ML14188C348

Proposed Relief Request 13-MN-002

ML14188C348
Person / Time
Site:	McGuire, Mcguire
Issue date:	07/14/2014
From:	Robert Pascarelli Plant Licensing Branch II
To:	Capps S Duke Energy Carolinas
Miller G
References
TAC MF2609, TAC MF2610
Download: ML14188C348 (16)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Steven D. Capps Vice President McGuire Nuclear Station Duke Energy Carolinas, LLC 12700 Hagers Ferry Road Huntersville, NC 28078 July 14, 2014

SUBJECT:

MCGUIRE NUCLEAR STATION, UNITS 1 AND 2-PROPOSED RELIEF REQUEST 13-MN-002 {TAC NOS. MF2609 AND MF2610)

Dear Mr. Capps:

By letter dated August 13, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13234A069), as supplemented by letter dated March 3, 2014 (ADAMS Accession No. ML14073A464), Duke Energy Carolinas, LLC (the licensee) submitted RR 13-MN-002, to the U.S. Nuclear Regulatory Commission (NRC) as an alternative to the requirements of the American Society of Mechanical Engineering (ASME) Boiler and Pressure Vessel Code {B&PV Code),Section XI, 2007 Edition through 2008 Addenda. The alternative would implement a risk-informed/safety-based (RIS_B) inservice inspection program for piping at the McGuire Nuclear Station, Units 1 and 2 (MNS1 and MNS2). The proposed program is based, in part, on the ASME,Section XI, Code Case N-716 The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the proposed alternative provides an acceptable level of quality and safety. Therefore, the NRC staff authorizes the proposed alternative in accordance with 10 CFR 50.55a(a)(3)(i) for the remainder of the fourth 1 0-year inservice inspection interval at MNS1 and MNS2. The NRC staff's approval of the licensee's RIS_B inservice inspection program does not constitute approval of Code Case N-716.

. All other ASME Code,Section XI, requirements, for which relief was not specifically requested and authorized herein by the NRC staff, remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.

S. If you have any questions, please contact the Project Manager, G. Edward Miller at 301-415-2481 or via e-mail at ed.miller@nrc.gov.

Docket Nos. 50~369 and 50-370

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ Sincerely, Robert J. Pascarelli, Chief' Plant Licensing Branch 11-1 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 NO. 13-MN-002 DUKE ENERGY CAROLINAS, LLC MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By letter dated August 13, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13234A069), as supplemented on March 3, 2014 (ADAMS Accession No. ML14073A464), Duke Energy Carolinas, LLC (Duke Energy or the licensee) submitted Relief Request 13-MN-002, "Risk-Informed lnservice Inspection Program, Fourth lnservice Inspection Interval" (the submittal), pursuant to paragraph 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations (10 CFR). The relief request proposed to use a risk-informed/safety based (RIS_B) program as an alternative to the requirements in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI inservice inspection (lSI) program for Class 1 and 2 piping for the McGuire Nuclear Station (MNS). The basis for the relief request is that the RIS_B program provides an acceptable level of quality and safety. The licensee requested implementation of the RIS_B program for the fourth lSI interval for McGuire, Units 1 and 2. The fourth lSI interval for McGuire, Unit 1, began on December 1, 2011, and ends on November 30, 2021, and for McGuire, Unit 2, the fourth lSI interval is scheduled to start on July 15, 2014, and ends on July 14, 2024. The proposed RIS_B program is based on the ASME Code Case N-716, "Alternative Piping Classification and Examination Requirements,Section XI Division 1" (Code Case N-716). The provisions of Code Case N-716 define additional requirements for Class 3 or non-Class piping.

Code Case N-716 is founded in large part on the risk-informed inservice inspection (RI-ISI) process as described in the Electric Power Research Institute (EPRI) Topical Report (TR)-112657, Revision B-A, "Revised Risk-Informed lnservice Inspection Evaluation Procedure" (ADAMS Accession No. ML013470102). EPRI TR-112657, Revision B-A, was previously reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC).

Code Case N-716 has not been endorsed for generic use by the NRC. The licensee refers to the methodology described in Code Case N-716 instead of describing the details of the methodology in the relief request. Duke Energy has, however, modified the methodology described in Code Case N-716 while developing its proposed RIS_B program. When the methodology used by the licensee is accurately described in Code Case N-716; this safety Enclosure evaluation (SE) refers to the details found in Code Case N-716. When the methodology used by the licensee deviates from or expands upon the methodology described in Code Case N-716, this SE refers to details in the relief request. Therefore, Code Case N-716 is incorporated in this SE only as a source for some of the detailed methodology description as needed and the NRC staff is not endorsing the use of Code Case N-716 in this SE.

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 set forth in the ASME Code to the extent practical within the limitations of the design, geometry, and materials of construction of the components.

Paragraph 50.55a(g) of 10 CFR 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 written relief has been granted by the NRC.

The regulations, in 10 CFR 50.55a(g)(4)(i)-(ii), also require during the first 1 0-year lSI interval and during subsequent intervals, that the licensee's lSI program comply with the requirements in the latest edition and addenda of the ASME Code incorporated by reference into 10 CFR 50.55a(b) 12 months before the start of the 120-month inspection interval, subject to the conditions listed in 10 CFR 50.55a(b). The applicable edition of Section XI of the ASME Code for McGuire, Unit 1, during Period 1 of the fourth lSI interval is the 1998 Edition with the 2000 Addenda. For Periods 2 and 3 of the fourth lSI interval, the applicable edition of Section XI of the ASME Code is the same as the McGuire, Unit 2, fourth lSI interval. The applicable code requirement during the fourth lSI interval for McGuire, Unit 2, is the 2007

• Edition with the 2008 Addenda of the ASME Section XI Code.

Pursuant to 10 CFR 50.55a(g), a certain percentage of ASME Code Category 8-F, B-J, C-F-1, and C-F-2 pressure retaining piping welds must receive lSI during each 10-year lSI interval.

The ASME Code requires 100 percent(%) of B-F welds and 25% of B-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% of non-exempt welds are selected for volumetric or surface examination, or both.

According to 10 CFR 50.55a(a)(3), the NRC may authorize alternatives to the requirements of 10 CFR 50.55a(g), if an applicant or licensee demonstrates that the proposed alternatives would_provide an acceptable level of quality and safety, or that compliance with the specified requirements of 10 CFR 50.55a would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The licensee states that Code Case N-716 is founded in large part on the RI-ISI process described in EPRI TR-112657, Revision 8-A, which was previously reviewed and approved by the NRC.' The licensee further states that the risk-informed application based on CC N-716 meets the intent and principles of Regulatory Guide (RG) 1.17 4, "An Approach for Using Probabilistic Risk Assessm~nt In Risk-Informed DeCisions on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML100910006), and RG 1.178, "An ~pproach for Plant-Specific Risk-Informed Decisionmaking-lnservice Inspection* of Piping" (ADAMS Accession No. ML032510128).

The objectives of the licensee's proposed RIS_B program are to identify high safety significant (HSS) piping at the plant, to determine the locations to be inspected within the identified piping, and to identify appropriate inspection methods. The licensee's proposed program will inspect ten percent of HSS welds using an inspection method capable of detecting the degradation mechanism(s) of concern and most likely to occur at each location to be inspected.

In general, the licensee simplified the EPRI TR-112657 method because the licensee does not evaluate system parts that have been generically identified as HSS, and uses probabilistic risk assessment (PRA) analyses to evaluate in detail only system parts that cannot be screened out as low safety significant (LSS). The NRC staff reviewed and evaluated the licensee's proposed RIS_B program based on guidance and acceptance criteria provided in the following documents:

RG 1.174; RG 1.178; RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (ADAMS Accession No. ML090410014);

NUREG-0800, Chapter 3.9.8; "Standard Review Plan for the Review of Risk-Informed*

lnservice Inspection of Piping" (SRP 3.9.8) (ADAMS Accession No. ML032510135; and EPRI TR-112657, Revision 8-A Since the issuance of the NRC Staff SE for EPRI TR-112657 (ADAMS Accession No. ML993190477), several instances of primary water stress-corrosion cracking (PWSCC) of Alloy 82/182 dissimilar metal welds have occurred at pressurized-water reactors (PWRs). In response, the NRC sent a letter (ADAMS Accession No. ML053480359) to the Chairman of the ASME Subcommittee on Nuclear lnservice Inspection, stating that the operating experience with leakage and flaws caused by PWSCC at PWRs supports a position that current ASME Code inspection requirements are not sufficient for managing PVVSCC-susceptible butt welds in the reactor coolant pressure boundary for PWRs. This letter represents a departure from the NRC staff's conclusions about PWSCC in theSE for EPRI TR-112657. The NRC staff is including this information to demonstrate that as issues arise, modifications to RI-ISI programs and RIS_B

• programs may be warra11ted, as identified in the NRC approval of EPRI TR-112657. The licensee stated in Section 2.2, "Augmented Programs," of the submittal that the requirements of

. ASME Code Case N-770-1, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1,"will be used for the inspection and management of PWSCC susceptible welds and will supplement the RIS_B program selection process.

3.0 TECHNICAL EVALUATION

RG 1.17 4 provides guidance on the use of PRA findings and risk insights in support of licensee

• request for changes to a plant's licensing basis. RG 1.178 describes an 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. EPRI TR-112657 provides a detailed methodology that the NRC staff has previously concluded will result in an acceptable RI-ISI program. The RIS_B program proposed by MNS also incorporates risk insights to focus inspection on more important locations which differs in several respects from the methodology in EPRI TR-112657. This SE describes and evaluates the differences between the endorsed methodology in EPRI TR-112657 and the proposed RIS_B methodology to reach a conclusion about the acceptability of the proposed method.

An acceptable RI-ISI program replaces the number and locations of non-destructive examination (NDE) inspections based on ASME Code Section XI requirements by the number and locations of these inspections based on the RI-ISI guidelines. The proposed RIS_B program permits alternatives to the requirements of IWB-2420, IWB-2430, and IWB-2500 (Examination Categories B-F and B-J) and IWC-2420, IWC-2430, and IWC-2500 (Examination Categories C-F-1 and C-F-2), or as additional requirements for Subsection IWD. Also, the proposed RIS_B program may be used for lSI and preservice inspection of Class 1, 2, 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. Visual examinations (VT-2) are scheduled in accordance with the MNS pressure and leak test program which remains unaffected by the proposed RIS_B program.

The process described in EPRI TR-112657 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 lnspection/NDE Selection Risk Impact Assessment Implementation Monitoring and Feedback These steps 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 adequate protection of public health and safety. In general, the methodology in Code Case N-716 replaces a detailed evaluation of the safety significance of each pipe segment with a generic population of HSS segments, followed by a flooding analysis to identify any plant-specific HSS segments. The flooding analysis was performed in accordance with Part 3 of ASME/American Nuclear Society (ANS) RA-Sa-2009, Standard for Levei1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addenda to ASME/ANS RA-S-2008 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 the licensee has applied to develop its program with the steps from EPRI TR-112657.

3.1 Scope Definition The scope of evaluation to support RIS_B program development, and of the proposed changes, in Code Case N-716 and the licensee's submittal 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 SRP 3.9.8 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 MNS's evaluation included all piping where ASME inspections could be discontinued by providing assurance that the change in risk estimate would, as a minimum, capture the risk increase associated with implemented the RIS_B program in lieu of the ASME program. The change in risk estimate is used by the licensee to demonstrate that ~isk increases are small. RG 1.178 identifies different grouping 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 Code Case N-716 and as evaluated by the licensee is consistent with the definition of full-scope in RG 1.178. Therefore, the NRC staff concludes that the full-scope" extent of the piping included in the RIS_B program changes satisfies the guidelines in SRP 3.9.8 and RG 1.178, and is acceptable.

• 3.2 Consequence Evaluation The methodology described in RG 1.178 and EPRI TR-112657 divide all piping within the scope of the proposed EPRI-TR 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 EPRI TR-112657 that yield equivalent results. The consequences are used to determine the safety significance of segments.

In contrast to the EPRI TR-112657 methodology, Code Case N-716 does not require that the consequence of each segment failure be estimated to determine the safety significance of piping segments. Instead, Code Case 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 assigned as HSS because there is no higher safety-significance category to which the system part can be assigned and degradation mechanisms, not consequence, are used to select inspection locations in the HSS weld population. The licensee's PRA is subsequently used to search for any additional, plant-specific HSS segments that are not included in the generic HSS population.

Sections 2(a)(1) through 2(a)(4) in Code Case 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 RI-ISI programs were 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, EPRI TR-112657 and SRP 3.9.8.. Therefore, the generic results are derived from acceptable analyses.

Section 2(a)(5) in Code Case 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 used its PRA of pressure boundary failures (internal flooding analysis) to search for additional plant-specific HSS segments. The internal flooding analysis considered both the direct and indirect effects 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 internal 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 SRP 3.9.8 and is, therefore, acceptable.

3.3 Degradation Mechanism Evaluation EPRI TR-112657 requires a determination of susceptibility to all degradation mechanisms of every weld within the scope of the proposed program. The degradation mechanisms which

$hould be identified are described in EPRI TR-112657. This information is used to support the safety significance determination for all segments, to target inspections toward the locations with degradation mechanisms in.the segments that require inspections, and to provide estimates of weld failure frequencies to support the change in risk evaluation. Once a segment is placed in the LSS category, the degradation mechanisms at the welds in that segment are used no further in the development of the EPRI TR-112657 RI-ISI program because no inspections are required in LSS segments and the discontinued inspections in LSS segments are not included in the change in risk estimate.

Code Case N-716 identifies a generic population of HSS welds, followed by a search for plant-specific HSS welds and requires a determination of the susceptibility to all degradation mechanisms of all welds assigned generically as HSS. The degradation mechanisms considered in Code Case N-716 are consistent with those identified in EPRI TR-112657 which the NRC staff has previously concluded is a sufficiently comprehensive list of the applicable degradation mechanisms.

Pipe failure frequencies are used in the screening analysis searching for plant-specific HSS welds described in Section 3.2 of this SE, and then in the change in risk estimate. If welds are exposed to any degradation mechanism, aside from flow-accelerated corrosion (FAC) and water hammer, a medium failure frequency is assigned to those welds. If welds are exposed to FAC or water hammer, a high failure frequency is assigned. Finally, if there are no damage mechanisms, a low failure frequency is assigned. This is consistent with the approved methodology in EPRI TR-112657.

The licensee stated in Section 3.4.1 of Reference 1 that the presence of FAC was adjusted for in the quantitative analysis by excluding its impact on the failure potential rank. The exclusion of the impact of FAC was performed because the licensee manages this damage mechanism through the plant augmented inspection program for FAC per Generic Letter (GL) 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning" (ADAMS Accession No. ML031200731). EPRI TR-112657 notes that the plant's existing FAC program in response to GL 89-09 would not be affected by the RI-ISI program. The.NRC staff found this to be'acceptable in Section 3.2.1 of theSE approving EPRI TR-112657:

For HSS welds, and in response to NRC staff questions, the licensee confirmed that HSS piping is not susceptible to water hammer. For LSS welds, the licensee stated in Section 3.4.1 of their original submittal that a review was conducted to verify that LSS piping was not susceptible to water hammer. In lieu of conducting a degradation mechanism evaluation for all LSS piping, the licensee assigned these locations to the medium failure potential for the purpose of assigning a failure frequency to calculate the change in risk. This results in conservative failure frequency estimates because these failure frequencies would always be equal or greater than those used in the analysis if the susceptibility of all LSS welds to all degradation mechanisms was determined.

The approach proposed by the licensee used failure frequency estimates that reflected applicable degradation mechanisms while searching for plant-specific HSS piping. Failure frequency estimates are further refined for use in the change in risk estimate by identifying degradation mechanisms at all HSS welds and in LSS welds with potential high failure '

frequency (i.e., susceptible to FAC or water hammer). Therefore, the NRC staff concludes that the screening evaluation relying on a plant-specific update of generic failure frequencies, followed by a bounding analysis for specific welds where inspections will be added or discontinued, is acceptable because the licensee's methods fulfills the requirements for identifying locations that should be inspected (i.e., identifying additional plant-specific HSS segments) and developing a bounding estimate for the change in risk.

3.4 Piping Segment Definition Previous guidance on RI-ISI including RG 1.178, SRP 3.9.8, and both approved industry methodologies 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 the segments is to group welds so that consequence analyses can be done for the smaller number of segments instead of for each weld. Sections 2(a)(1) to 2(a)(4) in Code Case N-716 identifies 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 Code Case N-716 uses a PRAto identify plant-specific piping that might be assigned HSS. The process described by the licensee to search for plant-specific HSS piping first identifies zones that may be sensitive to flooding, and then evaluates 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 in the same zone. 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 category, 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. Case Code N-716 generically defines a large population of welds as HSS. An additional population of welds may be added based on the risk-informed search for plant-specific HSS segments. When complete, the Code Case N-716 process yields a well-defined population of HSS welds from which inspections must be selected. This accomplishes the same objective as accounting for each weld throughout the*

analysis by using segments. Code Case N-716, as applied by the licensee, 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. for the MNS RI-ISI program. The risk-informed search for HSS segments based on a flooding PRA divides up piping systems into segments based on consequences, which is consistent with the segment definition in RG 1.178.

Therefore, 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.

3.5 Risk Categorization Sections 2(a)(1) through 2(a)(4) in CC N-716 identify the portions of systems that should be generically assigned 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 CC N-716 identifies plant-specific piping segments that are not assigned to the generic HSS category but that are risk-significant at a particular plant. Code Case N-716 requires that any segment with a total estimated CDF greater than 1 x 1 o*6 per year be assigned to the HSS category. The licensee augmented this Code Case N-716 metric on CDF with the requirement to also assign the HSS category to any segment with a total estimated large early release frequency (LERF) greater than 1 x 1 o*7 per year. 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 the submittal, the licensee clarified that these ancillary metrics (i.e., CDF and LERF) 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 Code Case N-716 is required to be included in the change in risk assessment and an acceptable change in risk estimate is used to demonstrate compliance with the acceptance guidelines in RG 1.174. 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) of CC N-716.

The NRC staff concludes that a plant-specific analysis to identify plant-specific locations that are important to safety is a necessary element of RI-ISI prograni development. The results of the 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. Any evaluation that categorizes the safety significance of structures, systems, and components requires metrics and guideline values, such as the Fussei-Vessley and risk achievement worth guidelines endorsed in RG 1.201, "Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance," Revision 1 (ADAMS Accession No. ML061090627). Such metrics are subordinate to the change in risk metrics in RG 1.17 4 which are used to determine whether the increase in risk associated with a proposed change is small and consistent with the intent of the Commission's Safety Goal Policy Statement (51 FR 28044).

Satisfying the guidelines in Section 2(a)(5) 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 compliance with the attributes of an NRC-endorsed industry PRA standard (the licensee performed a gap assessment using the ASME/ANS RA-Sa-2009 PRA standard in its application) 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 Attachment A of Reference 1, the licensee described the resolution of the findings (i.e., a subset of the facts and observations) from the focused-scope peer review of the MNS internal flooding analysis. The licensee reviewed the results of its internal flooding analysis and found that the Class 3 nuclear service water system piping in the auxiliary feedwater pump room was HSS due to CDF being greater than 1 x 1 o-6 per year.

The NRC staff concludes that the CDF and LEAF 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 (i.e., CDF and LEAF) 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., that may not be used to reassign any generic HSS segment into the LSS category).

3.6 lnspection/NDE Selection The licensee's submittals discuss the impact of the proposed RIS_B application on the various augmented inspection programs.

The licensee's augmented inspection program for high energy line breaks, implemented per Table 3-24 of the UFSAR, i*s not affected or changed by the RIS-B Program.

The plant augmented inspection program for flow accelerated corrosion per Generic Letter 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning (ADAMS Accession No. ML031470660), is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RIS-B Program.

The augmented inspection program for localized corrosion per NRC Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment (ADAMS Accession No. ML031150348), is used to detect localized corrosion damage mechanisms. Class 3 RN (nuclear service water) system piping in the auxiliary feedwater pump room was determined to be high safety significant. While the sampling percentages of Code Case N-716 will be applied to this piping, it will be inspected under the existing effective localized corrosion program, per Section 3.6.7 of EPRI TR-112657.

The requirements of ASME Code Case N-770-1, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 Or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, will be used for the inspection and management of PWSCC susceptible welds and will supplement the RIS_B Program selection process.

The licensee has conducted an evaluation in accordance with MRP-146, Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-/so/able Reactor Coolant System Branch Lines, and these results have been incorporated into the RIS_B Program.

The NRC staff finds the licensee's approach to the integration of the proposed Rl-181 program with augmented inspection programs as described above acceptable because it is consistent with the EPRI TR.

Additionally, Code Case 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 finds this acceptable because the information provided in the relief request is consistent with that required by EPRI TR-112657 which has been reviewed and approved by the NRC.

In addition, in the additional information provided in the March 3, 2014 submittal, the licensee stated that weld selections will consider whether >90 percent coverage can be obtained. If limitations are encountered, the licensee will evaluate the availability of weld substitutions that will meet both the >90 percent coverage requirement and the Code Case N-716 requirements.

The NRC staff reviewed the tables provided in the relief request which address degradation mechanisms, failure potential and the number of welds selected for examination. The NRC staff finds that the data contained in these tables along with the additional information provided in the March 3, 2014 supplement is consistent with the requirements of EPRI TR-112657 and is therefore acceptable.

3.7 Risk Impact Assessment The licensee uses the change in risk estimation process approved by the NRC staff in EPRI TR-112657. The change in risk assessment in 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. Code Case N-716 also includes both alternatives, and the bounding values to be used in the bounding analyses are the same as those approved for use in EPRI TR-112657. The MNS Rl-181 program 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 first table of Section 3.4.1 of the licensee's original submittal, the licensee identified the different types of pipe failures that cause major plant transients such as those causing loss-of-coolant accidents (LOCAs), isolable LOCAs, potential LOCAs, and feedwater line breaks.

Conservative CCDP and CLERP estimates were developed from the PRA based on these initiating events. When the scenario was not explicitly modeled in the PRA, the licensee performed a bounding analysis using associated plant-specific equipment failure probabilities.

The NRC staff concludes that the scenarios described are reasonable because they identify the appropriate equipment failure modes that cause a 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 internal flooding analysis to identify the appropriate consequence bin for welds whose failure does not cause major plant transients and for which a consequence estimate is required. The licensee's internal flooding analysis indicated that the LSS Class 2 piping did not have high-consequence segments (lower bound of CCDP and CLERP of 1 x 1 o-4 and 1 x 1 o-5, respectively). The licensee placed the LSS Class 2 piping into the medium consequence bin.

Section 5 of Code Case N-716 requires that any piping that has NDE inspections1 added or removed per CC N-716 be inCluded in the change in risk evaluation. The licensee used the upper-bound values found in the first table of Section 3.4.1 of the licensee's original submittal for CCDP and CLERP. Acceptance criteria provided in Section 5(d) of Code Case N-716 include limits of 1 x 1 o-7 per year and 1 x 1 o-8 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.

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 that change 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 approved method in EPRI TR-112657 with the exception that the change in risk calculation in Code Case N-716 includes the risk increase from discontinued inspection in LSS segments. Based on the detailed analysis of every segment required by EPRI TR-112657, the NRC staff concluded that there is a high confidence that the total increase in risk from all

~ discontinued inspections in LSS segments would be negligible. 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 Code Case N-716 does not result in an unanticipated and potentially unacceptable risk increase.

. The licensee provided the results of the change in risk calculations in the submittal (found in the second (for MNS, Unit 1) and third (for MNS, Unit 2) tables of Section 3.4.1 of the licensee's original submittal) and noted that all the estimates satisfy both the system level and total CDF and LERF guidelines. Therefore, the NRC staff finds that any increase in risk is small and acceptable.

1 Code Case N-716 requires no estimated risk increase for discontinuing surface examinations at locations that are nQt susceptible to outside diameter attack (e.g., external chloride stress-corrosion cracking). The NRC staff determined during the review and approval ofEPRI 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.

3.8 Implementation Monitoring and Feedback The objective of this element of RGs 1.174 and 1.178 is to assess performance of the affected piping systems under the proposed RI-ISI program by implementing monitoring strategies that conform to the assumptions and analysis used in developing the RIS_B program. In Section 3.5 of Reference 1, the licensee states that upon approval of the RIS_B program, procedures that comply with the guidelines described in Code Case N-716 will be prepared to implement and monitor the program.

The list of possible changes includes all changes at the facility or in the PRA that could affect the evaluation* used to develop the RIS_B program and performing the reevaluation every lSI period coincides with the inspection periods in the inspection program requirements contained in the ASME Code,Section XI. The NRC staff finds that the proposed procedures are consistent with the performance monitoring guidelines described in RG 1.178 and are, therefore, acceptable.

3.9 Examination Methods Section 4 of EPRI TR-112657 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 EPRI TR-112657 summarizes the degradation mechanisms expected and the examination methods which are appropriate.

Specific references are provided to the ASME Code,Section XI concerning the manner in which the examination is conducted and the acceptance standard.

Code Case N-716 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 ASME Code,Section XI concerning the manner in which the examination is conduced and the acceptance standard.

In its relief request, the licensee states that the implementation of the RI-ISI program will conform to Code Case N-716, 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 NRC finds this approach acceptable because it is consistent with EPRI TR-112657 which has been reviewed and approved by the NRC and Code Case N-716 includes additional code item numbers to assign NDE requirements to all HSS locations including those segments where no degradation mechanism has beeri identified.

4.0 CONCLUSION

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 the EPRI-TR methodology, the existing ASME Code performance measurement strategies will remain in place. In addition, the Code Case N-716 methodology provides for increased inspection volumes for those locations that are included in the NDE portion of the program.

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. The EPRI-TR RI-ISI methodology contains details for developing an acceptable an RI-ISI program. The Code Case N-716, modified as described by the licensee in its submittal, describes a methodology similar to EPRI TR-112657 methodology but with several differences as described in this SE. The NRC staff has evaluated each of the differences and determined 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 1.174.

The NRC staff concludes that the proposed RIS_B program will provide an acceptable level of quality and safety pursuant to 1 0 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 program is authorized for its fourth 1 0-year lSI interval for MNS, Units 1 and 2, 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.

Principal Contributors: M. Audrain, NRR J. DeJesus, NRR Date: July 14, 2014

ML14188C348

• via email OFFICE NRR/DORULPL2-1 /PM NRR/DORULPL2-1/LA NRR/DRA/APLA/BC NRR/DE/EPNB/BC NRR/DORULPL2-1 /BC NAME GEMiller SFigueroa HHamzehee*

Tlupold (DAiley for)* A Pascarelli DATE 07/09/14 07/09/14 07/09/14 07/11/14 07/14/14