ML100610470

Safety Evaluation of VEGP-ISI-ALT-02 Version 1, Risk-Informed/Safety-Based Inservice Inspection Alternative for Class 1 and Class 2 Piping

ML100610470
Person / Time
Site:	Vogtle
Issue date:	03/03/2010
From:	Kulesa G Plant Licensing Branch II
To:	Ajluni M Southern Nuclear Operating Co
Boyle, Patrick, NRR/DORL/LPL2-1/415-3936
References
TAC ME1097, TAC ME1098, VEGP-ISI-ALT-02
Download: ML100610470 (15)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 l'v1arch 3, 2010 Mr. Mark J. Ajluni Manager, Nuclear Licensing Southern Nuclear Operating Company, Inc 40 Inverness Center Parkway Birmingham, Alabama 35201

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - RISK INFORMED SAFETY-BASED INSERVICE INSPECTION ALTERNATIVE FOR CLASS 1 AND CLASS 2 PIPING WELDS (TAC NOS. ME1 097 AND ME1098)

Dear Mr. Ajluni:

By letter dated April 15, 2009, as supplemented by letter dated October 12, 2009, Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted a relief request proposing a risk-informed safety-based inservice inspection (RIS_B) program as an alternative to a portion of its current inservice inspection (lSI) program for the Vogtle Electric Generating Plant (VEGP),

Units 1 and 2. The proposed program is based, in part, on the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Code Case N-716.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed relief request VEGP-ISI-ALT-02, and concludes that the proposed alternative provides an acceptable level of quality and safety. Therefore, the NRC staff authorizes the proposed alternative in accordance with Title 10 of the Code of Federal Regulations, 50.55a(a)(3)(i) for the licensee's third 10-year lSI interval. The NRC staff's approval of the license's RIS_B program does not constitute approval of Code Case N-716.

The NRC staff's safety evaluation is enclosed.

Sincerely, Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RISK-INFORMED/SAFETY BASED INSERVICE INSPECTION PROGRAM RELIEF REQUEST VEGP-ISI-ALT-02 SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2.

DOCKET NUMBERS 50-424 AND 50-425

1.0 INTRODUCTION

By letter dated April 15, 2009 (Reference 1), as supplemented by letter dated October 12, 2009 (Reference 2), Southern Nuclear Operating Company (SNC, the licensee), requested the U.S.

Nuclear Regulatory Commission's (NRC) authorization to implement a risk-informed/safety based inservice inspection (RIS_B) program as an alternative to a portion of its current inservice inspection (lSI) program for Vogtle Electric Generating Plant (VEGP), Units 1 and 2. VEGP 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. The licensee requested implementation of this alternative during the third 1O-year interval.

VEGP requests to implement an RIS_B program based, in part, on ASME Code Case N-716 (CC N-716), "Alternative Piping Classification and Examination Requirements,Section XI Division 1" (Reference 3). The provisions of CC N-716 may be used in lieu of the requirements of IWB-2420, IWB-2430, Table IWB-2500-1 (Examination Categories B-F and B-J), IWC-2420, IWC-2430 and Table IWC-2500-1 (Examination Categories C-F-1 and C-F-2) for lSI of Class 1 or 2 piping and IWB-2200 and IWC-2200 for preservice inspection of Class 1 or 2 piping, or as additional requirements for Class 3 piping or Non-Class piping, for plants issued an initial operating license prior to December 31, 2000. The CC N-716 requirements are expected to reduce the number of inspections required but may also define additional requirements for Class 3 piping or non-Class piping.

Code Case N-716 has not been endorsed for generic use by the NRC. The licensee's relief request refers to the methodology described in CC N-716 instead of describing the details of the methodology in the relief request. The licensee has, however, modified the methodology described in CC N-716 while developing its proposed RIS_B program. When the methodology used by the licensee is accurately described in CC N-716, this safety evaluation (SE) refers to the details found in CC N-716. When the methodology used by the licensee deviates or expands upon the methodology described in CC N-716, this SE refers to the licensee's submittals cited above.

Enclosure

- 2 Therefore, CC N-716 is incorporated in this SE only as a source for some of the detailed methodology descriptions as needed and the NRC staff is not endorsing the use of CC N-716.

2.0 REGULATORY EVALUATION

Pursuant to Title 10 of the Code of Federal Regulations (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 desjgn, geometry, and materials of construction of the components. Paragraph 10 CFR 50.55a(g) also states that the 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 objective of the lSI program, as described in Section XI of the ASME Code and applicable addenda, is to identify conditions (Le., flaw indications) that are precursors to leaks and ruptures in the pressure boundary of these components that may impact plant safety.

The regulations also require, during the first 10-year lSI interval and during subsequent intervals, that the licensee's lSI program complies 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 limitations and modifications listed therein. VEGP is in the third 10-year lSI interval.

Pursuant to 10 CFR 50.55a(g), a certain percentage of ASME Code Category B-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 all B-F welds and 25 percent of all 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 percent 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 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 licensee has proposed to use an 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. As stated in Section 1.0 of this SE, the provisions of CC N-716 are expected to reduce the number of required examinations but may also define additional requirements for Class 3 piping or non-Class piping. The application states that this proposed program will be substituted for the current program in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety.

The licensee states that CC N-716 is founded in large part on the risk-informed inservice inspection (Rl-ISI) process as described in Electric Power Research Institute (EPRI) Topical Report (TR) -112657, which was previously reviewed and approved by the NRC. The licensee further states that the risk-informed application based upon CC N-716 meets the intent and principles of Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" (Reference 4), and RG 1.178, "An Approach For Plant-Specific Risk-Informed Decisionmaking - Inservice Inspection of Piping" (Reference 5). Regulatory Guide 1.174 provides guidance on the use of probabilistic risk analysis (PRA) findings and risk insights in

- 3 support of licensee requests for changes to a plant's licensing basis. Regulatory Guide 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.

The NRC staff has reviewed and evaluated the licensee's proposed RIS_B program based on guidance and acceptance criteria provided in the following documents:

RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

RG 1.178, "An Approach for Plant-Specific Risk-Informed Decision making for Inservice Inspection of Piping,"

RG 1.200, "An Approach for Determining Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

NRC NUREG-0800, Chapter 3.9.8 Standard Review Plan For the Review of Risk-Informed Inservice Inspection of Piping, EPRI TR-112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure,"

3.0 TECHNICAL EVALUATION

Code Case N-716 is founded, in large part, on the RI-ISI process as described in the EPRI TR 112657, Revision B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure,"

December 1999 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML013470102), which was previously reviewed and approved by the NRC. In general, the licensee simplified the EPRI TR method because it does not evaluate system parts that have been generically identified as high-safety-significant (HSS), and uses plant-specific PRA to evaluate in detail only system parts that cannot be screened out as low-safety significant (LSS).

An acceptable RI-ISI program replaces the number and locations of nondestructive examination (NDE) inspections based on ASME Code,Section XI, requirements with 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, and 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.

- 4 The EPRI TR RI-ISI process includes the following steps which, when successfully applied, satisfy the guidance provided in RGs 1.174 and 1.178.

Scope definition Consequence evaluation Degradation mechanism evaluation Piping segment definition Risk categorization Inspection/NDE selection Risk impact assessment Implementation monitoring and feedback 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 CC N-716 replaces a detailed evaluation of the safety significance of each pipe segment with a generic population of HSS segments, followed by a screening flooding analysis to identify any plant-specific HSS segments. The screening flooding analysis is performed in accordance with the flooding analysis described in Section 4.5.7 of ASME RA-Sb-2005, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum B, to ASME RA-S-2002 (Reference 7) as endorsed by RG 1.200, Revision 1 (Reference 8).

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 process.

3.1 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. Standard Review Plan (SRP) 3.9.8 and Reference 5 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 SNC's evaluation included all piping where ASME Code 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 Code program. RG 1.178, identifies different groupings of plant piping that should be included in a RI-ISI program, and also clarifies that a "full-scope" risk-informed evaluation is acceptable. The scope of the RIS_B program as defined in CC N-716 is consistent with the definition of fUll-scope in RG 1.178. Therefore, the !\\IRC staff concludes that the "full-scope" extent of the piping included in the RIS_B program changes satisfies the SRP and RG gUidelines and is acceptable.

3.2 Consequence Evaluation The methodology described in RG 1.178 and the EPRI TR divide 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 that yield

-5 equivalent results. The consequences are used to determine the safety significance of the segments.

In contrast to the EPRI TR methodology, CC N-716 does not require that the consequence of each segment failure be estimated to determine the sa.fety-significance of piping segments.

Instead, CC 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 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 CC 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, and SRP 3.9.8, and, therefore, the generic results are derived from acceptable analyses.

Section 2(a)(5) in CC 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) that considers both the direct and indirect effects of pressure boundary failure and the different failure modes of the piping. This is consistent with the guidelines for evaluating pipe failures with PRA described in RG 1.178, the EPRI TR, 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 SRP 3.9.8 and is, therefore, acceptable.

3.3 Degradation Mechanism Evaluation The EPRI TR requires a determination of the susceptibility to all degradation mechanisms of every weld within the scope of the proposed program. The degradation mechanisms which should be identified are described in the EPRI TR. This information is used to support the safety significance determination for all segments, to target inspections toward the locations with damage mechanisms in the segments that require inspections, and to provide estimates of weld failure frequencies to support the change in risk calculation. Once a segment is placed in the LSS category, the degradation mechanisms at the welds in that segment are not further used in the development of an EPRI RI-ISI program because inspections are not 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. CC N-716 requires a determination of the susceptibility to all degradation

- 6 mechanisms of all welds assigned to the HSS category. The degradation mechanisms to be considered in the CC N-716 are consistent with those identified in the EPRI TR report which the staff has previously concluded is a sufficiently comprehensive list of the applicable mechanisms except for primary water stress corrosion cracking (PWSCC) at pressurized-water reactors (PWR) units.

Since the issuance of the SE on the EPRI TR, several instances of PWSCC of alloy 82/182 dissimilar metal welds have occurred at PWRs. This has prompted the NRC to send a letter (Reference 6) 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 PWSCC-susceptible butt welds in the reactor coolant pressure boundary of PWRs.

This letter represents a departure from the NRC staff's conclusions about PWSCC in the EPRI TR's SER. The NRC staff is including this information to demonstrate that, as issues arise, modifications to RI-ISI programs may be warranted as required in the NRC approval of the RIS_B program. The nuclear power industry, through the Materials Reliability Program (MRP),

developed guidance for inspection and evaluation of primary system piping butt welds in MRP 139 (Reference 9). The licensee states in Reference 1 that a plant augmented inspection program has been implemented at VEGP to meet the requirements of MRP-139.

Pipe failure frequencies are used in the screening analysis searching for HSS welds described above, and then in the change in risk estimate. In Reference 1, the licensee stated that a review was conducted to further verify that LSS piping was not susceptible to water hammer.

The LSS piping may be susceptible to flow-accelerated corrosion (FAC); however, the examination for FAC is performed per the FAC augmented program. In lieu of conducting 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 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 segments with potential high failure frequency (Le., 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 process fulfills the requirements for identifying locations that should be inspected (Le., identifying plant-specific HSS segments) and developing a bounding estimate for the change in risk respectively.

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 segments is to group welds so that consequence analyses can be done

- 7 for the smaller number of segments instead of for each weld. Sections 2(a)(1) to 2(a)(4) in CC 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 CC N-716 uses a PRA to 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.

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 the EPRI TR 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 CC 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. CC 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. 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 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 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. Case Code N-716 requires that any segment with a total estimated core damage frequency (CDF) greater than 1E-6/year be assigned the HSS category. The licensee augmented this CC N-716 metric on CDF with the requirement to also assign the HSS category to any segment with a total estimated large early relief frequency (LERF) greater than 1E-7/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 1.174.

- 8 In the 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 CC 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 RG 1.174 acceptance criteria.

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 Sections 2(a)(1) through 2(a)(4).

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 program 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 Fussel-Vessleyand 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." Such metrics are subordinate to the change in risk metrics in RG 1.174 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.

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. RegUlatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

states that compliance with the attributes of an NRC-endorsed industry PRA standard (the licensee application used ASME RA-Sb-2005) may be used to demonstrate that a PRA analysis is adequate to support a risk-informed application. Regulatory Guide 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 Reference 2, the licensee states that no significant issues pertaining to an independent contractor's evaluation of VEGP's flooding model were found. Three internal flooding related comments were provided and shown to be properly resolved for this application. The licensee states that the updated internal flooding analysis considers submergence, spray, jet impingement, pipe whip, humidity, condensation, and temperature effects in determining the flooding effects on equipment, thereby meeting ASME Capability Category III for identifying susceptibility of each safety-related structure, system, and component in a flood area to flood-induced failures. The updated VEGP internal flooding analysis did not take credit for any isolation by human actions and flooding scenarios were not grouped or subsumed, therefore the use of the updated VEGP internal flooding PRA is conservative. The NRC staff concurs that the licensee's evaluation adequately demonstrates that the resulting RI-ISI program is consistent with the RI-ISI methodology.

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

- 9 acceptance guidelines in RG 1.174, 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 at VEGP provides confidence that HSS segments have been identified. Sections 2(a)(1) through 2(a)(4) in CC N-716 which identify generic HSS portions of systems were applied to VEGP piping. The licensee's PRA analysis used to fulfill the guideline in Sections 2(a)(5) was performed using a PRA of adequate technical quality based on consistency between the PRA and the applicable characteristics of the NRC-endorsed industry standard ASME RA-Sb-2005. The licensee reviewed the results of its flooding analysis and did not identify any segments that had a CDF greater than 1E-6/year or a LERF greater than 1E-7/year.

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

Case Code N-716 contains no provisions for reducing the number of inspections in the inspection program for high-energy line breaks (HELB) outside containment, implemented in response to VEGP Final Safety Analysis Report Section 6.6 and Technical Specification 5.5.16.

However, CC N-716 does include a provision to increase the number of HELB inspections if the HELB program is inspecting less than 10 percent of the welds in this region. Changes to the HELB program may be made as authorized by EPRI TR-1006937, "Extension of the EPRI Risk Informed lSI Methodology to the Break Exclusion Region Programs," (ADAMS Accession No. ML021790518) or by another process found acceptable by the NRC staff.

Case Code N-716 contains no provisions for changing the FAC augmented program developed in response to NRC Generic Letter 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning."

VEGP's FAC program is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RIS_B program.

The MRP-139 will be used as an augmented inspection program for the inspection and management of PWSCC susceptible dissimilar metal welds and will supplement the RI-ISI program.

The VEGP 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 RIS_B Program.

Section 4 in CC N-716 requires that 10 percent of HSS welds shall be selected for examination.

Sections 4(b)(1) through 4(b)(3) in CC N-716 describe how the inspection locations will be selected. The selection process includes guidance that ensures that inspection locations are distributed physically throughout the HSS piping systems and that all degradation mechanisms will be represented in the selected locations. The guidance provides some flexibility in the distribution of locations to satisfy all the guidelines but the number of inspections must be increased beyond 10 percent, if necessary, to meet the quantitative risk acceptance guidelines in Section 5(b).

- 10 In contrast to the EPRI TR which only changed the types of lSI inspections and the locations of inspections, CC N-716 also discontinues preservice inspection requirements for LSS welds.

These preservice examinations are performed to obtain a baseline inspection using the examination method that will be used for subsequent lSI examinations. Similar to the ASME Code which requires preservice examination of all Class 1 welds, CC N-716 requires preservice examination of all HSS welds. Preservice examinations are performed on ASME Code Class 2 welds that are initially selected for lSI, which is 7.5 percent of Class 2 piping welds. Any Class 2, Class 3, or non-code welds that are selected for inspection in the RIS_B program will be HSS welds. Therefore, preservice examinations will continue to be performed on all welds selected for examination using CC N-716 to obtain a baseline inspection using the examination method that will be used for subsequent lSI examinations.

In addition to the preservice exams as required, repair/replacement activities involving welding or brazing areas and welded joints made for installation of items shall be examined in accordance with the Construction Code identified in the licensee's Repair/Replacement Plan.

The licensee's use of CC N-716 does not affect the examinations required to verify the integrity of welds associated with repairlreplacement activities. Therefore, the NRC staff finds that there is no effect on the change in risk calculations associated with repair activities under the RIS_B program because the examinations required to verify the integrity of repaired or replaced welds are not affected and will continue to be performed.

3.7 Risk Impact Assessment The licensee uses a change in risk estimation process approved by the NRC staff in the EPRI TR. The change in risk assessment in the EPRI TR 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.

CC 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. 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 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 corresponding types of feedwater and steam piping breaks. Conservative CCDP estimates were developed from the PRA for these initiating events. The licensee assumed a conditional containment failure probability of 0.1 to convert CCDP to CLERP unless the pipe break could both cause an initiating event and fail the containment barrier. For these scenarios, the CLERP was assigned the same value as the CCDP. 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. The NRC staff also concludes that the licensee uses generally acceptable values for any required additional failure modes, including the conservative 0.1 conditional containment failure probability for a non-bypassed large dry PWR containment.

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 Reference 8. The licensee stated that its flooding analysis did not identify any high

- 11 consequence segments (lower bound CCDP and CLERP of 'I E-4 and 1E-5, respectively) for LSS Class 2 piping that was being inspected under the ASME Code lSI program. Only segments with locations at which an inspection is being discontinued need to be included in the change in risk calculation so limiting the consequence evaluation to segments that are inspected is acceptable.

Section 5 in CC N-716 requires that any piping that has NDE inspections1 added or removed per CC N-716 be included in the change in risk assessment. The licensee used nominally the upper-bound estimates for CCDP and CLERP. Acceptance criteria provided in Section 5(d) in CC N-716 include limits of 1E-7/year and 1E-8/year for increase in CDF and LERF for each system, and limits of 1E-6/year and 1E-7/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 the EPRI TR and are, therefore, acceptable.

The change in risk evaluation approved in the EPRI TR method 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 the approved EPRI TR method with the exception that the change in risk calculation in CC N-716 includes the risk increase from discontinued inspection in LSS locations. Based on the detailed analysis of every segment required by the EPRI TR, 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 and does not need to be quantified. The NRC staff concludes that the licensee's method described in the submittal is acceptable because the deviation from the approved EPRI TR method expands the scope of the calculated change in risk providing confidence that the less detailed analyses of LSS segments required by CC 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 submittals and noted that most of the results indicate a decrease in risk and that all the estimates satisfy both the system level and the total guidelines. Therefore, the NRC staff finds that any increase in risk is small and acceptable.

3.8 Implementation Monitoring and Feedback The objective of this element of References 4 and 5 is to assess performance of the affected piping systems under the proposed RI-ISI program by implementing monitoring strategies that conform with the assumptions and analysis used in developing the RIS_B program. In Reference 1, the licensee states that upon approval of the RIS_B program, procedures that comply with the guidelines described in CC N-716 will be prepared to implement and monitor the program.

1Code Case 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 the EPRI TR that the surface exams do not appreciably contribute to safety and need not be included in the change in risk quantification and, therefore, exclusion of surface examinations from the change in risk evaluations is acceptable.

- 12 This 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 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 In accordance with CC N-716, LSS welds will be exempt from the volumetric, surface, VT-1 and VT-3 visual examination requirements of Section XI. Ten percent of the HSS welds will be selected for examination as addressed in Section 3.6 of this SE. Section 4 of CC N-716 directs users to Table 1 for the examination requirements of the welds selected for examination. The examination method is based on the postulated degradation for the selected weld. Table 1 of CC N-716 is consistent with the traditional RI-ISI approach for examination methods as approved in EPRI TR-112657. The examination methods are based on an inspection-for-cause philosophy so that when there is a potential for a certain degradation mechanism, the examination method selected would be one that would be able to detect that type of degradation. This is consistent with the guidelines for inspection strategies described in SRP 3.9.8 and is, therefore, acceptable.

4.0 CONCLUSION

Regulatory Guide1.174 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 RI-ISI methodology contains details for developing an acceptable RI-ISI program. CC N-716, modified as described by the licensee in its submittals, describes a methodology similar to the EPRI methodology but with several differences as described above 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 the EPRI TR, is consistent with the guidance provided in RG 1.178, and satisfies the guidelines established in RG 1.174 As set forth above, the NRC staff determines that the proposed alternative provides an acceptable level of quality and safety. The implementation strategy is consistent with the Reference 5 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. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i) the NRC staff authorizes the use of the risk-informed/safety-based inservice inspection (RIS_B) program described in VEGP-ISI-ALT-02 at VEGP Units 1 and 2 for the third 10-year lSI interval.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

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5.0 REFERENCES

1.

Letter, M.J. Ajluni (Southern Company) To U.S. Nuclear Regulatory Commission containing Request to Use ASME Code Case N-716, Vogtle Electric Generating Plant, Units 1 and 2, April 15, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML0910700750)

2.

Letter, M.J. Ajluni (Southern Company) To U.S. Nuclear Regulatory Commission containing Response to Request for Additional Information on Relief Request VEGP-ISI-ALT-02, October 12, 2009 (ADAMS Accession No. ML0928600972)

3.

ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1, ASME Code, New York, New York, April 19, 2006.

4.

RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, November 2002 (ADAMS Accession No. ML02320437).

5.

RG 1.178, An Approach for Plant-Specific Risk-Informed Decision making for Inservice Inspection of Piping, September 2003 (ADAMS Accession No. ML032510128).

6.

Letter from J.E. Dyer, U.S. Nuclear Regulatory Commission to G.C. Park, ASME Code Subcommittee on Nuclear Inservice Inspection, related to codification of inspection requirements of reactor coolant system dissimilar metal welds at pressurized water reactors, dated December 20, 2005 (ADAMS Accession No. ML053480359).

7.

ASME RA-Sb-2005, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum B to ASME RA-S-2002, ASME Code, New York, New York, December 30,2005.

8.

RG 1.200, An Approach for Determining The Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, January 2007 (ADAMS Accession No. ML070240001 )

9.

MRP-139, Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline, Revision 1, December 2008, Electric Power Research Institute, Palo Alto, California.

Principal Contributors: Jigar Patel, NRR Keith M. Hoffman, NRR Date: March 3, 2010

ML100610470 "transmitted by memo dated OFFICE NRRlLPL2-1/PM NRRlLPL2-1/PM NRRlLPL2-1/LA NRRlAPLAlBC NRRlDCI/APLA NRRlLPL2-1/BC NAME PBoyle RMartin SRohrer DHarrison" TLupold" GKulesa DATE 3/3/10 3/3/10 3/3/10 3/02/10 3/02/10 3/3/10