ML12306A088

Request to Utilize an Alternative to the Requirements of 10 CFR 50.55a(g) for Implementation of a Risk-Informed, Safety-Based Inservice Inspection Program Based on ASME Code Case N-716

ML12306A088
Person / Time
Site:	Nine Mile Point
Issue date:	10/25/2012
From:	Swift P Constellation Energy Group, EDF Group, Nine Mile Point
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML12306A088 (45)

Text

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a joint venture of P.O. Box 63 Lycoming, NY 13093 Cofnstellation 64F,%eDF Energy 0 NINE MILE POINT NUCLEAR STATION October 25, 2012 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

Nine Mile Point Nuclear Station Unit No. 2; Docket No. 50-410 Request to Utilize an Alternative to the Requirements of 10 CFR 50.55a(g) for Implementation of a Risk-Informed, Safety-Based Inservice Inspection Program Based on ASME Code Case N-716 In accordance with 10 CFR 50.55a(a)(3)(i), Nine Mile Point Nuclear Station, LLC (NMPNS) requests authorization to implement a risk-informed, safety-based inservice inspection (ISI) program based on American Society of Mechanical Engineers (ASME) Code Case N-716, "Alternative Piping Classification and Examination Requirements,Section XI, Division 1," as documented in the enclosed 10 CFR 50.55a Request Number 21SI-011, Rev. 00. The information provided in the enclosed request demonstrates that the proposed alternative provides an acceptable level of quality and safety.

NMPNS plans to implement the proposed alternative during the third ten-year ISI interval, which began on April 5, 2008 and is scheduled to end on April 4, 2018, and requests NRC approval by October 31, 2013 to facilitate planning for the 2014 refueling outage and for remainder of the third ten-year ISI interval.

This letter contains no new regulatory commitments. Should you have any questions regarding the information in this submittal, please contact John J. Dosa, Director Licensing, at (315) 349-5219.

Very truly yours, Paul M. Swift Manager Engineering Services PMS/DEV

Document Control Desk October 25, 2012 Page 2

Enclosure:

Nine Mile Point Nuclear Station, Unit 2 - Third Inservice Inspection Interval, 10 CFR 50.55a Request Number 21SI-01 1, Rev. 00 cc: Regional Administrator, Region I, NRC Project Manager, NRC Resident Inspector, NRC

ENCLOSURE NINE MILE POINT NUCLEAR STATION, UNIT 2 THIRD INSERVICE INSPECTION INTERVAL 10 CFR 50.55a REQUEST NUMBER 21SI-011, REV. 00 Nine Mile Point Nuclear Station, LLC October 25, 2012

• "" i Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

A. COMPONENT IDENTIFICATION System: Various Class 1 and 2 Systems Class: Quality Groups A, and B (ASME Code Class 1, and 2)

Components Affected: All Class 1 and 2 Piping Welds - Examination Categories B-F, B-J, C-F-1, and C-F-2 B. APPLICABLE CODE REQUIREMENTS Pursuant to 10 CFR 50.55a(g), American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PVC),Section XI, 2004 Edition, No Addenda, Examination Tables IWB-2500-1 and IWC-2500-1, Examination Categories B-F, B-J, C-F-i, C-F-2 must receive inservice inspection during each successive 120-month (ten-year) interval.

The Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-01 requires Intergranular Stress Corrosion Cracking (IGSCC) Category A welds to be examined over the 10-year interval in accordance with the staff positions on schedule, methods, personnel and sample expansion.

The required examinations in each Examination Category shall be completed during each successive inspection interval in accordance with Inspection Program B, Tables IWB-2412-1 and IWC-2412-1 and GL 88-01 guidelines, as modified by BWRVIP-75-A. Table 1 below reflects these requirements.

Table 1 ASME Section XI and GL 88-01 Examination Requirements ASME Examination Types of Welds Examination Methods Percentage Code Category Requirements Class 1 B-F Dissimilar Metal Volumetric and Surface 100% Required Welds or Surface 1 B-J Piping Welds Volumetric and Surface 25% Required or Surface 1 GL-A Resistant Material Volumetric 25% Required 2 C-F-1 Piping Welds Volumetric and Surface 7.5% Required or Surface 2 C-F-2 Piping Welds Volumetric and Surface 7.5% Required or Surface Page 1 of 3

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 C. REASON FOR REQUEST FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), Nine Mile Point Nuclear Station, LLC (NMPNS) requests an alternative to the requirements of the ASME B&PVC, 2004 Edition, No Addenda, of Section XI, Division 1, Tables IWB-2500-1 and IWC-2500-1, Examination Categories B-F, B-J, C-F-1 and C-F-2.

NMPNS also requests an alternative to GL 88-01 staff positions, as modified by BWRVIP-75-A, on schedule, methods, personnel and sample expansion for Examination Category A welds (resistant materials) only.

NMPNS also requests authorization to use ASME Code Case N-716, "Alternative Piping Classification and Examination Requirements, Section Xl, Division 1," for risk-informed / safety-based insights.

D. BASIS FOR RELIEF AND ALTERNATIVE EXAMINATIONS The basis for this request for alternative is to document the application of ASME Code Case N-716 to Class 1 and 2 piping systems at Nine Mile Point Nuclear Station Unit 2 using risk-informed and safety based (RISB) insights.

The objective of the inservice inspection (ISI) program is to identify service-induced degradation that might lead to pipe leaks and ruptures, thereby meeting, in part, the requirements set forth in the General Design Criteria and 10 CFR 50.55a. ISI programs are intended to address all piping locations that are subject to degradation. Incorporating risk insights into ISI programs can focus examinations on the more important locations and reduce personnel exposure, while at the same time maintaining or improving the public health and safety.

Electric Power Research Institute (EPRI) Topical Report (TR) EPRI-TR-1 12657, Revision B-A, "Revised Risk-Informed In-service Inspection Evaluation Procedure" (hereafter referred to as EPRI-TR), was submitted for NRC review by letter dated July 29,1999. The NRC review, documented in a safety evaluation dated October 28, 1999, concluded that the EPRI-TR was acceptable for referencing in licensing applications to the extent specified and under the limitations delineated in the EPRI-TR and the associated NRC safety evaluation.

In addition, the NRC staff concluded that the proposed risk-informed inservice inspection program (RI-ISI) as described in the EPRI-TR is a sound technical approach and will provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a for the proposed alternative to the piping ISI requirements with regard to the number of locations, locations of inspections, and methods of inspection.

EPRI provided support in the development of this submittal.

As stated within the EPRI-TR, no changes to the augmented inspection programs for Flow Accelerated Corrosion (FAC) or Intergranular Stress Corrosion Cracking (IGSCC) GL 88-01 (as modified by BWRVIP-75-A) Categories B through G welds are being made in the proposed RIS_B inspection program. The proposed RISB program will supersede augmented inspection programs for IGSCC resistant Category A welds.

In addition to development of the proposed risk-informed ISI program utilizing the EPRI methodology, NMPNS will convert from implementing ASME Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B,Section XI, Division 1," to the implementation of ASME Code Case N-716, which was approved by ASME on April 19, 2006.

As a result of the above insights, more efficient and technically sound means for selecting and scheduling inservice examinations of piping can be achieved, which will provide an acceptable level of quality and safety as required by 10 CFR 50.55a(a)(3)(i).

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 E. IMPLEMENTATION SCHEDULE In accordance with 10 CFR 50.55a(a)(3)(i), the proposed RIS_B inspection program is an alternative to the ASME Code Section Xl inservice inspection requirements for piping with regard to the number of inspections, locations of inspections, and methods of inspections as summarized in Attachment 1 of this request.

NMPNS proposes to implement the alternative RISB inspection plan and schedule in accordance with ASME Code Case N-716, utilizing the EPRI methodology applied to plant specific ASME Code Class 1, and 2 piping in accordance with the EPRI-TR and Regulatory Guide 1.178.

NMPNS plans to complete the current Third Ten-Year ISI Interval by implementing the ASME Code Case N-716 based RISB program during the Second and Third Inspection Periods. Examinations shall be performed such that the period percentage requirements of ASME Section Xl are met for the current Interval, which began on April 5, 2008 and is scheduled to end on April 4, 2018.

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 the current ASME Section XI pressure testing program.

F. PRECEDENTS NRC Safety Evaluation for Seabrook Station, Unit 1, Relief For Alternative 3AR-1, Use of a Risk-Informed, Safety-Based Inservice Inspection Program, dated June 21, 2012 (ML121320552).

NRC Safety Evaluation for Millstone Power Station, Unit No. 2, Issuance of Relief Request RR 11 Regarding Risk-Informed Inservice Inspection Program, dated March 27, 2012 (ML120800433).

NRC Safety Evaluation for Joseph M. Farley Nuclear Plant, Units 1 and 2, Risk-Informed Safety-Based Inservice Inspection Alternative for Class 1 and Class 2 Piping Welds, dated January 18, 2012 (ML12012A135).

NRC Safety Evaluation for River Bend Station, Unit 1, Relief for Alternative RBS-ISI-013, Use of a Risk-Informed, Safety-Based Inservice Inspection Program, dated June 30, 2010 (ML101730157).

NRC Safety Evaluation for Nine Mile Point Nuclear Station, Unit No. 1, Request for Alternative 1 ISI-003, Request to Use ASME Code Case N-716 Associated with the Fourth 10-Year Inservice Inspection Interval, dated March 15, 2010 (ML100700034).

G. ATTACHMENTS Attachment 1, Summary Submittal (Template), Application of ASME Code Case N-716, Risk-Informed / Safety-Based Inservice Inspection Program H. REFERENCES Refer to Attachment 1.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 ATTACHMENT I

SUMMARY

SUBMITTAL (TEMPLATE)

APPLICATION OF ASME CODE CASE N-716 RISK-INFORMED / SAFETY-BASED INSERVICE INSPECTION PROGRAM REQUEST FOR ALTERNATIVE 21SI-011, Rev. 00 Page 1 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table of Contents I Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 Probabilistic Risk Assessment (PRA) Quality 2 Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section XI 2.2 Augmented Programs 3 Risk-Informed / Safety-Bases Inservice Inspection Process 3.1 Safety Significance Determination 3.2 Failure Potential Assessment 3.3 Element and NDE Selection 3.3.1 Successive and Additional Examinations 3.3.2 Program Relief Requests 3.4 Risk Impact Assessment 3.4.1 Quantitative Analysis 3.4.2 Defense-in-Depth 4 Implementation and Monitoring Program 4.1 Implementation 4.2 Feedback (Monitoring) 5 Proposed Inservice Inspection Program Plan Change 6 References/Documentation 7 List of Acronyms and System Abbreviations Appendix A Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Page 2 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Introduction Nine Mile Point Nuclear Station, Unit 2 (NMP2) is currently in the Third Ten-Year Inservice Inspection (ISI) Interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section XI Code for inspection Program B. Nine Mile Point Nuclear Station, LLC (NMPNS) plans to complete the current (Third) ISI Interval by implementing a Risk-Informed Safety-Based (RISB) Program based on ASME Code Case N-716 during the Second and Third Inspection Periods of the Third Interval. The NMP2 Third Interval began on April 5, 2008 and is scheduled to end on April 4, 2018. The Second Inspection Period (of the Third Interval) began on April 5, 2011.

The ASME Section Xl Code of record for the Third ISI Interval is the 2004 Edition, no Addenda, for Examination Categories B-F, B-J, C-F-i, and C-F-2, and Generic Letter (GL) 88-01 IGSCC resistant Category A Class 1 and 2 piping components. In the Second ISI Inspection Period of the Second Interval, NMPNS implemented a Risk-Informed ISI (RI-ISI) Program based on ASME Code Case N-578. NRC approval to adopt the Code Case N-578 alternative was documented in a letter dated May 31, 2001 (ML011420195). NMPNS requested and received approval to continue using the Code Case N-578 alternative in the current (Third) Interval, as documented in NRC letter dated December 1, 2008 (ML083190494). The delta-risk evaluations for both the approved Code Case N-578 program and the proposed Code Case N-716 program are based on a comparison to a traditional program based on ASME Section XI 1989 Edition requirements. The 1989 Edition of ASME Section Xl was the code of record during development of the initial Code Case N-578 RI-ISI submittal.

The objective of this submittal is to provide the information required to support the NMPNS request to use an alternate RIS_B process for the inservice inspection of Class 1 and 2 piping. The RIS_B process used in this submittal is based upon Code Case N-716, "Alternative Piping Classification and Examination Requirements,Section XI, Division 1," which is founded in large part on the RI-ISI process as described in the Electric Power Research Institute (EPRI) Topical Report (TR) 112657, Revision B-A, "Revised Risk-Informed In-service Inspection Evaluation Procedure" (Reference 1).

1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal 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 2), and RG 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmaking for In-service Inspection of Piping" (Reference 3). Additional information is provided in Section 3.4.2 relative to defense-in-depth.

1.2 Probabilistic Risk Assessment (PRA) Quality The NMP2 PRA (Reference 4) is based on a detailed model of the plant that was originally developed from the NMP2 Individual Plant Examination (IPE) and NMP2 Individual Plant Examination for External Events (IPEEE) projects. The original model was reviewed by the NRC and underwent Boiling Water Reactor Owner's Group (BWROG) certification. NRC reviews of the IPE and IPEEE are documented in the NRC Staff evaluations on the IPE dated August 18, 1994 (TAC No M74437) and the IPEEE dated August 12, 1998 (TAC No M83646). The NRC concluded that the NMP2 process is capable of identifying the most likely severe accidents and no significant impacts on the PRA were identified.

The NMP2 PRA has since been upgraded. It is a Level 2, at-power model that includes both internal and external events. A major upgrade of the internal events portion of the model to meet the guidance of RG 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (Reference Page 3 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 7), as well as the American Society of Mechanical Engineers and American National Standard (ASME/ANS) PRA Standard RA-Sa-2009, was completed in July 2009. A formal, BWROG-sponsored industry peer review of the upgraded internal events model was completed in August 2009. The peer review utilized the process described in Nuclear Energy Institute document NEI 05-04, "Process for Performing PRA Peer Reviews Using the ASME PRA Standard," January 2005, and the ASME/ANS PRA Standard. This review to ASME Capability Category II requirements confirmed that the PRA model met the guidance of RG 1.200, Revision 1, and ASME/ANS RA-Sa-2009. There were 18 findings identified by the peer review team. Appendix A contains a summary of these findings, including the status of the resolution for each finding and the potential impact of each finding on this RIS_B application. In summary, a majority of the findings were related to documentation that has no material impact on the results of this application. Resolution of the peer review findings to date has had a minor impact on the model and its quantitative results. Assessment of the remaining open peer review findings has determined that required model changes would result in minor reductions in model quantification results and, therefore, would have a negligible, if any, impact on the conclusions of this application.

Section 2 of EPRI TR 1021467-A, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs," concludes that quantification of external events will not change the conclusions derived from the RI-ISI process. As a result, there is no need to further consider external events.

The latest revision of the PRA model (Reference 4), which takes into account the NMP2 extended power uprate that was implemented during the spring 2012 refueling outage, was used in the development of the RIS-B evaluation.

Based on the above, NMPNS believes that the current PRA model, used in the RIS_B evaluation, has an acceptable quality to support this application.

2 Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section XI ASME Section Xl, Tables IWB-2500-1 and IWC-2500-1, Examination Categories B-F, B-J, C-F-i, and C-F-2, currently provide the requirements for inservice examination of piping welds, utilizing nondestructive examination (NDE) methods as amended by the application of Code Case N-578.

The alternative RIS_B Program for piping is described in Code Case N-716. The RIS_B Program will be implemented as an alternative for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by providing an acceptable level of quality and safety. Non-related portions of the ASME Section Xl Code will remain unaffected by the proposed RISB program.

2.2 Augmented Programs The impact of the RIS_B application on the various plant augmented inspection programs listed below were considered. This section documents only those plant augmented inspection programs that address common piping within the RISB application scope (e.g.,

Class 1 and 2 piping).

• The original plant augmented inspection program for high-energy line breaks, implemented in accordance with the NMP2 Updated Safety Analysis Report (USAR), was revised in accordance with the risk-informed break exclusion region methodology (RI-Page 4 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 BER) described in EPRI TR 1006937, Revision 0-A, "Extension of EPRI Risk Informed ISI Methodology to Break Exclusion Region Programs." EPRI TR 1006937 was approved by the NRC in 2002. The results of the RI-BER application demonstrated that the volumetric examination requirement for this scope of piping could be reduced from 100% to approximately 12%. As a result, a minimum of 12% of the BER population will be examined during the course of each ten-year interval, which exceeds the 10%

requirement imposed by Code Case N-716. NMP2 was a pilot plant for the Risk-Informed BER application.

The NMP2 augmented inspection program for intergranular stress corrosion cracking (IGSCC) per GL 88-01, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping," as modified by BWRVIP-75-A (Reference 15), is relied upon to manage this damage mechanism. GL 88-01 specifies the examination extent and frequency requirements for austenitic stainless steel welds classified as Categories A through G, depending on their susceptibility to IGSCC. In accordance with EPRI TR 112657, piping welds identified as Category A are considered resistant to IGSCC and are assigned a low failure potential provided no other damage mechanisms are present. Consequently, the weld examinations identified as CategoryA inspection locations are subsumed by the RIS_B Program. The existing NMP2 augmented inspection program for the other piping welds susceptible to IGSCC (Categories D and E) remains unaffected by the RISB Program submittal.

• The plant augmented inspection program for flow-accelerated corrosion (FAC) per Generic Letter (GL) 89-08, "Erosion/Corrosion-Induced Pipe Wall Thinning," is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RIS_B Program.

3 Risk-Informed / Safety-Based Inservice Inspection Process The process used for the development of the RISB program conformed to the methodology described in Code Case N-716. The process applied involves the following steps:

• Safety Significance Determination

• Failure Potential Assessment

• Element and NDE Selection

• Risk Impact Assessment

• Implementation Program

• Feedback Loop 3.1 Safety Sigqnificance Determination The systems assessed in the RISB Program are provided in Table 3.1. Piping and instrumentation diagrams and additional plant information, including the existing plant ISI Program, were used to define the system boundaries.

Per Code Case N-716 requirements, piping welds are assigned safety-significance categories, which are used to determine the treatment requirements. High safety-significant (HSS) welds are determined in accordance with the requirements below.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 (1) Class I portions of the reactor coolant pressure boundary (RCPB), except as provided in 10 CFR 50.55a(c)(2)(i) and (c)(2)(ii) of 10CFR50.55a.

(2) Applicable portions of the shutdown cooling pressure boundary function; i.e., Class 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flow path either:

a. As part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or

b. Other systems or portions of systems from the RPV to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds.

(3) That portion of the Class 2 feedwater system (greater than 4 inch nominal pipe size (NPS)) of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve. This does not apply to NMP2, which is a boiling water reactor (BWR).

(4) Piping within the BER (greater than 4 inch NPS) for high-energy piping systems as defined by the owner. This may include Class 3 or non-class piping. As discussed in Section 2.2, NMP2 has a plant specific BER Program.

(5) Any piping segment, including segments subsumed into internal event initiating events, whose contribution to core damage frequency (CDF) is greater than 1 E-06 (or 1E-07 for large early release frequency (LERF)) based upon a plant-specific PRA of pressure boundary failures (e.g., pipe whip, jet impingement, spray, inventory losses). This may include Class 3 or non-class piping.

Low Safety Significance (LSS) is applied to all remaining Class 2, 3 and non-class piping welds that are not determined to be HSS based on the above criteria.

3.2 Failure Potential Assessment Failure potential estimates were generated utilizing industry failure history, plant specific failure history, and other relevant information. These failure estimates were determined using the guidance provided in EPRI TR 112657 (i.e., the EPRI traditional RI-ISI methodology).

Table 3.2 summarizes the failure potential assessment by system for each degradation mechanism that was identified as potentially operative.

3.3 Element and NDE Selection Code Case N-716 and lessons learned from the RISB pilot applications provide criteria for identifying the number and location of required examinations. Ten percent of the HSS welds shall be selected for examination as follows:

(1) Examinations shall be prorated equally among systems to the extent practical, and each system shall individually meet the following requirements:

a. A minimum of 25% of the population identified as susceptible to each degradation mechanism and degradation mechanism combination shall be selected.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00

b. If the examinations selected above exceed 10% of the total number of HSS welds, the examinations may be reduced by prorating among each degradation mechanism and degradation mechanism combination, to the extent practical, such that at least 10% of the HSS population is inspected.

c. If the examinations selected above are not at least 10% of the HSS weld population, additional welds shall be selected so that the total number selected for examination is at least 10%.

(2) At least 10% of the RCPB welds shall be selected.

(3) For the RCPB, at least two-thirds of the examinations shall be located between the first isolation valve (i.e., the isolation valve closest to the RPV) and the RPV.

(4) A minimum of 10% of the welds in that portion of the RCPB that lies outside of containment (e.g., portions of the main feedwater system in BWRs) shall be selected.

(5) A minimum of 10% of the welds within the BER shall be selected.

In contrast to a number of RI-ISI Program applications where the percentage of Class 1 piping locations selected for examination has fallen substantially below 10%, this application results in selection of greater than 10% of the Class 1 welds. A brief summary is provided below, and the results of the selections are presented in Table 3.3. Section 4 of EPRI TR 112657 was used as guidance in determining the examination requirements for these locations.

Class 1 Welds( 1 Class 2 Welds(2) ( Class 3 and All Piping Non-class Welds(3) Welds(4)

Total I Selected Total Selected Total I Selected Total Selected 997 109 1390 1 6 0 2393 110 Notes (1) Includes all Category B-F and B-J locations. All 997 Class 1 piping weld locations are HSS.

(2) Includes all Category C-F-1 and C-F-2 locations. Of the 1390 Class 2 piping weld locations, 1383 weld locations are LSS and the remaining 7 are HSS (BER) welds in the ICS and MSS systems (system abbreviations are defined in Section 7).

(3) There are 4 Class 3 BER welds in the WCS system and 2 non-class BER welds in the FWS system.

(4) Regardless of safety significance, Class 1, 2 and 3 piping components will continue to be pressure tested as required by the ASME Code,Section XI. VT-2 visual examinations are scheduled in accordance with the pressure test program, which remains unaffected by the RISB Program.

3.3.1 Successive and Additional Examinations RISB examinations will be performed to the requirements specified within Table 1 of Code Case N-716. The RISB program will determine, through an engineering evaluation, the root cause of any unacceptable flaw or relevant condition (exceeding the Code Case N-716 Table 1 acceptance standards) determined to be service-related (e.g., fatigue, wall loss, IGSCC, etc.) found during examination. The flaw evaluation, performed in accordance with ASME Section XI, IWB-3600, will account Page 7 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-O11 Rev. 00 for responsible service conditions and degradation mechanisms to determine whether the element(s) will still perform their intended safety function during subsequent operation. If the flaw is acceptable for continued service, successive examinations will be scheduled per Section 6 of Code Case N-716. Elements not meeting this requirement will be repaired, replaced, or analyzed in accordance with the applicable ASME Code Edition and Addenda as identified in the IS[ Program including approved Code Cases and alternatives.

The need for extensive root cause analysis beyond that required for IWB-3600 evaluation will be dependent on practical considerations (i.e. the practicality of performing additional NDE or removing the flaw for further evaluation during the outage). ASME Section Xl, IWB-3134(b) and IWB-3144(b) require submission of the analytical evaluation to the NRC. In addition, the evaluation will be documented in the Corrective Action Program and the Owner submittals required by ASME Section Xl or approved alternatives.

The evaluation will include whether other elements on the segment or additional segments are subject to the same root cause and degradation mechanism.

Additional examinations will be performed, to the requirements of Section 6 of Code Case N-716, on these elements up to a number equivalent to the number of elements requiring examinations on the segment or segments initially examined during the current outage. If unacceptable flaws are determined to be service related or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined during the current outage. No additional examinations will be performed if there are no additional elements identified as being susceptible to the same service related root cause conditions or degradation mechanism.

3.3.2 Program Relief Requests Consistent with previously approved RI-ISI submittals, NMPNS will calculate coverage and use additional examinations or techniques in the same manner as for traditional Section XI examinations and previous RI-ISI examinations. Experience has shown this process to be weld-specific (e.g., joint configuration). As such, the effect on risk, if any, will not be known until that time. In instances where a location may be found that does not meet >90 percent coverage (limited examination), an evaluation will be performed to ensure that the impact of the limited examination is acceptable as required by Footnote 3 to Table 1 of Code Case N-716. This evaluation will be completed as part of the periodic program update required by Section 7 of Code Case N-716.

A relief request will be submitted for all limited examinations per the guidance of 10 CFR 50.55a(g)(5)(iv) within one (1) year after the end of the interval.

Request for Alternative 21SI-007 pertaining to the application of Code Case N-578 will be withdrawn for use at NMP2 upon NRC approval of this RISB Program submittal.

3.4 Risk Impact Assessment The RISB Program has been conducted in accordance with RG 1.174 and the requirements of Code Case N-716, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 This evaluation categorized segments as high safety significant or low safety significant in accordance with Code Case N-716, and then determined what inspection changes are proposed for each system. The changes include changing the number and location of inspections and in many cases improving the effectiveness of the inspection to account for the findings of the RISB degradation mechanism assessment. As an example, for locations subject to thermal fatigue, inspection locations have an expanded volume and the examination is focused to enhance the probability of detection (POD) during the inspection process.

3.4.1 Quantitative Analysis Code Case N-716 has adopted the EPRI TR 112657 process for risk impact analyses whereby limits are imposed to ensure that the change in risk of implementing the RIS_B Program meets the requirements of RGs 1.174 and 1.178.

The EPRI criterion requires that the change in CDF and LERF be less than I E-07 and 1E-08 per year per system, respectively.

For LSS welds, the conditional core damage probability (CCDP) and conditional large early release probability (CLERP) values of 1 E-4 and 1E-5, respectively, were conservatively used except for the high pressure core spray (CSH) and reactor core isolation cooling (ICS) systems, where the CCDP for suction piping off the suppression pool had a higher CCDP. The rationale for using these values is that the change-in-risk evaluation process of Code Case N-716 is similar to that of the EPRI RI-ISI methodology. As such, the goal is to determine CCDP and CLERP threshold values. For example, the threshold values between High and Medium consequence categories are 1 E-4 (CCDP) / 1E-5 (CLERP) and between Medium and Low consequence categories are 1 E-6 (CCDP) / I E-7 (CLERP) from the EPRI RI-ISI Risk Matrix. Using these threshold values streamlines the change-in-risk evaluation as well as stabilizes the update process. For example, if a CCDP changes from 1E-5 to 3E-5 due to an update, it will remain below the 1E-4 threshold value and the change-in-risk evaluation would not require updating.

With respect to assigning failure potential for LSS piping, the criteria are defined by Table 3 of Code Case N-716. That is, those locations identified as susceptible to FAC (or another mechanism and also susceptible to water hammer) are assigned a high failure potential. Those locations susceptible to thermal fatigue, erosion-cavitation, corrosion or stress corrosion cracking are assigned a medium failure potential and those locations that are identified as not susceptible to degradation are assigned a low failure potential.

In order to streamline the risk impact assessment, a review was conducted that verified that the LSS piping was not susceptible to water hammer, as documented in Reference 5. LSS piping may be susceptible to FAC; however, the susceptibility evaluation and examination for FAC is governed by the site FAC program. This review was conducted similar to that done for a traditional RI-ISI application. 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 a medium failure potential ("Assume Medium" in Table 3.4) for use in the change-in-risk assessment. Experience with previous industry RI-ISI applications shows this to be conservative.

NMPNS has conducted a risk impact analysis per the requirements of Section 5 of Code Case N-716 that is consistent with the "Simplified Risk Qualification Method" described in Section 3.7 of EPRI TR 112657. The analysis estimates the net change in risk due to the positive and negative influences of adding and removing locations Page 9 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 from the inspection program.

The CCDP and CLERP values used to assess risk impact were estimated based on pipe break location. Based on these estimated values, a corresponding consequence rank was assigned per the requirements of EPRI TR 112657 and upper bound threshold values were used as provided in Table 3.5. Consistent with the EPRI risk-informed methodology, the upper bound for all break locations that fall within the high consequence rank range was based on the highest CCDP value obtained (e.g., Medium loss of coolant accident (LOCA) in the CSH piping for NMP2 bounds large and small LOCA initiating events as well as other medium LOCA events).

The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability. The basic likelihood of PBF for a piping location with no degradation mechanism present is 1E-08 per Code Case N-716. Piping locations identified as medium failure potential have a likelihood of 2E-07 per Code Case N-716. These PBF likelihoods are consistent with References 9 and 14 of EPRI TR 112657. In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to increased POD from application of the RISB approach.

Table 3.4 presents a summary of the change-in-risk (delta risk) for the proposed RISB Program versus ASME Section XI Code program requirements on a "per system" basis. The impact of FAC is not accounted for in Table 3.4 because the FAC degradation mechanism is addressed via the site augmented FAC program.

The RISB Program credits and relies upon this plant augmented inspection program to manage this degradation mechanism. The plant FAC Program will continue to determine where and when examinations shall be performed. Hence, since the number of FAC examination locations remains the same "before" and "after" and no delta exists, there is no need to include the impact of FAC in the performance of the risk impact analysis.

As indicated in Table 3.6, this evaluation, using estimated CCDP/CLERP values, has demonstrated that unacceptable risk impacts will not occur from implementation of the RISB Program, and satisfies the acceptance criteria of RG 1.174 and Code Case N-716.

The Inspection selections for the original ASME Section XI program, the proposed Code Case N-716 program and the difference between those selections, are contained in Table 3.4 under the column headings SXI, RIS_B, and Delta respectively. The risk impact (change-in-risk) analysis included changes made to the original ASME Section XI inspections as a result of implementing Code Case N-716 and the results are displayed in the Delta column as either no change (represented by 0), an increase (represented by a positive number) or a decrease (represented by a negative number). To show that the use of a conservative upper bound CCDP/CLERP does not result in an optimistic calculation with regard to meeting the acceptance criteria (Code Case N-716 Section 5), a conservative sensitivity was conducted where the RISB selections were set equal to the ASME Section XI selections (Delta changed from positive number to zero). The acceptance criteria are met when the number of RISB selections is not allowed to exceed Section Xl.

Page 10 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 3.4.2 Defense-In-Depth The intent of the inspections mandated by ASME Section XI for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system pressure boundary. Currently, the process for picking inspection locations is based upon structural discontinuity and stress analysis results. As referenced in Section 2.3 of EPRI TR 112657 and depicted in the Summary of the ASME White Paper 92-01-01, Revision 1, "Evaluation of In-service Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds" (Reference 8), this method has been ineffective in identifying leaks or failures. EPRI TR 112657 and Code Case N-716 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.

This process has two key independent ingredients: a determination of each location's susceptibility to degradation and an independent assessment of the consequence of the piping failure. These two ingredients assure that defense-in-depth is maintained. First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leaks or ruptures is increased. Second, a generic assessment of high-consequence sites has been determined by Code Case N-716 as supplemented by plant-specific evaluations, thereby requiring a minimum threshold of inspection for important piping whose failure would result in a LOCA or BER break. Finally, Code Case N-716 requires that any piping on a plant-specific basis that has a contribution to CDF of greater than 1E-06 or LERF of greater than 1 E-7 be included in the scope of the application. NMP2 did not identify any such piping (as documented in References 4 and 5).

All locations within the Class 1, 2, and 3 pressure boundaries will continue to be pressure tested in accordance with ASME Section Xl Code requirements, regardless of their safety significance.

4 Implementation and Monitoring Program 4.1 Implementation Upon approval of the proposed RISB Program, appropriate procedures and/or revisions to the existing inspection program that implement the guidelines described in EPRI TR 112657 and/or Code Case N-716 will be completed to implement and monitor the program. The new program will be integrated into the existing and subsequent ASME Section XI inservice inspection intervals. No changes to the Technical Specifications or Updated Safety Analysis Report are necessary for the alternative RIS_B Program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as implementation of the Code Case N-716 prescribed examination methods, details of the Code Case N-716 prescribed acceptance standards, pressure testing, corrective measures, documentation requirements, reporting requirements, and quality control requirements.

Existing ASME Section XI program implementation documents will be retained and modified to address the RISB process.

4.2 Feedback (Monitoring)

The RISB Program is a living program that is required to be monitored periodically for changes that could impact the basis for which welds are selected for examination.

Monitoring encompasses numerous facets, including the review of changes to the plant Page 11 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 configuration, changes to operations that could affect the degradation assessment, a review of NMP2 NDE results, a review of site failure information from the NMPNS corrective action program, and a review of industry failure information from industry operating experience.

Also included is a review of PRA changes for their impact on the RISB program. These reviews provide a feedback loop such that new relevant information is obtained that will ensure that the appropriate locations selected for examination are maintained. As a minimum, this review will be conducted on an ASME period basis. In addition, more frequent adjustment may be required as identified by NRC Bulletins or Generic Letters, or by industry and plant-specific feedback. Periodic updates will meet the guideline recommendations contained within Nuclear Energy Institute (NEI) 04-05, "Living Program Guidance to Maintain Risk-Informed Inservice Inspection Programs for Nuclear Plant Piping Systems." Changes will be reflected, as appropriate, in the future 10-Year inspection plan and schedule submittals as required by ASME Section XI, IWA-1400(c).

If a flaw or relevant condition is detected during examination, this adverse condition will be addressed by the corrective action program and procedures, and the ISI Program Plan. The following are appropriate actions to be taken:

(1) Identify - Examination results conclude there is an unacceptable flaw.

(2) Characterize - Determine if regulatory reporting is required and assess if an immediate safety or operation impact exists.

(3) Evaluate - Determine the cause and extent of the condition identified and develop a corrective action plan or plans.

(4) Decide - Make a decision to implement the corrective action plan.

(5) Implement - Complete the work necessary to correct the problem and prevent recurrence.

(6) Monitor - Ensure that the RISB program has been updated based on the completed corrective action.

(7) Trend - Identify conditions that are significant based on accumulation of similar issues.

For preservice examinations, NMPNS will follow the rules contained in Section 3.0 of Code Case N-716. Welds classified as HSS require a preservice inspection. The examination volumes, techniques, and procedures shall be in accordance with Table 1 of Code Case N-716. Welds classified as LSS do not require preservice inspection.

5 Proposed Inservice Inspection Program Plan Change A comparison between the RISB Program and ASME Section XI inspection program requirements for in-scope piping is provided in Table 5.1.

NMP2 is currently in the Second Period of the Third Ten-Year ISI Interval. NMPNS plans to complete the current (Third) ISI Interval by implementing a Code Case N-716 based Risk-Informed Safety-Based Program during the Second Inspection Period of the Third Interval. The NMP2 Third Interval began on April 5, 2008. The Second Inspection Period (of the Third Interval) began on April 5, 2011 and includes the 2012 and 2014 refueling outages (RFO13 and RF014). NMP2 has completed the First Period examinations as defined in the current ISI Program Plan including the approved Code Case N-578 alternatives satisfying ASME Section XI percentage requirements. In anticipation of implementation of the Code Case N-716 RISB program, weld exams have been rescheduled within the current Second ISI Period exam schedule (RFO13 and RFO14). Upon approval of this RIS_B submittal, NMPNS will remove the exams (moved from RFO13 to RFO14) from the RFO14 schedule to make the Second Inspection Period consistent with the proposed Code Case N-716 RISB exam schedule. Examinations shall be performed such that the period percentage requirements of ASME Section XI are met for the current Interval.

Page 12 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 As discussed in previous sections, implementation of the RISB program will not alter the augmented examination requirements for FAC, GL 88-01 (IGSCC) welds, or RI-BER welds in high-energy piping.

6 ReferenceslDocumentation

1. Electric Power Research Institute Topical Report 112657, Revised Risk-Informed In-service Inspection Evaluation Procedure, Revision B-A, dated December 1999

2. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis

3. Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking for Inservice Inspection of Piping

4. NMP2 PRA (Model 10U2Model) 09152011 EPU

5. CNG-NMP2-ISI-003-RI-001, ASME Code Case N-716 Evaluation - Nine Mile Point Unit 2 dated January 2012, Revision 00, prepared by J.H. Moody Consulting, Inc.

6. ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI, Division 1

7. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 1, January 2007

8. Summary of the ASME White Paper 92-01-01, Revision 1, Evaluation of In-service Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds, dated July 1995

9. NER-2A-025, NMP2 RI-ISI BER Evaluation

10. Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, January 25, 1988

11. NUREG-0313, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, Revision 2, January 1988

12. Generic Letter 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, dated May 2, 1989

13. Electric Power Research Institute Topical Report 1021467-A, Nondestructive Evaluation:

Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs, June 2012

14. Nuclear Energy Institute (NEI) 04-05, Living Program Guidance to Maintain Risk-Informed Inservice Inspection Programs for Nuclear Plant Piping Systems, April 2004

15. BWRVIP-75-A, BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules, Final Report, October 2005 (EPRI Report 1012621)

Page 13 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 7 List of Acronyms and System Abbreviations Acronyms BER Break Exclusion Region BWROG Boiling Water Reactor Owners Group BWRVIP Boiling Water Reactor Vessel and Internals Project cc Crevice Corrosion CCDP Conditional Core Damage Probability CDF Core Damage Frequency CLERP Conditional Large Early Release Probability DM Degradation Mechanism ECSCC External Chloride Stress Corrosion Cracking E-C Erosion-Cavitation FAC Flow-Accelerated Corrosion HELBCUU High Energy Line Break - Cleanup System HSS High Safety Significant IGSCC Intergranular Stress Corrosion Cracking ILOCA-OC Isolable Loss of Coolant Accident - Outside Containment IPE Individual Plant Examination IPEEE Individual Plant Examination External Events ISI Inservice Inspection LERF Large Early Release Frequency LOCA Loss of Coolant Accident LSS Low Safety Significant MIC Microbiologically-Influenced Corrosion MLOCAHS Medium Loss of Coolant Accident - CSH System NDE Non-destructive Examination NNS Non-nuclear Safety NPS Nominal Pipe Size PBF Pressure Boundary Failure PIT Pitting PLOCA Potential Loss of Coolant Accident PLOCA-OC Potential Loss of Coolant Accident - Outside Containment POD Probability of Detection PRA Probabilistic Risk Assessment PWSCC Primary Water Stress Corrosion Cracking RCPB Reactor Coolant Pressure Boundary RCPB (IFIV) Reactor Coolant Pressure Boundary Inside First Isolation Valve RCPB (OC) Reactor Coolant Pressure Boundary Outside Containment RI-ISI Risk-Informed Inservice Inspection RIS B Risk-Informed / Safety-Based Inservice Inspection RIS-BER Risk-informed Break Exclusion Region SDC Shutdown Cooling SP Suction Piping RPV Reactor Pressure Vessel SXl ASME Section Xl TASCS Thermal Stratification, Cycling, and Striping TGSCC Transgranular Stress Corrosion Cracking TT Thermal Transients Page 14 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 System Abbreviations ASS Auxiliary Steam CSH High Pressure Core Spray CSL Low Pressure Core Spray DER Drywell Equipment Drains FWS Feedwater ICS Reactor Core Isolation Cooling ISC Nuclear Boiler and Process Instrumentation MSS Main Steam RCS Reactor Recirculation RDS Control Rod Drive (CRD) Scram Discharge Volume RHS Residual Heat Removal RPV Reactor Pressure Vessel SLS Standby Liquid Control WCS Reactor Water Cleanup Page 15 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 3.1 Code Case N-716 Safety Significance Determination Safety t m()

Sy System0) ount CWeld N-716 Safety Significance Determination Significance RCPB SDC PWR: FW BER CDF> I E-6(2 ) High Low ASS 4 1 21 1 CSH 164 1 19 1 1 CSL 117 "

DER 2 1 1 72 " I FWS 27 1 " "

2 " 1 65 " "

5 1" 1" 1 5

ICS 3 _" #

207 "

ISC 19 " "

209 1 I 44 V" 1 MSS1 v -

4 90 "

RCS 106 1 1 22 # #

RDS 76 "

86 1 /

RHS 78 1 1 I 725 RPV 34 1 1 SLS 50 1 1 112 1 1 WCS 46 1 "

_ _ _ 4 " 1" 797 1 1

SUMMARY

122 V" "

RESULTS 78 FOR ALL SYSTEMS 13 ___

1383 1 TOTALS 2393 1010 1383 Notes:

(1) System abbreviations are defined in Section 7.

(2) Piping is also evaluated for impact on LERF > 1 E-7.

Page 16 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 3.2 Failure Potential Assessment Summary Thermal Localized Flow Thermal ~Stress Corrosion Cracking LclzdFo Fatigue Stress Corrosion Sensitive System(1 ) TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC ASS"3 )

CSH(3) v 3

CSL( )

DER FWS (2)

ICS(3)

ISC _

MSS(3) "

RCS RDS(3 )

RHS(3 ) ____

RPV "

SLS (2)

WCS Notes:

(1) System abbreviations are defined in Section 7.

(2) The FAC Program has previously identified areas for inspection in these systems, but this has no impact on this application.

(3) A degradation mechanism assessment was not performed on low safety significant piping segments. This includes the ASS in its entirety, as well as portions of the CSH, CSL, ICS, MSS, RDS, and RHS systems.

Page 17 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 3.3 Code Case N-716 Element Selections Weld Count N-716 Selection Considerations System(I) HSS LSS DIs RCPB RCPB RCPB BER Selections (IFIV) (OC)

ASS 4 None 0 CSH 11 TASCS , / 2 CSH 8 None " 0 CSH 2 None V / 1 CSH 164 None 0 CSL 8 TASCS , V 2 CSL 9 None , 0 CSL 2 None , , 1 CSL 117 None 0 DER 2 None , 1 FWS 25 TASCS " " 3 FWS 6 TASCS V / V 5 FWS 3 TASCS / V 0 FWS 12 TASCS / $ V 2 FWS 47 None " / 0 FWS 6 None / , , 1 FWS 2 None V 0 ICS 9 TASCS V " 3 ICS 4 TASCS / 1 ICS 14 None " / 1 ICS 30 None /" 0 ICS 2 None V , ' 1 ICS 2 None V V 0 ICS 8 None V 1 ICS 1 None , / $ 1 ICS 3 None " 0 ICS 207 None 0 ISC 6 TASCS , / 2 ISC 13 None " / 0 MSS 10 TASCS " " 3 MSS 181 None V , 6 MSS 18 None / / V 9 MSS 8 None V 0 MSS 6 None V V 1 MSS 10 None , V 0 MSS 20 None V V V 7 MSS 4 None V 0 MSS 90 None 0 RCS 1 IGSCC V V 1 RCS 105 None V V 10 RDS 2 None V V 1 RDS 76 None 0 RHS 22 TASCS , V 6 RHS 4 EC V V 1 RHS 46 None V V 6 Page 18 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 3.3 Code Case N-716 Element Selections Weld Count N-716 Selection Considerations System(I) HSS LSS D~s RCPB RCPB RCPB BER Selections (IFIV) (OC)

RHS 68 None / 0 RHS 24 None / 4 RHS 725 None 0 RPV 30 IGSCC , V 4 RPV 4 None V V 0 SLS 10 TASCS V V 4 SLS 26 None V 0 SLS 14 None V V 2 WCS 8 TASCS,IGSCC V V 5 WCS 10 IGSCC , V 3 WCS 11 TASCS V V 2 WCS 29 TASCS V V V 5 WCS 75 None V V 1 WCS 1 None V V V 1 WCS 8 None V 0 WCS 2 None V V 0 WCS 14 None V V V 0 WCS 4 None , 0 112 TASCS V V 27 6 TASCS V V V 5 3 TASCS V V 0 41 TASCS V V V 7 4 TASCS V 1 8 TASCS,IGSCC V V 5 Summary 41 IGSCC V V 8 Results 4 EC V V 1 All 487 None I/ V" 25 Systems 21 None V V V 11 159 None V 1 10 None V V 1 60 None V V 9 41 None V V V 9 13 None _ _ 0 1383 0 Totals 1010 1383 110 Note:

(1) System abbreviations are defined in Section 7.

Page 19 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 3.4 Risk Impact Analysis Results Safety Break Failure Potential(3) Inspections CDF Impact LERF Impact ysem Significance Location DMs Rank SXI()2 RIS B(4) Delta w/POD wlo POD w/POD w/o POD ASS Total Low Class 2 LSS Assume Medium 1 0 -1 1.OOE-1 1 1.OOE-1 1 1.OOE-12 1.OOE-12 CSH High LOCA TASCS Medium 3 2 -1 -5.40E-11 3.OOE-11 -5.40E-12 3.OOE-12 CSH High PLOCA None Low 1 0 -1 5.00E-13 5.OOE-13 5.OOE-14 5.00E-14 CSH High PLOCA-OC None Low 1 1 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 CSH Low Class 2 SP Assume Medium 13 0 -13 2.60E-08 2.60E-08 2.60E-09 2.60E-09 CSH Total 2.59E-08 2.60E-08 2.59E-09 2.60E-09 CSL High LOCA TASCS Medium 3 2 -1 -5.40E-11 3.OOE-11 -5.40E-12 3.OOE-12 CSL High PLOCA None Low 1 0 -1 5.OOE-13 5.OOE-13 5.OOE-14 5.OOE-14 CSL High PLOCA-OC None Low 1 1 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 CSL Low Class 2 LSS Assume Medium 10 0 -10 1.O0E-10 1.00E-10 1.OE-11 1.OE-11 CSL Total 4.65E-11 1.31E-10 4.65E-12 1.31E-11 DER High PLOCA None Low 0 1 1 -5.00E-13 -5.OOE-13 -5.OOE-14 -5.OOE-14 DER Total -5.OOE-13 -5.OOE-13 -5.OOE-14 -5.OOE-14 FW High LOCA TASCS Medium 10 8 -2 -2.52E-10 6.OOE-11 -2.52E-11 6.00E-12 FW High PLOCA TASCS Medium 2 0 -2 1.20E-11 2.00E-11 1.20E-12 2.00E-12 FW High PLOCA-OC TASCS Medium 4 2 -2 -3.60E-1 1 6.OOE-1 1 -6.OOE-1 2 1.OOE-1 1 FW High LOCA None Low 2 0 -2 3.OOE-12 3.OOE-12 3.OOE-13 3.OOE-13 FW High PLOCA-OC None Low 4 1 -3 4.50E-12 4.50E-12 7.50E-13 7.50E-13 FW Total -2.69E-10 1.48E-10 -2.90E-11 1.91 E-11 ICS High LOCA TASCS Medium 3 3 0 -1.08E-10 0.00E+00 -1.08E-11 0.00E+00 ICS High PLOCA TASCS Medium 0 1 1 -1.80E-11 -1.O0E-11 -1.80E-12 -1.OOE-12 ICS High LOCA None Low 4 2 -2 3.00E-12 3.00E-12 3.OOE-13 3.OOE-13 ICS High PLOCA None Low 2 0 -2 1.OOE-12 1.00E-12 1.OOE-13 1.OOE-13 ICS High PLOCA-OC None Low 3 2 -1 1.50E-12 1.50E-12 2.50E-13 2.50E-13 ICS Low Class 2 SP Assume Medium 12 0 -12 2.40E-08 2.40E-08 2.40E-09 2.40E-09 ICS Total 2.39E-08 2.40E-08 2.39E-09 2.40E-09 ISC High LOCA TASCS Medium 0 2 2 -1.08E-10 -6.OOE-11 -1.08E-11 -6.OOE-12 Page 20 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 3.4 Risk Impact Analysis Results st" Safety Break Failure Potential(3 ) Inspections CDF Impact LERF Impact ysem Significance Location DMs Rank SXI( 2 ) RIS_B(4) Delta w/POD w/o POD w/POD w/o POD ISC High LOCA None Low 0 0 0 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 ISC Total -1.08E-10 -6.OOE-11 -1.08E-11 -6.OOE-12 MSS High LOCA TASCS Medium 0 3 3 -1.62E-10 -9.00E-11 -1.62E-11 -9.OOE-12 MSS High LOCA None Low 46 13 -33 4.95E-11 4.95E-11 4.95E-12 4.95E-12 MSS High PLOCA None Low 7 1 -6 3.00E-12 3.00E-12 3.00E-13 3.00E-13 MSS High PLOCA-OC None Low 21 7 -14 2.10E-11 2.10E-11 3.50E-12 3.50E-12 MSS Low Class 2 LSS Assume Medium 5 0 -5 5.00E-1 1 5.00E-1 1 5.OOE-12 5.00E-12 MS Total -3.85E-11 3.35E-11 -2.45E-12 4.75E-12 RCS High LOCA IGSCC Medium 1 1 0 0.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 RCS High LOCA None Low 26 10 -16 2.40E-11 2.40E-11 2.40E-12 2.40E-12 RCS Total 2.40E-11 2.40E-11 2.40E-12 2.40E-12 RDS High LOCA None Low 1 1 0 0.00E+00 0.00E+00 0.00E+00 0.00E+00 RDS Low Class 2 LSS Assume Medium 6 0 -6 6.OOE-1 1 6.OOE-1 1 6.OOE-12 6.OOE-12 RDS Total 6.OOE-11 6.OOE-11 6.OOE-12 6.OOE-12 RHR High LOCA TASCS Medium 9 6 -3 -1.62E-10 9.00E-11 -1.62E-11 9.00E-12 RHR High PLOCA-OC E-C Medium 2 1 -1 3.00E-11 3.00E-11 5.00E-12 5.00E-12 RHR High LOCA None Low 15 6 -9 1.35E-11 1.35E-11 1.35E-12 1.35E-12 RHR High PLOCA None Low 6 0 -6 3.00E-12 3.OOE-12 3.OOE-13 3.OOE-13 RHR High PLOCA-OC None Low 4 4 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 RHR Low Class 2 LSS Assume Medium 62 0 -62 6.20E-10 6.20E-10 6.20E-11 6.20E-11 RHR Total 5.05E-10 7.57E-10 5.25E-11 7.77E-11 RPV High LOCA IGSCC Medium 30 4 -26 3.90E-11 3.90E-11 3.90E-12 3.90E-12 RPV High LOCA None Low 2 0 -2 3.OOE-12 3.OOE-12 3.OOE-13 3.OOE-13 RPV Total 4.20E-11 4.20E-11 4.20E-12 4.20E-12 SLS High LOCA TASCS Medium 0 4 4 -2.16E-10 -1.20E-10 -2.16E-11 -1.20E-11 SLS High PLOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 SLS High PLOCA-OC None Low 0 2 2 -3.OOE-12 -3.OOE-12 -5.OOE-13 -5.OOE-13 Page 21 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 3.4 Risk Impact Analysis Results System() Safety Break Failure Potential(3) Inspections CDF Impact LERF Impact 2

P Significance Location DMs Rank SXIM ) RIS B(4 ) Delta wIPOD w/o POD w/POD w/o POD SLS Total -2.19E-10 -1.23E-10 -2.21 E-1 1 -1.25E-1 I WCS High LOCA TASCS,IGSCC Medium 0 5 5 -1.50E-10 -1.50E-10 -2.50E-11 -2.50E-11 WCS High LOCA IGSCC Medium 4 3 -1 3.OOE-11 3.00E-11 5.OOE-12 5.OOE-12 WCS High LOCA TASCS Medium 0 2 2 -1.08E-10 -6.00E-11 -1.08E-11 -6.00E-12 WCS High PLOCA-OC TASCS Medium 15 5 -10 0.00E+00 3.00E-10 0.00E+00 5.00E-11 WCS High LOCA None Low 4 2 -2 3.00E-12 3.00E-12 3.00E-13 3.OOE-13 WCS High PLOCA None Low 1 0 -1 5.OOE-13 5.00E-13 5.00E-14 5.OOE-14 WCS High PLOCA-OC None Low 4 0 -4 6.OOE-12 6.00E-12 1.00E-12 1.00E-12 WCS Total -2.19E-10 1.30E-10 -2.95E-1 1 2.54E-1 1 Grand 351 108 -243 4.97E-08 5.12E-08 4.97E-09 5.14E-09 Total __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _

Notes:

(1) System abbreviations are defined in Section 7.

(2) Only those ASME Section XI Code inspection locations that received a volumetric examination in addition to a surface examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-1 12657.

(3) The failure potential rank for high safety significant (HSS) locations is then assigned as "High", "Medium", or "Low" depending upon potential susceptibly to the various types of degradation. Note: Low safety significant (LSS) locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").

(4) Only those RIS_B inspection locations that receive a volumetric examination are included in the count. Inspection locations receiving VT2 exams per Code Case N-716 were not considered.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 3.5 CCDP and CLERP Values Based on Break Location Estimated Consequence Upper!/ Lower Bound Break Location i Description of Affected Piping CCDP CLERP Rank CCDP CLERP LOCA 3E-04 3E-05 (U) 3E-04 (U) 3E-05 The highest CCDP for Medium LOCA in CSH (MLOCAHS) HIGH (L) 1E-04 (L) 1E-05 Unisolable RCPB piping of all sizes was used (0.1 margin used for CLERP) (LE-4 ()_E0 ILOCA( 1 ) 1E-06 1 E-07 Calclate ML1H C P o6 - a(U) E7 1E-04 (U) 1E-05 Piping between 1st and 2nd normally open isolation Calculated based on MLOCAHS CCDP of 3E-4 and valve fail MEDIUM (LIE06 L)1E7 vaeindectimntWCCSMS F )

to close probability of 3E-3 (0.1 margin used for CLERP) (L) 1E-06 (L) 1E-07 valve inside containment (WCS, CS, MSS, FWS)

PLOCA 1E-06 1E-07 Calculated based on MLOCAHS CCDP of 3E-4 and valve MEDIUM (U) 1E-04 (U) 1E-05 Piping between 1st and 2nd normally closed isolation rupture probability of <1 E-3 (0.1 margin used for CLERP) (L) 1E-06 (L) 1E-07 valve inside containment (RHS, CSL, CSH, SLS)

ILOCA-OC(2) 5E-05 5E-05 (U) 3E-04 (U) 5E-05 Piping between penetration and outside containment Isolable LOCA outside containment CCDP based on HIGH isolation valve with normally open isolation valve inside initiating event HELBCUU CCDP of 5E-5 (CCDP = CLERP) (L) 1E04 (L) 1E05 containment (WCS, ICS, MSS, FWS)

PLOCA-OC 1 E-05 1 E-05 Potential LOCA outside containment CCDP based on valve (U) 1E-04 (U) 1E-05 Piping between penetration and outside containment vMEDIUM isolation valve with normally closed isolation valve inside rupture probability <1E-3 and CCDP for ISLOCA <1E-2 (L) 1E-06 (L) 1E-07 containment (RHS, CSL, CSH, SLS)

(CCDP = CLERP)

Class 2 LSS 1E-04 1E-05 (U) 1E-04 (U) 1E-05 All other Class 2 system piping designated as low safety MEDIUM (ignificant except for ICS and CSH suction from Estimated based on upper bound for Medium Consequence (L) 1E-06 (L) 1E-07 siuppression pool Class 2 SP 2E-02 2E-03 CS and CSL suction piping from the suppression pool HIGH (U) 2E-02 (U) 2E-03 Class 2 ICS and CSH suction lines from the suppression although low frequency and low risk has a CCDP -2E-2 and (L) 1E-04 (L) 1E-05 pool designated as low safety significant.

CLERP -2E-3 (0.1 margin used for CLERP) (3)

Notes:

(1) All welds located inside containment and beyond the first isolation valve are designated as PLOCA whether normally closed or normally open auto closed.

(2) All welds located outside containment and beyond the first isolation valve are designated as PLOCA-OC whether normally closed or normally open auto closed. Quantification is conservatively based on ILOCA-OC CCDP and CLERP.

(3) All ICS and CSH Class 2 LSS welds were conservatively assigned the "Class 2 SP" CCDP and CLERP in the risk impact quantification.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011I Rev. 00 Table 3.6 System With POD Credit Without POD Credit Delta CDF Delta LERF Delta CDF Delta LERF ASS - Auxiliary Steam 1.OOE-1 1 1.OOE-12 1.OOE-1 1 1.OOE-12 CSH - High Pressure Core Spray 2.59E-08 2.59E-09 2.60E-08 2.60E-09 CSL - Low Pressure Core Spray 4.65E-11 4.65E-12 1.31 E-10 1.31 E-11 DER - Drywell Equipment Drain -5.OOE-13 -5.OOE-14 -5.OOE-13 -5.OOE-14 FWS - Feedwater -2.69E-10 -2.90E-11 1.48E-10 1.91E-11 ICS - Reactor Core Isolation Cooling 2.39E-08 2.39E-09 2.40E-08 2.40E-09 ISC - Nuclear Boiler and Process Instrument Lines -1.08E-10 -1.08E-11 -6.OOE-11 -6.OOE-12 MSS - Main Steam -3.85E-11 -2.45E-12 3.35E-11 4.75E-12 RCS - Reactor Recirculation 2.40E-11 2.40E-12 2.40E-11 2.40E-12 RDS - Control Rod Drive Scram Discharge Volume 6.OOE-11 6.OOE-12 6.OOE-11 6.OOE-12 RHS - Residual Heat Removal 5.05E-10 5.25E-11 7.57E-10 7.77E-11 RPV - Reactor Pressure Vessel 4.20E-11 4.20E-12 4.20E-11 4.20E-12 SLS - Standby Liquid Control -2.19E-10 -2.21 E-1 1 -1.23E-10 -1.25E-1 1 WCS - Reactor Water Cleanup -2.19E-10 -2.95E-11 1.30E-10 2.54E-11 Total 4.97E-08 4.96E-09 5.12E-08 5.14E-09 Page 24 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 5.1 Insoection Location Selection Comoarisons Between ASME= Section XI and Code Case N-716 System(l) Safety Significance Break Failure Potential Code Weld Section XI Code Case N-716 High Low Location DMs Rank Category Count Volume Surface RISB Other(2)

ASS LSS n/a Assume Medium C-F-2 4 1 0 NA CSH V LOCA TASCS Medium B-J 11 3 2 NA CSH V PLOCA None Low B-J 8 1 0 NA CSH V PLOCA-OC None Low B-J 2 1 1 NA CSH V LSS n/a Assume Medium C-F-i, C-F-2 164 13 1 0 NA CSL " LOCA TASCS Medium B-J 8 3 2 NA CSL V" PLOCA None Low B-J 9 1 0 NA CSL V PLOCA-OC None Low B-J 2 1 1 NA CSL V LSS n/a Assume Medium C-F-i, C-F-2 117 10 0 NA DER V PLOCA None Low B-J 2 0 1 NA FW V LOCA TASCS Medium B-J 31 10 8 NA FW V, PLOCA TASCS Medium B-J 3 2 0 NA FW V PLOCA-OC TASCS Medium B-J 12 4 2 NA FW V LOCA None Low B-J 47 2 0 NA FW " PLOCA-OC None Low B-J, CL4 8 4 1 NA ICS " LOCA TASCS Medium B-J 9 3 3 NA ICS V, PLOCA TASCS Medium B-J 4 0 1 NA ICS V LOCA None Low B-J 16 4 2 NA ICS V PLOCA None Low B-J 32 2 0 NA ICS V" PLOCA-OC None Low B-J, C-F-2 12 3 1 2 NA ICS V LSS n/a Assume Medium C-F-i, C-F-2 207 12 0 NA ISC V LOCA TASCS Medium B-J 6 0 1 2 NA ISC V" LOCA None Low B-F, B-J 13 0 11 0 NA MSS V LOCA TASCS Medium B-J 10 0 2 3 NA MSS V LOCA None Low B-J 199 46 15 13 2 VT-2 MSS V PLOCA None Low B-J 14 7 1 NA MSS V PLOCA-OC None Low B-J, C-F-2 34 21 7 NA MSS V LSS n/a Assume Medium C-F-2 90 5 0 NA RCS V LOCA IGSCC Medium B-J 1 1 1 NA RCS V LOCA None Low B-J 105 26 10 NA Page 25 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 5.1 Inspection Location Selection Comparisons Between ASME Section XI and Code Case N-716 System(l) Safety Significance Break Failure Potential Code Weld Section XI Code Case N-716 High Low Location DMs Rank Category Count Volume Surface RISB Other(2)

RDS LOCA None Low B-J 2 1 1 NA RDS LSS n/a Assume Medium C-F-2 76 6 0 NA RHR V" LOCA TASCS Medium B-J 22 9 6 NA RHR __ _ PLOCA-OC E-C Medium B-J 4 2 1 NA RHR _ _ LOCA None Low B-J 46 15 6 NA RHR " PLOCA None Low B-J 68 6 1 0 NA RHR __ _ PLOCA-OC None Low B-J 24 4 4 NA RHR " LSS n/a Assume Medium C-F-i, C-F-2 725 62 3 0 NA RPV " LOCA IGSCC Medium B-F 30 30 4 NA RPV LOCA None Low B-F, B-J 4 2 1 0 NA SLS LOCA TASCS Medium B-J 10 0 4 4 NA SLS PLOCA None Low B-J 26 0 1 0 NA SLS PLOCA-OC None Low B-J 14 0 1 2 NA WCS LOCA TASCS,IGSCC Medium B-J 8 0 5 NA WCS I LOCA IGSCC Medium B-J 10 4 3 NA WCS " LOCA TASCS Medium B-J 11 0 3 2 NA WCS " PLOCA-OC TASCS Medium B-J 29 15 5 NA WCS " LOCA None Low B-J 76 4 2 NA WCS PLOCA None Low B-J 10 1 0 NA r"

WCS PLOCA-OC None Low B-J, CL3 18 4 0 NA Page 26 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Table 5.1 Inspection Location Selection Comparisons Between ASME Section XI and Code Case N-716 Notes:

(1) System abbreviations are defined in Section 7.

(2) The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716.

Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) in a BWR to be credited toward the 10% requirement. Also, this column is used to denote those welds 2-inch and smaller that will receive a VT2 exam.

(3) The failure potential rank for high safety significant (HSS) locations is then assigned as "High", "Medium", or "Low" depending upon potential susceptibility to the various types of degradation. Note: Low safety significant (LSS) locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").

Page 27 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 APPENDIX A

SUMMARY

OF INDUSTRY PEER REVIEW FINDINGS FOR THE NMP2 INTERNAL EVENTS PRA MODEL UPDATE Page 28 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description Assoc. SR Basis for Peer Review Team Fniiigeciino SSignificance Suggested Resolution Code Case N-716 Impact 1-1 Demands from causes other than DA-C6 SR requires all types Include demands from the Open - Insignificant Impact surveillance tests were not included in DA-C7 of demands be four causes listed in the This was looked at during the Unit 1 update the collection of plant-specific data. counted or estimated. SR. Perhaps use and considered again during the Unit 2 Mitigating System update. It is slightly conservative and not (This Finding originated from Performance Indicator considered significant to estimate using Supporting Requirement (SR) (MSPI) estimates for surveillance procedures. Note that MSPI no DA-C6) MSPI components longer counts actual events.

because that program includes all demands (except post maintenance test).

1-2 Maintenance Rule unavailability data DA-C13 SR specifically says to Either exclude Closed - Minor Impact (Reduction) were used, which include include UA events Maintenance Rule Section 3 of the Data Analysis (DA) unavailability during plant shutdowns only occurring while unavailability data while Notebook and the model were updated with if that component is required to be the plant is at power. the plant is shut down, or a maintenance unavailability calculation operable. SR states that only at power provide more justification that does not include unavailability during unavailability should be used. why using such data does non-power operation.

NUREG/CR-6890 Vol. 2, Table A-2, not significantly affect the data indicate that DG unavailability results if only at power during shutdown is 5 to 10 times unavailability were to be higher than during power operation. used.

(This Finding originated from SR DA-C-13)

Page 29 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description 1Assoc. SR Basis for Significance Peer Review Team Suggested Resolution Code Case N-716 Impact 1-9 The selection of a failure probability LE-D4 More realistic failure Reconsider the 1.OE-4 Closed - Minor Impact (Increase) of 1.OE-4 for the low-pressure system probabilities of 0.1 or failure probability or Section 5 of the DA Notebook was revised component(s) rupturing given 0.01 would increase provide detailed to provide a more detailed evaluation of the exposure to RCS pressure and the frequencies of justification for such a NMP2 piping and heat exchanger temperature is optimistic given the these ISLOCA low probability. fragilities. As a result, the probability of information provided in the sequences by a factor rupture was revised in the model, which referenced NUREG/CR-5603. of 100 to 1000. varies for each system from 0.05 to 0.003.

(This Finding originated from SR LE-D4) 1-11 Several spray events identified (for IFEV-A5 Incorrect frequencies Use the spray frequencies example, FDSWCB 1 and FDSWCB2 (too low) were used for these initiating events. Closed - Minor Impact (Increase) in Table 5.1 of the Internal Flooding for these internal flood Check other internal Reviewed the IF Notebook Main Report (IF) Notebook, use flood frequencies initiators, flooding initiators for and Appendix B for potential spray events rather than spray frequencies from correct type and and frequency. The following changes were EPRI Report 1013141. There could frequency. required:

be others. (1) Initiators FDSWCB 1, FDSWCB2 and FDSWCB5 were changed to spray (This Finding originated from frequency initiating events because there is SR IFEV-A5) no detection and no propagation from these rooms.

(2) North Auxiliary Bay panel impact corrected in Appendix B (no PRA impact).

(3) Sections 4.3, 4.5, 4.6 and 5.4 of the IF Notebook were updated to include the screened spray events where PRA equipment was affected.

Page 30 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Basis for Peer Review Team Finding J Finding Description Assoc. SR Significance Suggested Resolution Code Case N-716 Impact 2-5 P. 2-7 of the DA Notebook states that DA-D1 It is not acceptable to Perform Bayesian update Closed - Minor Impact (Decrease) a Bayesian analysis was not done skip performing a when data is available and Section 2 of the DA Notebook and model when there are no plant-specific Bayesian update when zero plant-specific were updated with Bayesian analysis for failures. This is unacceptable for zero plant-specific failures are observed, or, zero events down to failure rates on the Category II or Category III. failures are observed. alternatively, show that it order of 1E-3. The conservatism of not is unlikely to get the performing this update for lower failure The discussion justifying not required number of rates is shown to be minor.

performing such updates on p. 2-6 demands to significantly and 2-7 of the DA Notebook is change the failure misleading because of the very small probability for specific failure probabilities involved in the equipment showing zero example given. failures.

Based on NUREG/CR-6928 parameters for distributions with as few as 200 to 1000 demands, the posterior mean could drop by a factor of 2.

(This Finding originated from SR DA-D I)

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update T Basis for Peer Review Team Code Case N-716 Impact Finding Finding Description Assoc. SR Significance Suggested Resolution 2-6 A critical test of the posterior that is DA-D4 Consistency between Perform recommended Closed - Minor Impact (Increase) suggested in this Supporting the plant-specific data consistency analyses for Section 2.7 of DA Notebook was updated Requirement is: and the prior was not all data. to include a test of key distributions with (c) examination of inconsistencies evaluated. A documentation of methodology. A few between the prior distribution and the representative distributions were identified as potentially plant-specific evidence to confirm example of such an inconsistent (prior versus posterior and that they are appropriate, inconsistency is plant data). As a result, the uncertainty in provided, the prior distribution was increased to be There is at least one case in which more representative of plant data.

data is inconsistent-Motor Operated Valve (MOV) (lake) fails to open.

There were 6 failures in 150 demands.

The prior from NUREG/CR-6928 for MOV FTO/C has a mean of 1.07 E-3.

The method from NUREG/CR-6823, Sections 6.2.3.5 & 6.3.3.4, describe a method for consistency evaluation that suggests that greater than or equal to 2 failures would be inconsistent and that another prior should be used.

There is no documentation of any NMP2 analysis like this.

(This Finding originated from SR DA-D4)

Page 32 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description Assoc. SR Basis for Peer Review Team FindingFindingDescriptionAssoc._SR Significance Suggested Resolution Code Case N-716 Impact 2-9 Section 2.12 of the Service Water SY-B8 This is an isolated Provide discussion of Closed - Documentation Only System (SWS) Notebook, which deals example of weakness effects on SWS of Section 2.12 of SY.04 was corrected to with Component Spatial Information, in the treatment of containment failure, address the fact that SWS is not affected by needs a small improvement. It is spatial effects. They containment failure.

stated that SWS is credited for are treated well in operation after containment failure, other notebooks.

but no justification is given for why it However, treatment of would be available, given spatial spatial effects is a effects from containment failure. clear requirement of the Standard.

(This Finding originated from SR SY-B8) 2-11 The list of sources of uncertainty has SY-C3 This is an isolated Discuss sources of Closed - Documentation Only been omitted from Section 3.5 of the occurrence of failing uncertainty in the 125 A potential important uncertainty is 125 Vdc SY Notebook. to provide this Vdc SY Notebook. associated with battery life, which was information; however, added to the Notebook.

(This Finding originated from requirements of the SR SY-C3) ASME Standard to list sources of uncertainty are clear.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description Assoc. SR Basis for Peer Review Team Finding____Finding ___Description _ _Assoc._ SR Significance Suggested Resolution Code Case N-716 Impact 2-16 This SR requires identification of QU-D6 Since Category II Identify CDF contribution Closed - Documentation Only contributors to CDF. To satisfy requires including from SSCs and operator Support system initiating event fault trees Category II (and III) requires SSCs and operator actions that contribute to have been added to the model. The IE including structures, systems, and actions that contribute IE frequencies. Notebook refers to this. SY.00 Notebook components (SSCs) and operator to IE frequencies, this provides methodology. Applicable SY actions that contribute to Initiating is a finding. notebooks develop the models.

Event (IE) frequencies. These are not included for NMP2, so only Category Open - Documentation Only I has been met. Equipment and operator contributions will be developed in the Quantification (QU)

(This Finding originated from Notebook.

SR QU-D6) The IE Notebook will be updated with correction factors.

3-5 At the time of the Peer Review, MU-F1 The lack of signatures Obtain signatures from Closed - Documentation Only various PRA documentation was widespread the personnel who were The Peer Review issuance of all notebooks notebooks were not signed by throughout the PRA designated preparer, has been signed and issued.

performers, reviewers, or approvers, notebooks. The reviewer, or approver.

preparer, reviewer, Add lines for signature (This Finding originated from and approver dates. Ensure SR MU-F l) signatures normally documentation (PRA imply that they have notebooks) reflects proper concurred with the revision number.

statements made in the associated documentation.

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description Assoc. SR Basis for Peer Review Team FindingFindingDescriptionAssoc._SR Significance Suggested Resolution Code Case N-716 Impact 3-6 The IF Notebook describes a plant IFSO-B 1 This feature has a Revise documentation Closed - Documentation Only feature important in mitigation of significant impact on (and flooding model, if The IF Notebook was revised to indicate flooding that could disable Div 1 and IF results. The IF required) to accurately that doors are currently held open by door Div 2 switchgear - "There is an open Notebook and model reflect current plant stop and there is a future modification door that is held open by a latch, should accurately configuration. which will hold doors open by latch. This which actuates to close door on a fire reflect current plant was a documentation issue only.

alarm." (pg 4.1-6). This is cited configuration.

throughout the IF notebook in multiple places. This design change has not actually been installed, but an interim measure to block the door open has been taken.

(This Finding originated from SRIFSO-B 1) 3-8 An important plant modification MU-Al This modification has Enter and track this issue Closed - Documentation Only associated with an internal flood event a significant impact on in the CRMP database. CRMP 376 issued. No impact on model or that could disable Div I and II core damage results.

Switchgear is not entered into the frequency, and Configuration Risk Management tracking of the Program (CRMP) database. modification is required by this SR and CNG-CM-1.01-3003, "Probabilistic Risk Assessment Configuration Control."

Page 35 of 39

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description Assoc. SR Basis for Peer Review Team Finding___ Finding_______Description________ Assoc.__SR Significance Suggested Resolution Code Case N-716 Impact 4-7 Several system notebooks do not have SY-A4 There are only 3 Provide completed system Closed - Documentation Only a completed system walk down. systems. walk down checklist for Only 3 System Notebooks (Automatic those systems in Depressurization System, Vapor (This Finding originated from Appendix C. Suppression and Reactor Recirc) did not SR SY-A4) have documented walk downs (NA was included) and it is stated that they are in the Drywell (inaccessible).

5-2 Routine system alignments IE-A6 Does not meet IE-A6 Include routine system contributing to initiating event Category II alignments in the Closed - Documentation Only frequencies are not included. requirements. calculation of initiating Routine alignments are already included in event frequencies, where the average initiating event frequency (This Finding originated from applicable, development. In addition, the addition of SR IE-A6) support system initiating event fault trees to the model (see Finding 2-16) adds some important alignments for these systems.

Open - Insignificant It would be a significant effort to add the type of factors that are typically reserved for EOOS risk management modeling such as V2scram testing, etc. This will have to wait until a plant reliability program is developed (e.g., scram, turbine trip risk).

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description Assoc. SR Basis for Peer Review Team Finding____ Fdg siiAoSSignificance Suggested Resolution Code Case N-716 Impact 6-1 In some cases the assignment of a HR-Gl Failure to perform a Identify risk-significant Closed - Documentation Only conservative screening human error detailed analysis for HFEs in the PRA model, Section 1 of HRA Notebook updated to probability (HEP) value may not have the estimation of and perform detailed explicitly identify HEPs based on been appropriate given the risk HEPs that represent analysis using appropriate screening, the basis for screening, and their significance of the operator action it significant human human reliability analysis importance.

represents. In particular, the use of a failure events (HFEs). (HRA) methodology(ies).

conservative screening value of 1E-02 Open - Insignificant assigned to the HEP Detailed HRA will be considered in future ZHS05_HSROOMCOL, "Operator updates as appropriate.

Fails to open HPCS ROOM Doors and HVAC Duct," may not have been appropriate given the risk significance of the HPCS room cooling support system.

(This Finding originated from SR HR-G 1)

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Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 9

10 CFR 50.55a Request Number 21SI-011 Rev. 00 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description Assoc. SR Basis for Peer Review Team Significance Suegested Resolution Code Case N-716 Impact 6-4 The most significant operator action HR-H2 Failure to satisfy HR- Perform a review of all Closed - Documentation Only in terms of importance (RRW = 2, H2 criteria for significant operator ZZOHX is not an operator action. The RAW = 11) is ZZOHX, "Failure to Capability Category recovery actions, and modeling of recovery term ZZOHX Recover Heat Removal before 1I/III/1 for significant ensure that a detailed includes an operator action ZOHO 1, which Containment Failure." There does not operator action. analysis is presented is a direct dependency for operators appear to be a detailed analysis of this which includes performing containment heat removal.

operator action with regard to consideration of ZZOHX is an equipment recovery value for procedure availability and operator procedure availability and failure to recover loss of containment heat training (nor is justification given for operator training (or removal, given ZOHO 1 was previously omission), nor were shaping factors justification given for successful. Agree that the basis for ZZOHX and sufficiency of manpower for omission), as well as in Section 5 of the DA Notebook needs performing this recovery action consideration of the improvement and this has been updated.

included in the evaluation which shaping factors and Also, sufficiency of manpower for actions documents this recovery action. sufficiency of manpower required after one day is not considered an for performing the issue.

(This Finding originated from recovery actions.

SR HR-H2) 6-5 The Accident Sequence (AS) AS-Cl The AS analysis Revise the AS Notebook Closed - Documentation Only Notebook does not contain the event documentation does to include all applicable The final post Peer Review issuance of the tree top event fault trees, which are not provide sufficient top-logic fault trees, and AS Notebook has all the documentation in necessary for understanding the information to additional description in the AS Notebook as suggested versus accident sequence logic, facilitate PRA the notebook to explain external (facilitates review etc).

applications, the top event logic.

(This Finding originated from upgrades, and peer SRAS-C1) review.

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I..

Nine Mile Point Nuclear Station, Unit 2 Third Inservice Inspection Interval 10 CFR 50.55a Request Number 21SI-011 Rev. 00 9 Appendix A - Summary of Industry Peer Review Findings for the NMP2 Internal Events PRA Model Update Finding Finding Description F Assoc. SR Basis for Significance Peer Review Suggested Team Resolution Code Case N-716 Impact 6-10 Based on a review of the design IFSN-A14 Table 4-14 indicates Revise Table 4-14 to Closed - Documentation Only features, detection and response that the South Aux change YES to NO under Footnote (1) was added to the "Yes" which section, this supporting requirement Service Bldg can be the column for Criteria #3 states "There is no detection in the South appears to have been met for the screened based upon for the South Aux Service Aux Service Building. However, there is no above areas except for the South Aux the presence of flood Bldg. PRA equipment here, the piping is Service Bldg. detection. The NMP2 relatively small and there is reliable IF Notebook, Section detection, isolation and significant time (This Finding originated from 4.2.6, does not available when propagation occurs to SR IFSN-A14) indicate that there is Turbine and or Control buildings."

detection for this area.

The responsible Constellation engineer corroborated this conclusion.

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