments be identified and defined for all piping within the scope of the RIS_B program, it is therefore acceptable to the NRC staff.

-8-3.2.5 Risk Categorization Sections 2(a)(1) through 2(a)(4) of N-716 identify the portions of systems that should be generically classified as HSS, and Section 2(a)(5) requires a search for plant-specific HSS segments. Application of the guideline in Section 2(a)(5) in N-716 identifies plant-specific piping segments that are not assigned to the generic HSS category but that are risk-significant at a P,articular plant. N-716 requires that any segment with a total estimated CDF greater than 1 x 10-6 per year be assigned to the HSS category. The licensee augmented this N-716 metric on CDF with the requirement to also assign the HSS category to any segment with a total estimated LERF greater than 1 x 1 o-7 per year. The licensee stated that these guideline values are suitably small and consistent with the decision guidelines for acceptable changes in CDF and LERF found in RG 1.17 4.

• In its March 19, 2013, submittal, the licensee clarified that these ancillary metrics were added as a defense-in-depth measure to provide a method of ensuring that any plant-specific locations that are important to safety are identified. All piping that has inspections added or removed per N-716 is required to be included in the change-in-risk assessment and an acceptable estimate is used to demonstrate compliance with the acceptance guidelines in RG 1.17 4. The ancillary metrics and guidelines on CDF and LERF are only used to add HSS segments and not, for example, to remove system parts generically assigned to the HSS in Section 2(a)(1) through 2(a)(4) in N-716. The NRC staff determined that a plant-specific analysis to identify plant-specific locations that are important to safety is a necessary element of the RI-ISI program development. The results of a plant-specific risk categorization analysis provide confidence that the goal of inspecting the more risk-significant locations is met while permitting the use of generic HSS system parts to simplify and standardize the evaluation. Satisfying the guidelines in Section 2(a)(5) in N-716 requires confidence that the flooding PRA-is capable of successfully identifying all, or most, of the significant flooding contributors to risk that are not included in the generic results. RG 1.200 states that meeting the attributes of an NRC-endorsed industry PRA Standard may be used to demonstrate that a PRA is adequate to support a risk-informed application. RG 1.200 further states that an acceptable approach that can be used to ensure technical adequacy is to perform a peer review of the PRA. In its March 19, 2013, submittal,* the licensee stated that, in November 2010, the PBNP PRA model underwent a full scope peer review by Pressurized-Water Reactor Owners Group against the available versions of the ASME PRA Standard and RG 1.200, Revision 2. In August 2011, independent contractors performed a focused scope peer review of the updated .Internal Flooding Supporting Requirements (SRs) against the available versions of the ASME PRA Standard and Regulatory Guide 1.200, Revision 2. Another peer review of the updated PRA excluding Internal Flooding was conducted in October 2011. This review was conducted against the available version of the ASME PRA Standard and Regulatory Guide 1.200, Revision 2. The PBNP PRA model was further updated in March 2013. This updated PRA model was used in the licensee's submittal. The licensee described the resolution of the findings from the peer review of the PBNP flooding analysis. The NRC staff agrees that further resolution of the findings would have a minimal or conservative impact on the risk-insights for this application. The licensee also reported a review of the flooding PRAto identify any piping* whose failure could cause flooding that could significantly impact safety significant components.

-9 -The results of its flooding analysis identified two service water piping welds in the cable spreading room as HSS due to exceeding the above CDF criteria. One weld will be selected for inspection. This piping is shared between both PBNP units. The NRC staff concludes that the CDF and.LERF metrics proposed by the licensee are acceptable because they address the risk elements that form the basis for risk-informed applications (i.e., core damage and large early release). The NRC staff accepts the proposed guideline values because these ancillary guidelines are applied in addition to the change-in-risk acceptance guidelines in RG 1.17 4, and only add plant-specific HSS segments to the RIS_B program (i.e., they may not be used to reassign any generic HSS segment into the LSS category). \ . The NRC staff finds that the risk categorization performed by the licensee provides confidence that HSS segments have been identified. Sections 2(a)(1) through 2(a)(4) in N-716, which

• identify generic HSS portions of systems, were applied to PBNP piping. The licensee's PRA used to fulfill the guideline in Section 2(a)(5) was performed using a PRA of adequate technical quality based on consistency between the PRA and the applicable characteristics of the endorsed industry standard. 3.2.6 lnspection/NDE Selection The licensee's submittals discuss the impact of the proposed RIS_B application on the various augmented inspection programs. The EPRI TR and N-716 contain no provisions for changing the FAC augmented program developed in response to NRC Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning." The PBNP FAC program is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RIS_B program. The PBNP augmented inspection program implemented in response to NRC Bulletin 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems," will be subsumed by the .f31S_B Program. In the March 19, 2013, submittal the licensee indicated the four Unit 2 steam generator hot leg and cold leg primary nozzle to safe-end welds are the only Alloy 82/182 dissimilar metal welds at PBNP. These welds were factory clad with Alloy 52/152 weld material to enhance their resistance to primary water stress corrosion cracking. The licensee stated these welds will be examined in accordance with the augmented examination requirements of ASME Code Case N-770-1 and 10 CFR 50.55a(g)(6)(ii)(F). In its September 5, 2013, submittal, the licensee states that these four welds are not counted as exams selected to satisfy the RIS_B program. The NRC staff finds the licensee's approach to the integration of the proposed program, with augmented inspection programs as described above, to be acceptable because it is consistent with the EPRI TR. Additionally, N-716 contains requirements that inspection locations be divided among the systems under consideration and that certain percentages of inspections will be conducted in speCific locations. In its relief request, the licensee has addressed these issues. The NRC staff

-10-finds this acceptable because the information provided in the relief request is consistent with that required by the EPRI TR which has been reviewed and approved by the NRC. The NRC staff reviewed the tables provided in the relief request which address degradation mechanisms, failure potential, and the number of welds selected for evaluation. The staff finds that the data contained in these tables is consistent with the requirements of the EPRI TR and, therefore, is acceptable. 3.2.7 Risk Impact Assessment The licensee uses a change-in-risk estimation process approved by the NRC staff in EPRI TR-112657. The change-in-risk assessment in the EPRI TR-112657 permits using each segment's CCDP and CLERP or, alternatively, placing each segment into high-, medium-, or low-consequence "bins" and using a single bounding CCDP and CLERP for all segments in each consequence bin. N-716 also includes both alternatives and the bounding values to be used in the bounding analysis are the same as those approved for use in the EPRI TR-112657. The licensee uses the alternative of placing each segment into consequence bins and using the associated bounding values for all segments in each bin during the change-in-risk assessment. In the submittal, the licensee identified different types of pipe failures that cause major plant transients such as thos'e causing loss-of-coolant accidents (LOCAs), isolable LOCAs, potential LOCAs and feedwater breaks. Conservative CCDP estimates were developed from the PRA for these initiating events. When the break scenario was not appropriately modeled in the PRA, the licensee developed scenarios based on the PRA results and associated plant-specific equipment failures. The NRC staff concludes that the scenarios described are reasonable because they are modeled in the PRA or include the appropriate equipment failure modes that cause each sequence to progress, and the licensee uses generally accepted values for those failure modes. Based on these estimates, the segments were assigned into the appropriate consequence bin. The licensee relied on its flooding analysis to identify the appropriate consequence bin* for welds whose failure does not cause a major plant transient and for which a consequence estimate is required. As discussed above, the licensee performed its flooding analysis consistent with ASME PRA Standard. Only segments with locations at which an inspection is being discontinued or added need to be included in the change-in-risk calculation; therefore, limiting the consequence evaluation to segments that are inspected is acceptable. Section 5 in N-716 requires that any piping that has NDE inspections1 added or removed per N-716 be included in the change-in-risk assessment. The licensee nominally used the upper-bound estimates for CCDP and CLERP. Acceptance criteria provided in Section 5(d) of N-716 include limits of 1 x 1 o-7 per year and 1 x 1 o-a per year for increase of CDF and LERF for each system, and limits of 1 x 1 o-6 per year and 1 x 1 o-7 per year for the total increase in CDF and LERF associated with replacing the ASME Code Section XI program with the RIS_B program. These guidelines and guideline values are consistent with those approved by the NRC staff in EPRI TR-112657 and are, therefore, acceptable. N-716 requires no estimated risk increase for discontinuing surface examinations at locations that are not susceptible to outside diameter attack (e.g., external chloride stress-corrosion cracking). The NRC staff determined during the review and approval of EPRI TR-112657 that surface exams do not appreciably contribute to safety and need not be included in the change in risk evaluation and, therefore, exclusion of surface exam from the change-in-risk evaluations is acceptable.

-11 -The change-in-risk evaluation approved in EPRI TR-112657 is a final screening to ensure that a licensee replacing the Section XI program with the risk-informed alternative evaluates the potential change-in-risk resulting from change in method and implements it only upon determining, with reasonable confidence, that any increase in risk is small and acceptable. The licensee's method is consistent with the approved method in EPRI TR-112657 with the exception that the change-in-risk calculation in N-716 includes the risk increase from discontinued inspection in LSS segments. The licensee conducted a review and verified that the LSS piping was not susceptible to water hammer. LSS piping may be susceptible to FAC; however, the examination for FAC is performed per the FAC program. The RIS_B Program credits and relies upon this plantD augmented inspection program to manage this damage mechanism. In lieu of conducting a formal degradation mechanism evaluation for all LSS piping (e.g., to determine if thermal fatigue is applicable), these locations were conservatively assigned to the medium failure potential category for use in the change-in-risk assessment. The NRC staff concludes that the licensee's method described in the submittal is acceptable because the deviation from the approved method in EPRI TR-112657 expands the scope of the calculated change-in-risk providing confidence that the less detailed analyses of LSS segments required by N-716 do not result in an unanticipated and potentially unacceptable risk increase. The licensee provided the results of the change-in-risk calculations in the submittal and noted that results indicate a small and acceptable increase in risk and that all the estimates satisfy both the system level and the total CDF and LERF guidelines. Therefore, the NRC staff finds the change in risk acceptable for this application. 3.2.8 Implementation Monitoring and Feedback The program implementation process described in the EPRI TR and RG 1.178 requires that a licensee's RI-ISI program have a schedule for inspecting all piping segments categorized as safety significant. It further states that the inspection interval will normally be that prescribed by Section XI of the Code but that certain degradation mechanisms may require the interval to be altered. The performance monitoring category requires that a licensee's RI-ISI program be updated based on: changes in plant design features, changes in plant procedures, equipment performance changes, examination results, and plant or industry operating experience. Additionally, a licensee must update its program periodically to correspond to the requirements contained in Section XI of the Code, Inspection Program B. The corrective action category requires a corrective action program that is consistent with the requirements of Section XI of the Code for both Code class and non-Code class piping. Information concerning this topic was obtained from the relief request itself and from Sections 6 and 7 of the code case. The code case information was used by the NRC in this review based on the licensee's statement that it would develop implementation procedures for its program in accordance with the code case. In its relief request the licensee stated that it has a corrective action program and that it will review the RI-ISI program periodically as required by the Code or more frequently as directed by the NRC, or industry or plant-specific feedback. Sections 6 and 7 of the Code address inspection frequency and program updates. These s*ections indicate that inspection frequencies should normally be in accordance with Code requirements, and that

-12-updates should be made on a Code-dictated schedule or more frequently in response to plant and industry events or information. The NRC finds the licensee's approach to implementing the program to be acceptable because, in accordance with RG 1.178, the licensee indicated that it inspects components on a frequency based on the Code, that it has a corrective action program, and that it updates the program periodically and in response to plant and industry events and information. 3.2.9 Examination Methods Section 4 of the EPRI TR addresses the NDE techniques which must be used in a RI-ISI program. This section emphasizes the concept that the inspection technique utilized must be specific to the degradation mechanism expected. Table 4.1 of the EPRI TR summarizes the degradation mechanisms expected and the examination methods which are appropriate. Specific references are provided to the Code concerning the manner in which the examination is conducted and the acceptance standard. The code case addresses the issue of degradation mechanism/inspection technique in Table 1. Like Table 4.1 of the EPRI TR, Table 1 of the code case lists degradation mechanism and corresponding inspection techniques. This table also provides references to the Code concerning the manner in which the examination is conduced and the acceptance standard. In its relief request, the licensee stated that implementation of the RI-ISI program will conform to the code case, i.e., each HSS piping segment will be assigned to the appropriate item number within Table 1 of the code case. The NRC staff finds this acceptable because proper assignment of piping segments into Table 1 will ensure that appropriate inspections to detect the degradation mechanism under consideration are conducted. The staff finds this approach acceptable because it is consistent with the EPRI TR which has been reviewed and approved by the NRC, and the code case includes additional code item numbers to assign NDE requirements to all HSS locations including those segments where no degradation mechanism has been identified. Pursuant to 10 CFR 50.55a(a)(3)(i), alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if the licensee demonstrates that the proposed alternatives will provide an acceptable level of quality and safety. In this case, the licensee has proposed to use an alternative to the risk-informed process described in NRC-approved EPRI TR-112657. The implementation strategy is consistent with the RG 1.178 guidelines because the number and location of inspections is a product of a systematic application of the risk-informed process. Other aspects of the licensee's lSI program, such as system pressure tests and visual examination of piping structural elements will continue to be performed on all Class 1, 2, and 3 systems in accordance with ASME Code,Section XI. This provides a measure of continued monitoring of areas that are being eliminated from the NDE portion of the lSI program. As required by EPRI TR-112657 methodology, the existing ASME Code performance measurement strategies will remain in place. In addition, the N-716 methodology provides for increased inspection volumes for those locations that are included in the NDE portion of the program.

-13-RG 1.17 4 establishes requirements for risk-informed decisions involving a change to a plant's licensing basis. RG 1.178 establishes requirements for risk-informed decisions involving alternatives to the lSI program requirements of 10 CFR 50.55a(g), and its directive to follow the requirements of the ASME Code,Section XI. EPRI TR-112657 RI-ISI methodology contains details for developing an acceptable RI-ISI program. N-716, modified as described by the licensee in its submittal, describes a methodology similar to EPRI TR-112657 but with differences as described in this SE. The NRC staff has evaluated each of the differences and that the licensee's proposed methodology, when applied as described, meets the intent of all the steps endorsed in EPRI TR-112657, is consistent with the guidance provided in RG 1.178, and therefore satisfies the guidelines established in RG 1.17 4. 4.0 CONCLUSION As set forth above, the NRC staff determines that the proposed alternative provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a, and is in compliance with the ASME Code's requirements. Therefore, the NRC staff authorizes the use of the proposed RI-ISI RIS_B program as described in the licensee's submittals at the PBNP, Units 1 and 2, for the fifth 1 0-year lSI interval. Acco'rdingly, the NRC staff concludes that the proposed RIS_B program will provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i) for the proposed alternative to the piping lSI requirements with regard to (1) the number of locations, (2) the locations of inspections, and (3) the methods of inspection. Therefore, the proposed RI-ISI RIS_B program is authorized for its fifth 1 0-year lSI interval pursuant to 10 CFR 50.55a(a)(3)(i) on the basis that this alternative will provide an acceptable level of quality and safety.

• All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in the subject request remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector. 5.0 REFERENCES 1. ASME Code Case N-716, "Alternative Piping Classification and Examination Requirements,Section XI, Division 1 ," ASME, New York, New York, published April19, 2006. 2. EPRI TR-112657, Revision 8-A, "Revised Risk-Informed lnservice Inspection Evaluation Procedure," Final Report dated December 1999 (ADAMS Accession No. ML0134701 02). 3. ASME/ANS RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level I I Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Copyright© 2009, ASME, New York, New York. Principal Contributors: Date: January 30, 2014 Keith Hoffman Jigar Patel E. McCartney -2-If you have any questions, please contact Terry Beltz, Senior Project Manager, at (301) 415-3049. Docket Nos. 50-266 and 50-301 Enclosure: Safety Evaluation cc w/encl: Distribution via Listserv Distribution: PUBLIC LPL3-1 R/F RidsNrrLAMHenderson Resource RidsNrrPMPointBeach Resource RidsNrrDorllpl3-1 Resource RidsAcrsAcnw_MaiiCTR Resource ADAMS Accession No* ML14014A205 .. Sincerely, /RAJ Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDeEpnb Resource RidsNrrDraApla Resource* RidsRgn3MaiiCenter Resource RidsOgcRpMaiiCenter MReisiFard, NRR KHoffman, NRR

• via memo dated OFFICE LPL3-1/PM LPL3-1/LA DE/EPNB/BC

• DRA/APLA/BC

• LPL3-1/BC NAME TBeltz MHenderson Tlupold MHamzehee RCa rison DATE 01/24/14 01/23/14 11/25/13 12/04/13 01/30/14 OFFICIAL RECORD COPY