NL-10-2355, Request for Approval of Risk-Informed/Safety Based Inservice Inspection Alternative for Class 1 and 2 Piping FNP-ISI-AL T-12, Version 2.0

Request for Approval of Risk-Informed/Safety Based Inservice Inspection Alternative for Class 1 and 2 Piping FNP-ISI-AL T-12, Version 2.0

ML110060173
Person / Time
Site:	Farley
Issue date:	01/05/2011
From:	Ajluni M Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-10-2355
Download: ML110060173 (59)

Text

Mark J. Ajluni, P.E.

Southern Nuclear Nuclear licensing Director Operating Company. Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 T81205.992.7673 Fax 205.992.7885 January 5, 2011 SOUTHERN'\\'

COMPANY Docket Nos.: 50-348 NL-10-2355 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 Request for Approval of Risk-Informed/Safety Based Inservice Inspection Alternative for Class 1 and 2 Piping FNP-ISI-AL T-12, Version 2.0 Ladies and Gentlemen:

Pursuant to the requirements of 10 CFR 50.55a (a) (3) (i), Southern Nuclear Operating Company (SNC), the licensee for Farley Nuclear Plant (FNP) Units 1 and 2, requests authorization to implement Risk-Informed/Safety Based Inservice Inspection (RIS_B lSI) alternative FNP-ISI-ALT-12. This alternative will be used in lieu of the existing ASME Section XI Code Category B-F, B-J, C-F-1, and C-F-2 requirements for examination of Class 1 and 2 piping welds. This alternative, which is described in Enclosure 1 to this letter, has been developed in accordance with Code Case N-716, "Alternative Piping Classification and Examination Requirements. II Prior to submittal to the NRC, Version 1.0 of alternative FNP-ISI..ALT-12 was revised to provide justification that the exclusion of fire, seismic, and external hazards in the PRA model had no effect on the results of the Risk-Informed lSI program. The need for this information was identified in an NRC request for supplemental information received by Three Mile Island Nuclear Station on September 23, 2010.

SNC plans to implement the proposed alternative during the fourth 10-year inservice inspection interval that began on December 1, 2007. Implementation details are provided in the alternative. To facilitate the NRC's review, this alternative contains a template format modeled after previous submittals that the NRC has approved and a detailed evaluation of the PRA adequacy, including a gap analysis performed against Regulatory Guide 1.200. SNC requests approval of the RIS_B lSI Program by October 1, 2011, to support the Fall 2011 Unit 2 Outage.

U.S. Nuclear Regulatory Commission Log: NL-10-2355 Page 2 This letter contains no NRC commitments. If you have any questions, please contact Jack Stringfellow at (205) 992-7037.

Sincerely, M. J. Ajluni Nuclear Licensing Director MJAIT AH/Iac

Enclosure:

Proposed Alternative In Accordance With 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, Version 2.0 cc:

Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Ms. P. M. Marino, Vice President - Engineering RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley Mr. P. Boyle, NRR Project Manager

Farley Nuclear Plants Units 1 and 2 Request for Approval of Risk-Informed/Safety Based Inservice Inspection Alternative for Class 1 and 2 Piping FNP-ISI-ALT-12, Version 2.0

Plant Site - Unit:

Interval - Dates:

Requested Date for Approval:

ASMECode

. Components Affected:

Applicable Code Edition and Addenda:

Applicable Code Requirements:

Reason for Request:

Proposed Alternative and Basis for Use:

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3){i)

FNP-ISI-ALT-12, VERSION 2.0 Farley Nuclear Plant - Units 1 and 2 (FNP-1&2).

Fourth lSI Interval-December 1,2007 through November 30,2017 Approval is requested by October 1,2011 to support the Fall 2011 Unit 2 outage.

All Class 1 and 2 piping welds - Examination Categories B-F, B-J, C-F-1, and C-F-2.

The applicable Code edition and addenda is ASME Section XI, Rules for In service Inspection of Nuclear Power Plant Components, 2001 Edition through the 2003 addenda. In addition, as required by 10 CFR 50.55a, piping ultrasonic examinations are performed per ASME Section XI, 2001 Edition, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems.

For the current inservice inspection (lSI) program at FNP-1 &2, IWB-2200, IWB-2420, IWB-2430, and IWB-2500 provide the examination requirements for Category B-F and Category B-J welds. Similarly, IWC-2200, IWC-2420, IWC-2430, and IWC-2500 provide the examination requirements for Category C-F-1and C-F-2 welds.

The objective of this submittal is to request the use of a risk-informed/safety based (RIS_B) lSI process for the inservice inspection of Class 1 and 2 piping.

In lieu of the existing Code requirements, Southern Nuclear Operating company (SNC) proposes to use a RIS_B process as an alternate to the current lSI program for Class 1 and 2 piping. The RIS_B process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI, Division 1.

Code Case N-716 is founded, in large part, on the RI-ISI process described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. S A, Revised Risk-Informed Inservice Inspection Evaluation Procedure, December 1999 (ADAMS Accession No. ML013470102) which was previously reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC).

In general, a risk-informed 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 risk-informed guidelines. These processes result in a program consistent E1-1

Duration of Prol!osed ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WrrH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 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.

NRC approved EPRI TR 112657, Rev. B-A includes steps which, when successfully applied, satisfy the guidance provided in Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis and RG 1.178, An Approach For Plant-Specific Risk-Informed Decision Making for Inservice Inspection of Piping. These steps are:

Scope definition Consequence evaluation Degradation mechanism evaluation Piping segment definition Risk categorization Inspection/NDE selection Risk impact assessment Implementation monitoring and feedback These same steps were also applied to this RIS_B process and it is concluded that this RIS_B process alternative also meets the intent and principles of Regulatory Guides 1.174 and 1.178.

In general, the methodology in Code Case N-716 replaces a detailed evaluation of the safety significance of each pipe segment required by EPRI TR 112657, Rev. B-A with a generic population of high safety-significant segments, supplemented with a rigorous flooding analysis to identify any plant-specific high safety-significant segments (Class 1, 2, 3, or Non-Class).

The flooding analysis was performed in accordance with Regulatory Guide 1.200 and Addendum B to RA-S-2002 of the 2005 version of ASME RA-Sb 2005, Standard for Probabilistic Risk Assessment for Nuclear Plant Applications. (The flooding analysis did not identify any plant-specific high safety-significant segments).

By using risk-insights to focus examinations on more important locations while meeting the intent and principles of Regulatory Guides 1.174 and 1.178, this proposed RIS_B program will continue to maintain an acceptable level of quality and safety. Additionally, all piping components, regardless of risk classification, will continue to receive ASME Code-required pressure testing, as part of the current ASME Code,Section XI program. Therefore, approval for this alternative to the requirements of IWB-2200, IWB-2420, IWB-2430, and IWB-2500 (Examination Categories B-F and B-J) and IWC-2200, IWC 2420, IWC-2430, and IWC-2500 (Examination Categories C-F-1 and C-F-2) is requested in accordance with 10 CFR 50.55a(a)(3)(i). A Farley specific Template is attached that mirrors previous RIS B submittals to the NRC.

Through November 30, 2017.

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Alternative:

Precedents:

References:

Status:

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(I)

FNP-ISI-ALT-12, VERSION 2.0 Similar alternatives have been approved for Vogtle Electric Generating Plant, Donald C. Cook 1 and 2, Grand Gulf Nuclear Station, and Waterford-3.

Vogtle Electric Generating Plant Safety Evaluation - See ADAMS Accession No. ML100610470.

D. C. Cook Safety Evaluation - See ADAMS Accession No. ML072620553.

Grand Gulf Nuclear Station Safety Evaluation-See ADAMS Accession No. ML072430005.

Waterford-3 Safety Evaluation - See ADAMS Accession No. ML080980120.

Awaiting NRC approval.

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ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 TEMPLATE SUBMITTAL APPLICATION OF ASME CODE CASE N-716 RISK-INFORMED/SAFETY-BASED (RIS_B)

INSERVICE INSPECTION PROGRAM PLAN E1-4

AC AFW AS ASEP ASME ATWT BER BL-PRA CAFTA CC CC CCDP CCF CCW CDF CIV Class 2 LSS CLERP CS CVCS DA DC OM E-C ECSCC EOOS FAC F&O FLB FT FW HELB HEP HFE HR HRA HSS IE IF IFIV IGSSC I LOCA IPE ISEAL LE LERF LOCA ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-AL T -12, VERSION 2.0 Technical Acronyms/Definitions Used in the Template Alternating Current Auxiliary Feedwater Accident Sequence Analysis Accident Sequence Evaluation Program American Society of Mechanical Engineers Anticipated Transient Without Trip Break Exclusion Region (synonymous with HELB)

Base Line PRA Computer-Aided Fault Tree Analysis PRA abbreviation for Capacity Category Crevice Corrosion Conditional Core Damage Probability Common Cause Failure Component Cooling Water Core Damage Frequency Containment Isolation Valve Class 2 Pipe Break in LSS Piping Conditional Large Early Release Probability Containment Spray Chemical Volume and Control System Data analysis Direct Current Degradation Mechanism Erosion-Corrosion External Chloride Stress Corrosion Cracking Equipment Out of Service Flow-Accelerated Corrosion Facts and Observations Feedwater Line Break Fault tree Feedwater High Energy Line Break (synonymous with BER)

Human Error Probability Human Failure Event Human Reliability Human Reliability Analysis High Safety-Significant Initiating Events Analysis Internal Flooding Inside First Isolation Valve Intergranular Stress Corrosion Cracking Isolable Loss of Coolant Accident Individual Plant Evaluation Isolable RCP seal injection line LOCA LERF Analysis Large Early Release Frequency Loss of Coolant Accident E1-5

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(I)

FNP-ISI-ALT-12, VERSION 2.0 Technical Acronyms/Definitions Used in the Template (Continued)

LOSP LSS MAAP MIC MOV MS MU NDE NNS NPS PBF PIT PLOCA POD PPLOCA PRA PSA PSDC PPSDC PWR: FW PWROG PWSCC QU RC RCP RCPB RG RHR RI-BER RI-ISI RIS_B RM RPV SBO SC SDC SEAL SI SLB SGTR SSBI SSBO SSC SR SW SXI Sy Loss of Off-Site Power Low Safety-Significant Modular Accident Analysis Program Microbiologically-Influenced Corrosion Motor Operated Valve Main Steam Model Update Nondestructive Examination Non-Nuclear Safety Nominal Pipe Size Pressure Boundary Failure Pitting Potential Loss of Coolant Accident Probability of Detection Potential LOCA in Class 2 Piping Requiring Failure of Two Check Valves in Series Probabilistic Risk Assessment Probabilistic Safety Assessment Potential LOCA in a shutdown cooling line between the isolation valves Potential LOCA in a shutdown cooling line outside the two isolation valves Pressurized Water Reactor: Feedwater Pressurized Water Reactor Owner's Group Primary Water Stress Corrosion Cracking Quantification Reactor Coolant Reactor Coolant Pump Reactor Coolant Pressure Boundary Regulatory Guide Residual Heat Removal Risk-Informed Break Exclusion Region Risk-Informed Inservice Inspection Risk-Informed/Safety Based Inservice Inspection Risk Management Reactor Pressure Vessel Station Blackout Success Criteria Shutdown Cooling RCP Seal Injection Line LOCA Safety Injection Steam Line Break Steam Generator Tube Rupture Main Steam or Feedwater Break inside the Outer CIV Main Steam or Feedwater Break Beyond the Outer CIV Systems, Structures, and Components Supporting Requirements Service Water Section XI Systems Analysis E1-6

TASCS TGSCC THERP TR TT UET Vol WOG

%LOSPF

%LOSPG

%LOSSACF

%LOSSACG ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Technical Acronyms/Definitions Used in the Template (Continued)

Thermal Stratification, Cycling, and Striping Transgranular Stress Corrosion Cracking Technique for Human Error Rate Production Technical Report Thermal Transients Unfavorable Exposure Time Volumetric Westinghouse Owner's Group Specific Initiating Event Specific Initiating Event Specific Initiating Event Specific Initiating Event E1-7

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table of Contents

1. Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 PRA Quality

2. Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section XI 2.2 Augmented Programs

3. Risk-Informed/Safety-Based lSI Process 3.1 Safety Significance Determination 3.2 Failure Potential Assessment 3.3 Element and NDE Selection 3.3.1 Current Examinations 3.3.2 Successive Examinations 3.3.3 Scope Expansion 3.3.4 Program Relief Requests 3.4 Risk Impact Assessment 3.4.1 Quantitative Analysis 3.4.2 Defense-in-Depth 3.5 Implementation 3.6 Feedback (Monitoring)

4. Proposed lSI Plan Change

5. References/Documentation Attachment A - FNP PRA Quality Review E1-8

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(I)

FNP-ISI-ALT-12, VERSION 2.0

1. INTRODUCTION Farley Nuclear Plant Units 1 and 2 (FNP) is currently in the fourth inservice inspection (lSI) interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section XI Code for Inspection Program B. FNP plans to implement a risk informed/safety-based inservice inspection (RIS_B) program in the second inspection period of the fourth lSI interval. The fourth lSI interval began on December 1,2007.

The ASME Section XI Code of record for the fourth lSI interval at FNP is the 2001 Edition through the 2003 Addenda for Examination Category B-F, B-J, C-F-1, and C-F-2 Class 1 and 2 piping components.

The RIS_B process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1, which is founded in large part on the Risk-Informed lSI (RI-ISI) process as described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure.

The Unit 1 RIS_B application is based on Revision 9 of the PRA model. A complete Peer Review of the Unit 1 model was completed in March 2010. Peer Review findings and the resolution of each finding is presented in Attachment A. The Unit 2 PRA model is currently undergoing revisions similar to those performed for Unit 1. While changes that impact this RIS_B application are not expected, SNC will perform a review of the Unit 2 PRA model by September 30, 2011 (prior to the Second Period lSI examinations) and make any necessary changes to the HSS scope.

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 1.174, An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions On Plant-Specific Changes to the Licensing Basis, and Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.

1.2 Probabilistic Risk Assessment (PRA) Quality The methodology in Code Case N-716 provides for examination of a generic population of HSS segments, supplemented with a rigorous flooding analysis to identify any plant-specific HSS segments. Satisfying the requirement for the plant specific analysiS requires confidence that the flooding PRA is capable of successfully identifying any significant flooding contributors that are not identified in the generic population. Regulatory Guide (RG) 1.200, An approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities was used to demonstrate that the PRA analysis is adequate to support a risk-informed application. Regulatory Guide 1.200 further indicates that an acceptable approach for ensuring technical adequacy is to perform a peer review.

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ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3){i)

FNP-ISI-ALT-12, VERSION 2.0 The technical adequacy of the Farley PRA model Revision 9 was determined by demonstrating through a March 2010 Peer Team review that the Farley PRA model meets the technical elements and associated supporting requirements (SRs) of the ASMEIANS PRA Standard (RA-Sa-2009) as endorsed in NRC RG 1.200, Revision 2 (for Internal Events only). The resolutions to Peer Team results shown in Table A-2 of Appendix A demonstrate that the technical adequacy of the Farley PRA model is robust; however, there is still documentation to complete. Guidance from EPRI Report 1 021467, Probabilistic Risk Assessment Technical Adequacy Guidance for Risk Informed In-Service Inspection Programs was used to evaluate the documentation gap. The EPRI report indicates that peer-review findings and/or gaps related to documentation that do not impact the results would still allow the PRA model to support development of an RI-ISI program. This is the case for the Farley documentation; therefore, the model is suitable for use in support of this RIS_B Inservice Inspection application.

It should be noted that hazard groups such as internal fires, seismic events, high winds, external floods, and other external hazards as addressed in the ASME/ANS PRA Standard (RA-Sa-2009) are not included in the Farley PRA model that was used to support this submittal. However, based on the following, the Farley PRA model is considered technically adequate to support this RIS_B application without inclusion of these hazard groups.

Draft Regulatory Guide DG-12261 (Proposed Revision 2 of Regulatory Guide 1.174) indicates that:

A qualitative treatment of the missing modes and hazard groups may be sufficient when the licensee can demonstrate that those risk contributions would not affect the decision; that is, they do not alter the results of the comparison with the acceptance guidelines in Section 2.4 of this guide.

Section 2.2 of EPRI1021467, Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs provides justification for Risk-Informed lSI supporting analyses being based only on internal events PRAs (and exclusion of these other hazard groups). The basis for excluding each of these hazard groups is discussed below using the applicable information from EPR11021467.

Internal Fire Events - The potential contribution of piping failure to internal fire risk is insignificant as the failure probability of piping is insignificant compared to the failure probability of other systems, structures and components (SSCs), such as pumps, valves and power supplies. Fire events are also not likely to present significantly different challenges to the piping in the scope of this application.

Meeting defense-in-depth and safety margin principles provides additional assurance that this conclusion will remain valid. lSI is an integral part of defense in-depth, and the Risk-Informed lSI process will maintain the basic intent of lSI (Le.

identifying and repairing flaws) and thus provide reasonable assurance of an E1-10

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 ongoing substantive assessment of piping condition. In addition, there are no changes to design basis events, and thus Safety Margins are maintained.

Seismic Events - Well engineered systems and structures (e.g. piping systems) are seismically rugged. IPEEE and other industry and NRC studies (e.g. EPRI TR 1000895, NUREG/CR-5646) have shown piping systems to have seismic fragility capacities greater than the screening values typically used in seismic assessment and are not considered likely to fail during a seismic event. lSI is not considered in establishing fragilities of such SSCs. Meeting defense-in-depth and safety margin principles provides assurance that this conclusion will remain valid. lSI is an integral part of defense in depth, and the Risk-Informed lSI process will maintain the basic intent of lSI (I.e. identifying and repairing flaws) and thus provide reasonable assurance of an ongoing substantive assessment of piping condition.

In addition, there are no changes to design basis events and thus Safety Margins are maintained.

High Winds, External Floods, and Other External Hazards - The purpose of developing a Risk-Informed lSI program is to define an alternative inservice inspection strategy for piping systems. Other hazards (e.g. high wind, external floods) are not considered in the development of an in-service inspection program for piping. The reasons include: the structural ruggedness of the piping systems, location, as relevant systems are typically inside well engineered structure, and the consequence assessment for internal events already includes the consideration of spatial impacts. In addition, the substantial industry experience with plants implementing Risk-Informed lSI programs has not identified changes based upon insight from the evaluation of these other external hazards. The very small potential impact on the potential for piping failure of a Risk-Informed lSI process, and the approaches to maintaining defense in depth and safety margins summarized above, provide confidence in this conclusion.

In conclusion, excluding hazard groups such as internal fires, seismic events, high winds, external floods, and other external hazards as addressed in the ASMEIANS PRA Standard (RA-Sa-2009) has an insignificant impact on the RIS_B analysis and will not change the conclusions derived 'from the RIS_B process.

2. PROPOSED ALTERNATIVE TO CURRENT 151 PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components.

The alternative RIS_B Program for piping is described in Code Case N-716. The RIS_B Program will be substituted for the current program 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 alternatively providing an acceptable level of quality and safety.

Other non-related portions of the ASME Section XI Code will be unaffected.

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ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 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 with the RIS_B application scope (e.g., Class 1 and 2 piping).

• The plant augmented inspection program for high-energy line breaks outside containment, implemented in accordance with Technical Specification 5.5.16, has not been revised in accordance with the risk-informed break exclusion region methodology (RI-BER) described in EPRI Report 1006937, Extension of EPRI Risk Informed lSI Methodology to Break Exclusion Region Programs. Therefore, 100%

of these welds will continue to be examined per the Technical Specification 5.5.16 requirements. It is the intention of SNC to implement the RI-BER program later during the fourth lSI interval.

• A plant augmented inspection program has been implemented at FNP in response to NRC Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems. This program was updated in response to MRP-146, Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines. The thermal fatigue concern addressed was explicitly considered in the application of the RIS_B process and is subsumed by the RIS_B Program.

• The plant augmented inspection program for flow accelerated corrosion (FAC) per 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.

Since the issuance of the NRC safety evaluation for EPRI TR 112657, Rev. B-A, several instances of primary water stress corrosion cracking (PWSCC) of unmitigated Alloy 821182 welds has occurred at pressurized water reactors. For FNP, the unmitigated Alloy 82/182 Category B-F dissimilar metal welds (greater than NPS 1) subject to PWSCC are the three RPV hot leg nozzle to safe-end welds and the three cold leg nozzle to safe-end welds. The Steam Generator dissimilar metal welds are not subject to PWSCC because the welds are Alloy 52/152, and all of the pressurizer dissimilar metal welds greater than 1" Nominal Pipe Size (NPS) have been overlaid with Full Structural Weld Overlays (FSWOL). All of the overlaid welds have been removed from the risk-informed program and will be examined in accordance with the requirements set forth in the NRC safety evaluation for the weld overlays.

Even though Code Case N-716 only considers the RPV hot leg nozzle Alloy 82/182 weld locations to be susceptible to PWSCC, SNC has selected all six welds to be ultrasonically examined for PWSCC within the scope of Code Case N-716. Code Case N-716 requires the examination of these welds every ten years. However, the examination frequency for these eight welds is currently based on the frequencies established by the requirements of Materials Reliability E1-12

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Program (MRP)-139, Revision 1. MRP-139 currently requires that the unmitigated hot legs be examined on a five year frequency and the unmitigated cold legs be examined on a six year frequency. These frequencies are subject to change based on factors such as industry experience and issuance of NRC rule making. The RIS_B Program will not be used to eliminate any MRP-139 or regulatory requirements.

Per Code Case N-716 (Table 1, Item No. 1.15, Elements Subject to Primary Water Stress Corrosion Cracking (PWSCC), selected butt welds are subject to volumetric examination. Per Note 3 of Table 1, the examination includes essentially 100% of the examination location. When the required examination volume or area cannot be examined due to interference by another component or part geometry, limited examinations shall be evaluated for acceptability.

Areas with acceptable limited examinations (coverage less or equal to 90%), and their bases, shall be documented and submitted for relief per the requirements of 10 CFR 50.55a(g)(5)(iv).

3. RISK-INFORMED/SAFETY-BASED lSI PROCESS The process used to develop the RIS_B Program conformed to the methodology described in Code Case N-716 and consisted of the following steps:

Safety Significance Determination (see Section 3.1)

Failure Potential Assessment (see Section 3.2)

Element and NDE Selection (see Section 3.3)

Risk Impact Assessment (see Section 3.4)

Implementation Program (see Section 3.5)

Feedback Loop (see Section 3.6)

Each of these six steps is discussed below:

3.1 Safety Significance Determination The systems assessed in the RIS_B Program are provided in Table 3.1 a (Unit 1) and Table 3.1.b (Unit 2). The piping and instrumentation diagrams and additional plant information, including the existing plant lSI Program were used to define the piping system boundaries. Per Code Case N-716 requirements, piping welds are assigned safety-significance categories, which are then used to determine the examination treatment requirements. High safety-significant (HSS) welds are determined in accordance with the requirements below. Low safety-significant (LSS) welds include all other Class 2, 3, or Non-Class welds.

(1) Class 1 portions of the reactor coolant pressure boundary (RCPB), except as provided in 10 CFR 50.55a(c)(2)(i) and (c)(2)(ii);

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ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 (2) Applicable portions of the shutdown cooling pressure boundary function. That is, 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 RCPS from the reactor pressure vessel (RPV) to the second isolation valve (Le., 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 (Le., 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 [> 4 inch nominal pipe size (NPS)]

of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve; (4) Piping within the break exclusion region (SER) greater than 4" NPS for high energy piping systems as defined by the Owner. Per Code Case N-716, this may include Class 3 or Non-Class piping. SER piping at FNP is Class 2 and Non-Class.

(5) Any piping segment whose contribution to Core Damage Frequency (CDF) is greater than 1 E-06 [and per NRC feedback on the Grand Gulf and D. C. Cook RIS_S applications 1 E-07 for Large Early Release Frequency (LERF)] based upon a plant-specific PSA of pressure boundary failures (e.g., pipe whip, jet impingement, spray, inventory losses). This may include Class 3 or Non-Class piping. A review of the flooding PRA was performed to identify any piping whose failure could cause flooding that could significantly impact safety significant components. During the review, it was determined that in order to reduce the flooding scenario frequencies due to the postulated rupture of fire protection piping in auxiliary building areas (210, 211, 228, and 234 for Unit 1 and 2210, 2211, 2228, and 2234 for Unit 2) that supplementary visual inspection of the aSSOCiated fire protection piping is required every quarter. With these inspections, no piping segments with a contribution to CDF greater than 1 E-06 (1 E-07 for LERF) were identified. Upon NRC approval and subsequent implementation of this alternative, FNP will perform a supplementary visual inspection of fire protection piping in these locations every quarter.

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 NRC approved EPRI TR-112657 (Le., the EPRI RI-ISI methodology), with the exception of the deviation discussed below.

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

E1-14

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 A deviation to the EPRI RIS_B methodology has been implemented in the failure potential assessment for FNP. Table 3-16 of EPRI TR-112657 contains the following criteria for assessing the potential for Thermal Stratification, Cycling, and Striping (T ASCS). Key attributes for horizontal or slightly sloped piping greater than NPS 1 include:

1. The potential exists for low flow in a pipe section connected to a component allowing mixing of hot and cold fluids; or

2. The potential exists for leakage flow past a valve, including in-leakage, out-leakage and cross-leakage allowing mixing of hot and cold fluids; or

3. The potential exists for convective heating in dead-ended pipe sections connected to a source of hot 'fluid; or

4. The potential exists for two phase (steam/water) flow; or

5. The potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot Huid with turbulent flow; AND AND

~ Richardson Number> 4 (this value predicts the potential buoyancy of a stratified How)

These criteria, based on meeting a high cycle fatigue endurance limit with the actual

~T assumed equal to the greatest potential ~T for the transient, will identify locations where stratification is likely to occur, but allows for no assessment of severity. As such, many locations will be identified as subject to TASCS, where no significant potential for thermal fatigue exists. The critical attribute missing from the existing methodology, that would allow consideration of fatigue severity, is a criterion that addresses the potential for fluid cycling. The impact of this additional consideration on the existing TASCS susceptibility criteria is presented below.

);>

Turbulent Penetration TASCS Turbulent penetration is a swirling vertical flow structure in a branch line induced by high velocity flow in the connected piping. It typically occurs in lines connected to piping containing hot flowing fluid. In the case of downward sloping lines that then turn horizontal, signi'ficant top-to-bottom cyclic ~Ts can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASCS is considered for this configuration.

E1-15

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(I)

FNP-ISI-ALT-12, VERSION 2.0 For upward sloping branch lines connected to the hot fluid source that turn horizontal or in horizontal branch lines, natural convective effects combined with effects of turbulence penetration will tend to keep the line filled with hot water. If there is in-leakage of cold water, a cold stratified layer of water may be formed and significant top-to-bottom ATs may occur in the horizontal portion of the branch line. Interaction with the swirling motion from turbulent penetration may cause a periodic axial motion of the cold layer. Therefore, TASCS is considered for these configurations.

For similar upward sloping branch lines, if there is no potential for in-leakage, this will result in a well-mixed fluid condition where significant top-to-bottom ATs will not occur. Therefore, TASCS is not considered for these no in-leakage configurations. Even in fairly long lines, where some heat loss from the outside of the piping will tend to occur and some fluid stratification may be present, there is no significant potential for cycling as has been observed for the in-leakage case. The effect of T ASCS will not be significant under these conditions and can be neglected.

);>

Low flow TASCS In some situations, the transient startup of a system (e.g., shutdown cooling suction piping) creates the potential for fluid stratification as flow is established.

In cases where no cold fluid source exists, the hot flowing fluid will fairly rapidly displace the cold fluid in stagnant lines, while fluid mixing will occur in the piping further removed from the hot source and stratified conditions will exist only briefly as the line fills with hot fluid. As such, since the situation is transient in nature, it can be assumed that the criteria for thermal transients (IT) will govern.

);>

Valve leakage TASCS Sometimes a very small leakage flow of hot water can occur outward past a valve into a line that is relatively colder, creating a significant temperature difference. However, since this is generally a "steady-state" phenomenon with no potential for cyclic temperature changes, the effect of TASCS is not significant and can be neglected.

);>

Convection Heating TASCS Similarly, there sometimes exists the potential for heat transfer across a valve to an isolated section beyond the valve, resulting in fluid stratification due to natural convection. However, since there is no potential for cyclic temperature changes in this case, the effect of TASCS is not significant and can be neglected.

In summary, these additional considerations for determining the potential for thermal fatigue as a result of the effects of T ASCS provide an allowance for considering cycle severity. Consideration of cycle severity was used in previous NRC approved RIS_B program submittals for D. C. Cook, Grand Gulf Nuclear Station, Waterford-3, and the Vogtle Electric Generating Plant. The methodology used in the FNP RIS_B application for assessing TASCS potential conforms to these updated criteria. Additionally, E1-16

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 materials reliability program (MRP) MRP-146 guidance on the subject of TASCS was also incorporated into the FNP RIS_B application. It should be noted that the NRC has granted approval for RIS_B relief requests incorporating these TASCS criteria at several facilities, including Comanche Peak (NRC letter dated September 28, 2001) and South Texas Project (NRC letter dated March 5, 2002).

3.3 Element and NDE Selection Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RIS_B applications provided 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.

(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 inside first isolation valve (IFIV) (Le., 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 containment (not applicable for Farley) shall be selected.

(5)

A minimum of 10% of the welds within the break exclusion region (BER) shall be selected. Currently, there are 151 welds at Farley 1 and 138 welds at Farley 2 in the BER program. These BER welds consist of both Class 2 welds and non nuclear safety (NNS) welds in the main steam system located outside of the containment. A RI-BER program has not been implemented for these welds, so 100% of the population is currently being examined. It is the intention of SNC to implement the RI-BER program later during the fourth lSI interval.

In contrast to a number of RIS_B program applications, where the percentage of Class 1 piping locations selected for examination has fallen substantially below 10%, Code Case N-716 mandates that 10% of the HSS welds be chosen. A brief summary of the number of welds and the number selected is provided below, and the results of the E1-17

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3}(i)

FNP-ISI-ALT-12, VERSION 2.0 selections are presented in Table 3.3a (Unit 1) and Table 3.3b (Unit 2). Section 4 of EPRI TR-112657 was used as guidance in determining the examination requirements for these locations. Only those RIS_B inspection locations that receive a volumetric examination are included.

~t Class 1 Welds(1)

Total Selected Class 2 Welds(2)

Total Selected NNS Welds(3)

Total Selected All Piping Welds(4)

Total Selected 1

709 76 1,880 26 44 4

2,633 106 2

692 73 1,686 25 41 4

2,419 102 Notes:

(1) Includes all Category B-F and B-J locations. All Class 1 piping weld locations are HSS.

(2) Includes all Category C-F-1 and C-F-210cations. Of the Class 2 piping weld locations, 263 are HSS at Unit 1 and 251 are HSS at Unit 2; the remaining are LSS.

(3) The non-nuclear safety (NNS) piping weld locations are associated with Main Steam BEA.

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

3.3.1 Current Examinations FNP 1 &2 is currently using the traditional ASME Section XI inspection methodology for lSI examination of piping welds per the 2001 Edition of ASME Section XI through the 2003 Addenda.

3.3.2 Successive Examinations If indications are detected during RIS_B ultrasonic examinations, they will be evaluated per IWB-3514 (Class 1) or IWC-3514 (Class 2) to determine their acceptability. Any unacceptable flaw will be evaluated per the requirements of ASME Code Section XI, IWB-3600 or IWC-3600, as appropriate. As part of this evaluation, the degradation mechanism that is responsible for the flaw will be determined and accounted for in the evaluation. If the flaw is acceptable for continued service, successive examinations will be scheduled per Section 6 of Code Case N-716. If the flaw is found unacceptable for continued operation, it will be repaired in accordance with IWA-4000, applicable ASME Section XI Code Cases, or NRC approved alternatives. The IWB-3600 analytical evaluation will be submitted to the NRC. Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section XI. Evaluation of indications attributed to PWSCC and successive examinations of PWSCC indications will be performed in accordance with MRP-139 or a subsequent NRC rule making.

E1-18

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 3.3.3 Scope Expansion If the nature and type of the flaw is service-induced, then welds subject to the same type of postulated degradation mechanism will be selected and examined per Section 6 of Code Case N-716. The evaluation will include whether other elements in the segment or additional segments are subject to the same root cause conditions. Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms. The additional examinations will include HSS elements up to a number equivalent to the number of elements required to be inspected during the current outage. If unacceptable flaws 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 need be performed if there are no additional elements identified as being susceptible to the same root cause conditions. The need for extensive root cause analysis beyond that required for the IWB-3600 analytical evaluation will be dependent on practical considerations (Le., the practicality of performing additional NDE or removing the flaw for further evaluation during the outage).

Scope expansion for flaws characterized as PWSCC will be conducted in accordance with MRP-139 or subsequent NRC rule makings.

3.3.4 Program Relief Requests Consistent with previously approved RIS_B submittals, SNC will calculate coverage and use additional examinations or techniques in the same manner it has for traditional Section XI 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 the examinations are performed. Relief requests for those cases where greater than 90% coverage is not obtained will be submitted per the requirements of 10 CFR SO.SSa(g)(S)(iv).

No FNP relief requests are being withdrawn due to the RIS_B application.

3.4 Risk Impact Assessment The RIS_B Program development has been conducted in accordance with Regulatory Guide 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.

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 were proposed for each system. The changes included changing the number and location of inspections, and in many cases improving the effectiveness of the inspection to account for the findings of the RIS_B degradation mechanism assessment. For example, examinations of locations subject to thermal fatigue will be E1-19

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the inspection process.

3.4.1 Quantitative Analysis Code Case N-716 has adopted the NRC approved EPRI TR-1126S7 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 Regulatory Guides 1.174 and 1.17S. Section 3.7.2 of EPRI TR-1126S7 requires that the cumulative change in CDF and LERF be less than 1 E-07 and 1E-OS per year per system, respectively.

For LSS welds, Conditional Core Damage Probability (CCDP)/Conditional Large Early Release Probability (CLERP) values of 1 E-4/1 E-S were conservatively used. 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 risk informed lSI (RI-ISI) methodology. As such, the goal is to determine CCDPs/CLERPs threshold values. For example, the threshold values between High and Medium consequence categories is 1 E-4 (CCDP)/l E-S (CLERP) and between Medium and Low consequence categories are 1 E-6 (CCDP)/l 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 1 E-S to 3E-S due to an update, it will remain below the 1 E-4 threshold value; the change-in-risk evaluation would not require updating.

The updated internal flooding PRA was also reviewed to ensure that there is no LSS Class 2 piping with a CCDP/CLERP greater than 1 E-4/1 E-S. This review identified some piping in the RHR and CVCS systems located outside of containment with a CCDP greater than 1 E-4. As a result, all LSS RHR and CVCS welds were conservatively assigned CCDP/CLERP equal to 2E-3/2E-4.

With respect to assigning failure potentials for LSS piping, the criteria are defined in Table 3 of Code Case N-716. That is, those locations identified as susceptible to FAC 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, unless they have an identified potential for water hammer loads. In such cases, they will be assigned a high failure potential. Finally, 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. LSS piping may be susceptible to FAC; however, the examination for FAC is performed per the FAC program. This review was conducted similar to that done for a traditional RI-ISI application. Thus, the high failure potential category is not applicable to LSS piping. In lieu of conducting a formal degradation mechanism evaluation for all LSS piping (e.g. to determine if thermal fatigue is applicable), these locations were conservatively assigned to the Medium failure E1-20

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 potential ("Assume Medium" in Table 3.4-1 a and Table 3.4-1b) for use in the change-in-risk assessment. Experience with previous industry RIS_B applications shows this to be conservative.

FNP has conducted a risk impact analysis per the requirements of Section 5 of Code Case N-716 that is consistent with the "Simplified Risk Quantification 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 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 the table below. Consistent with the EPRI 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 LOCA CCDP bounds the large and small LOCA CCDPs for FNP). See Table below.

CCDP and CLERP Values Based on Break Location Break Estimated Upper Bound Consequence Location Designation CCDP CLERP Rank CCDP CLERP LOCA E-03 3.0E-04 HIGH 3.0E-03 3.0E-04 RCPB pipe breaks that result in a loss of coolant accident - The highest CCDP for a Medium LOCA was used (0.1 margin used for CLERP). Applies to unisolable RCPB piping of all sizes.

ILOCA(l) 1.0E-05 1.0E-06 MEDIUM 1.0E-04 1.0E-05 Isolable LOCA (1 open valve) - RCPB pipe breaks that result in an isolable LOCA - Calculated based on a Medium LOCA CCDP of 3E-3 and a valve failure to close probability of -3E-3 (0.1 margin used for CLERP). Applies to piping located between 1 st and 2nd isolation valve on charging, letdown and seal injection lines.

PLOCA(l) 3.0E-06 3.0E-07 MEDIUM 1.0E-04 1.0E-05 Potential LOCA (1 closed valve) - RCPB pipe breaks that result in a potential LOCA* Calculated based on a Medium LOCA CCDP of 3E-3 and a valve rupture probability of -1 E*3 (0.1 margin used for CLERP). Applies to piping located between 1 st and 2nd isolation valves on safety injection, alternate charging, drain lines on RC PSDC(l)

MEDIUM 1.0E-05 1.0E-06 1.0E-04 1.0E-05 Potential LOCA (1 valve)

• RCPB pipe breaks that occur in the shutdown cooling suction piping resulting in a potential LOCA at power and an isolable LOCA during shutdown. LOCA CCDP and MOV failure on demand is judged to be appropriate for lines inside containment (0.1 margin used for CLERP).

E1-21

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-AL T-12, VERSION 2.0 CCDP and CLERP Values Based on Break Location Break Estimated Upper Bound Consequence Location Designation CCDP CLERP Rank CCDP CLERP PPSDC(1) 1.0E-07 MEDIUM 1.0E-05 1.0E-06 1.0E-04 Potential LOCA (2 valves) - Class 2 potential LOCA in the shutdown cooling suction piping downstream of 2nd isolation valve. LOCA CCDP and failure of 2 MOVs to close on demand (3E-4) is judged to be appropriate for lines inside containment (0.1 margin used for CLERP)

PPLOCA(1)

<1 E-06

<1 E-07 MEDIUM 1.0E-04 1.0E-05 Potential LOCA (2 closed valves) - Class 2 potential LOCA breaks that require two check valves in series to rupture based on LOCA CCDP of 3E-3 and 2 valve ruptures <1 E-6 (0.1 margin used for CLERP). Class 2 SDC return lines between second RCS isolation valve inside containment and the containment penetration. Medium assumed rather than low because these lines support multiple cold leg injection paths.

SEAL 5.0E-05 5.0E-06 MEDIUM 1.0E-04 1.0E-05 RCP seal injection lines require a seal LOCA to occur - based on small LOCA CCDP of 5E-4 and 0.1 probability of seal LOCA (0.1 margin used for CLERP)

ISEAL

<1 E-06

<1 E-07 MEDIUM 1.0E-04 1.0E-05 i Isolable RCP seal injection lines require a seal LOCA to occur - based on small LOCA CCDP of 5E-4 and 0.1

• probability of seal LOCA and 3E-3 valve failure probability (0.1 margin used for CLERP)

SLB 5.01:-05 5.0E-06 MEDIUM 1.0E-04 1.0E-05 Main steam and feedwater breaks (SSBO, SSBI and FLB) all have the same CCDP and CLERP (0.1 margin used for CLERP)

Class 2 LSS 1.0E-04 1.0E-05 MEDIUM 1.0E-Q4 1.0E-05 Class 2 pipe breaks that occur in the remaining system piping designated as low safety significant except RHR and!

CVCS - Estimated based on upper bound for Medium Consequence.

Class 2 LSS 2.0E-03 2.0E-04 HIGH 2.0E-03 2.0E-04 AHA Class 2 RHR and CVCS pipe breaks that occur in the remaining system piping designated as low safety significant conservatively based on internal flood initiating event %FLOOD_AB223SP2 Note (1): The PRA does not explicitly model potential and isolable LOCA events, because such events are subsumed by the LOCA initiators in the PRA. That is, the frequency of a LOCA in this limited piping downstream of the first RCPB isolation valve times the probability that the valve fails is a small contributor to the total LOCA frequency. The N 716 methodology must evaluate these segments individually; thus, it is necessary to estimate their contribution. This is estimated by taking the LOCA CCDP and multiplying it by the valve failure probability.

E1-22

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 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 given as Xo and is expected to have a value less than 1 E-08. Piping locations identified as medium failure potential have a likelihood of 20xo* 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 an increased POD from application of the RIS_B approach.

Table 3.4-1 a (Unit 1) and Table 3.4-1b (Unit 2) presents a summary of the RIS_B Program versus the third lSI interval (1989 Edition of ASME Section XI Addenda) program requirements on a "per system" basis. The presence of FAG was adjusted for in the quantitative analysis by excluding its impact on the failure potential rank. The exclusion of the impact of FAG on the failure potential rank and therefore in the determination of the change-in-risk, was performed because FAG is a damage mechanism managed by a separate, independent plant augmented inspection program. The RIS_B Program credits and relies upon this plant augmented inspection program to manage this damage mechanism. The plant FAG program will continue to determine where and when examinations shall be performed. Hence, since the number of FAG examination locations remains the same "before" and "after" (the implementation of the RIS_B program) and no delta exists, there is no need to include the impact of FAG in the performance of the risk impact analysis.

As indicated in the following Unit 1 and Unit 2 tables, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RIS_B Program, and that the acceptance criteria of Regulatory Guide 1.174 and Gode Gase N-716 are satisfied.

FNP Unit 1 Risk Impact Summary With POD Credit Without POD Credit System Delta CDF Delta LERF Delta CDF Delta LERF Chemical & Volume Control 2.17E-09 2.17E-10 3.1SE-09 3.1SE-10 Main Feedwater 1.S0E-12 1.S0E-13 1.50E-12 1.S0E-13 Main Steam 1.20E-11 1.20E-12 1.20E-11 1.20E-12 Reactor Coolant

-3.89E-09

-3.89E-10 4.40E-09 4.40E-10 Residual Heat Removal 4.40E-09 4.40E-10 4.63E-09 4.63E-10 Safety Injection 3.21E-10 3.21 E-11 3.53E-10 3.53E-11 Auxiliary Feedwater O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO Containment Spray 1.20E-10 1.20E-11 1.20E-10 1.20E-11 Total 3.14E-09 3.14E-10 1.27E-OS 1.27E-09 E1-23

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3){i)

FNP-ISI-ALT-12, VERSION 2.0 FNP Unit 2 Risk Impact Summary System With POD Credit Delta CDF Delta LERF Without POD Credit Delta CDF Delta LERF Chemical & Volume Control 1.71 E-09 1.71E-10 3.17E-09 3.17E-10 Main Feedwater

-5.00E-13

-5.00E-14

-5.00E-13

-5.00E-14 Main Steam 3.65E-11 3.65E-12 3.65E-11 3.65E*12 Reactor Coolant

-1.92E-09

-1.92E-10 5.88E-09 5.88 Residual Heat Removal 4.81 E*09 4.81 E-10 5.00E-09 5.00E-10 Safety Injection Auxiliary Feedwater

~E.'0 E+OO 2.96E-11 O.OOE+OO 3.04E-10 O.OOE+OO 3.04E-11 O.OOE+OO Containment Spray

.30E-10 1.30E-11 1.30E-10 1.30E-11 Total 5.05E-09 5.05E-10 1.45E-08 1.45E-09 As shown in Tables 3.4-1 a and 3.4-1b, new RIS_B locations were selected such that the RIS_B selections exceed the Section XI selections for certain CVCS, MS, RHR and SI categories (Delta column has a positive 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, a very conservative sensitivity was conducted where the RIS_B selections were set equal to the Section XI selections (Delta changed from positive number to zero). The acceptance criteria are met when the number of RIS_B selections is not allowed to exceed Section XI.

3.4.2 Oefense-in-Oepth The intent of the inspections mandated by 10 CFR 50.55a for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system's pressure boundary. Currently, the process for selecting inspection locations is based upon terminal end locations, structural discontinuities, and stress analysis results. As depicted in ASME White Paper 92-01-01 Rev. 1, Evaluation of Inservice Inspection Requirements for Class 1f Category 8-J Pressure Retaining Welds, this methodology 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; that is, a determination of each location's susceptibility to degradation and secondly. an independent assessment of the consequence of the piping failure. These two ingredients assure 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 leak or ruptures is increased. Secondly, a generic assessment of high-consequence sites has been determined by Code Case N-716, 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 1 E-06 E1-24

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 (or 1 E-07 for LERF) be included in the scope of the application. FNP did not identify any such piping. [See paragraph 3.1 (5)].

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

3.5 Implementation Upon approval of the RIS_B Program, procedures that comply with the guidelines described in Code Case N-716 will be prepared to implement and monitor the program. The new program will be implemented during the fourth lSI interval. No changes to the Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section XI program implementing procedures will be retained and modified to address the RIS_B process, as appropriate.

3.6 Feedback (Monitoring)

The RIS_B Program is a living program that is required to be monitored continuously 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 configuration, changes to operations that could affect the degradation assessment, a review of Farley NDE results, a review of site failure information from the Farley corrective action program, and a review of industry failure information from industry operating experience (DE). Also included is a review of PRA changes for their impact on the RIS_B program. These reviews provide a feedback loop such that new relevant information is obtained that will ensure that the appropriate identification of HSS piping locations selected for examination is maintained. As a minimum, this review will be conducted on an ASME period basis. In addition, more frequent adjustment may be required as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant-specific feedback.

If an adverse condition, such as an unacceptable flaw is detected during examinations, the adverse condition will be addressed by the corrective action program and procedures. The following are appropriate actions to be taken:

A.

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

B.

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

C.

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

D.

Decide (make a decision to implement the corrective action plan).

E1-25

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 E.

Implement (complete the work necessary to correct the problem and prevent recurrence).

F.

Monitor (through the audit process ensure that the RIS_B program has been updated based on the completed corrective action).

G.

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

For preservtce examinations, SNC will follow the rules contained in Section 3.0 of N-716. Welds classified HSS require a preservice inspection. The examination volumes, techniques, and procedures shall be in accordance with Table 1 of N-716.

Welds classified as LSS do not require preservice inspection.

4. PROPOSED lSI PLAN CHANGE FNP 1 &2 is currently in the first period of the fourth lSI interval and is using the traditional ASME Section XI inspection methodology for lSI examination of piping welds. At least 16%

of the ASME Section XI piping examinations will be performed by the end of the first period of the fourth lSI interval to ensure compliance with the traditional ASME Section XI inspection methodology.

In anticipation of the approval of this RIS_B submittal, selected welds that are being examined during the 1 s1 Period, using the traditional ASME Section XI methodology, also meet the examination requirements of Table 1 of Code Case N-716. After approval of the RIS_B submittal, those welds in the RIS_B scope that were examined during the first period that also met Table 1 requirements may be credited toward the RIS_B requirements for the 1

s1 Period.

Alternatively, 1 s1 Period examinations will be completed using the traditional ASME XI methodology. Then, the 2nd and 3fd Period examinations will utilize the RIS_B methodology.

In this case, approximately one-third of the total number of RIS_B piping welds selected for examination will be examined in each of the two remaining periods.

As discussed in Section 2.2, implementation of the RIS_B program will not alter any PWSCC examination requirements for the Alloy 82/182 examinations.

A comparison between the RIS_B Program and the 1989 Edition of Section XI program requirements for in-scope piping is provided in Table 4a (Unit 1) and Table 4b (Unit 2).

5. REFERENCES/DOCUMENTATION EPRI Report 1006937, Extension of EPRI Risk Informed lSI Methodology to Break Exclusion Region Programs.

EPRI TR-112657, Revised Risk-Informed In service Inspection Evaluation Procedure, Rev.

B-A.

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

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

E1-26

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping.

Regulatory Guide 1.200, Rev 2 An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities.

USNRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for Alternative GG-ISI-002-lmplement Risk-Informed lSI based on ASME Code Case N-716, dated September 21, 2007.

USNRC Safety Evaluation for DC Cook Nuclear Plant, Units 1 and 2, Risk-Informed Safety Based lSI program for Class 1 and 2 Piping Welds, dated September 28,2007.

Appendix A of EPRI Report Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs.

Supporting Onsite Documentation Structural Integrity Calculation 0801605.302, N-716 Evaluation for Farley Units 1 &2, Rev O.

Structural Integrity Calculation 0801605.301 Degradation Mechanism Evaluation for Farley Units 1 & 2, Rev O.

E1-27

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table3.1a FNP-1 Code Case N*716 Safety Significance Determination Safety System Weld N*716 Safety Significance Determination Significance Description Count RCPB SOC PWR: FW BER CDF> 1E*6 High Low 35 Re 249 94 eves 312 116 RHR 75 363 70 SI 145 752 FW 81 151 MS 46 AFW 12 es 132 221 488

SUMMARY

RESULTS 75 FOR ALL 151 SYSTEMS 81 1617 TOTALS 2633 AFW = Auxiliary Feedwater portion of Main Feedwater CS = Containment Spray CVCS =Chemical Volume and Control System FW = Main Feedwater MS = Main Steam RC = Reactor Coolant RHR = Residual Heat Removal SI = Safety Injection J

I I

1 E1-28

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table3.1b FNP-2 Code Case N-716 Safety Significance Determination Safety System Weld N-716 Safety Significance Determination Significance Description Count RCPB SOC PWR: FW BER CDF> 1E-6 High Low 35

./

./

./

Ae 256

./

./

83

./

./

eves 297

./

117

./

./

./

AHA 73

./

./

361

./

63

./

./

./

I SI 138

./

./

576

./

i FW 81

./

./

138

./

./

MS 51

./

AFW 9

./

es 141

./

215

./

./

./

477

./

./

SUMMARY

73

./

./

RESULTS FOR ALL 138

./

./

SYSTEMS 81

./

./

./

1435 TOTALS 2419 AFW = Auxiliary Feedwater portion of Main Feedwater CS = Containment Spray CVCS Chemical Volume and Control System FW = Main Feedwater MS Main Steam RC Reactor Coolant RHR = Residual Heat Removal SI = Safety Injection E1-29

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table 3.2 Failure Potential Asses!rnent Summary Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive System(1)

TASCS IT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC RC CVCS(2)

", (3)

SI(2)

I RHR(2)

AFW(2)

I FW MS(2)

CS(2)

~~~~~

,~~~

~ ~~~ ~~~~~~~~

-~~~~~~

~~~~--~

Notes

1. Systems are described in Table 3.1a (Unit 1) and Table 3.1b (Unit 2).

2. A degradation mechanism assessment was not performed on low safety significant piping segments. This includes the AFW and es systems, as well as portions of the eves. SI, RHR, and MS systems.

3. TASeS only applies to Unit 2. (The corresponding Unit 1 welds have the same TASe degradation mechanism; however, they are administratively designated as being in the Re system).

E1-30

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table3.3a FNP-l Code Case N-716 Element Selections I

System Weld Count HSS LSS OMs N716 Selection Considerations RCPB RCPB(IFIV)

RCPB (OC)

BER Selections I

• AFW eves 4

12 TT

./

./

0 4

eves 18 TT

./

3 eves 20 None

./

./

3 eves 52 None

./

0 eves 312 0

• FW 81 None 9

MS 151 None

./

16 MS 46 0

Re 3

pwsee

./

./

3 Re 19 TASeS

./

./

5 I

IRe 23 TAses, TT

./

./

8 J

Re 17 TT

./

./

6 Re 212 None

./

./

7 I

Re 10 None

./

0 RHR 10 TASeS

./

4 I

RHR 77 TT

./

10 RHR 4

TT,lGSee

./

1 RHR 25 None

./

0 RHR 5

TASeS 0

~

RHR 45 TT 5

RHR 25 None 0

RHR 363 0

i Sl 8

IGSee

./

2 SI 12 TT

./

3 I

SI 4

TT,lGSee

./

1 SI 191 None

./

16

• SI 752 0

es 132 0

21 TT

./

./

10 107 TT

./

16 I

45 TT 5

3 pwsee

./

./

3 19 TASeS

./

./

5 10 TASeS

./

4 i

Summary 5

TASeS 0

Results All Systems 23 8

8 TASeS, TT IGsee TT,lGSee

./

./

./

./

8 2

2 232 None

./

./

10 i

278 None

./

16 106 None 9

151 None

./

16 I

1617 0

Totals 1016 1617 106 Note: Systems are described in Table 3.1 a.

E1-31

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Tale3.3b I

FNP-2 Code Case N-716 Element Selections i

System Weld Count HSS LSS OMs N716 Selection Considerations RCPB RCPB (IFIV)

RCPB (OC)

BER Selections I AFW 9

None 0

eves 3

TT

../

../

3 eves 18 TT

../

3 eves 3

TASeS, n

../

../

3 i

eves 23 None

../

../

0 eves 36 None

../

0 I

eves 297 0

FW 81 None 9

MS 138 None

../

14 I

MS 51 0

Re 3

pwsee

../

../

3 19 TASeS

../

../

5 I

~

21 TASeS, n

../

../

6 18 TT

../

../

5 Re 220 None

../

../

11 I

Re 10 None

../

0 RHR 6

TASeS

../

2 RHR 69 TT

../

9 RHR 6

TT,lGSee

../

2 RHR 36 None

../

0 RHR 48 TT 6

I RHR 25 None 0

RHR 361 0

i SI 8

IGSee

../

2 SI 1

TT

../

1 SI 2

TT,lGSee

../

1 I

SI 190 None

../

17 SI 576 0

es 141 0

I 21 TT

../

../

8 88 TT

../

13 48 TT 6

3 pwsee

../

../

3 19 TASeS

../

../

5 Summary Results 6

24 TASeS TASeS, n

../

../

../

2 9

All 8

IGSee

../

2 Systems 8

TT,lGSee

../

3 243 None

../

../

11 272 None

../

17 106 None 9

138 None

../

14 1435 0

Totals 984 1435 102 Note: Systems are described in Table 3.1b.

E1-32

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0

~~"---

System (1)

CVCS CVCS CVCS CVCS CVCS CVCS CVCS CVCS CVCS Total FWTotal MS MS MSTotal AC AC AC AC AC AC RC Total AHA AHA AHA AHA AHA AHA AHA AHA AHA RHRTotal Safety Significance High High High High High High High Low High High Low High High High High High High High High High High High High High High Low Table 3.4-1a: FNP-1 Risk Impact Analysis Results Break Failure Potential Inspections CDFlmpact LERF Impact Location (5)

DMs Rank (4)

SXI (2)

RIS_B (3)

Delta w/POD w/oPOD w/POD wlo POD LOCA IT Medium 0

4 4

-2. 16E-09

-1.20E-09

-2.16E-10

-1.20E-10 PLOCA IT Medium 0

1 1

-1.BOE-11

-1.00E-11

-1.BOE-12

-1.00E-12 ILOCA IT Medium 0

2 2

-3.60E-11

-2.00E-11

-3.60E-12

-2.00E-12 LOCA None Low 0

1 1

-1.50E-11

-1.50E-11

-1.50E-12

-1.50E-12 PLOCA None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO SEAL None Low 0

2 2

-1.00E-12

-1.00E-12

-1.00E-13

-1.00E-13 lSEAL None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO LSS Assume Medium 22 0

-22 4.40E-09 4.40E-09 4.40E-10 4.40E-10 2.17E-09 3.15E-09 2.17E-10 3.15E-10 SLB None Low 12 9

-3 1.50E-12 1.50E-12 1.50E-13 1.50E-13 SLB None Low 0

16 16

-B.OOE-12

-B.OOE-12

-B.OOE-13

-S.OOE-13 LSS Assume Medium 2

0

-2 2.00E-11 2.00E-11 2.00E-12 2.00E-12 1.20E-11 1.20E-11 1.20E-12 1.20E-12 LOCA PWSCC Medium 3

3 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO LOCA TASCS Medium 16 5

-11 1.S0E-10 3.30E-09 1.S0E-11 3.30E-10 LOCA TASCS, IT Medium 9

S

-1

-2.70E-09 3.00E-10

-2.70E-10 3.00E-11 LOCA IT Medium 6

6 0

-2.16E-09 O.OOE+OO

-2.16E-10 O.OOE+OO LOCA None Low 60 7

-53 7.95E-10 7.95E-10 7.95E-11 7.95E-11 PLOCA None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO

-3.89E-09 4.40E*09

• 3.89E*10 4.40E-10 PSDC TASCS Medium 1

4 3

-6.S0E-11

-3.00E-11

-S.SOE-12

-3.00E-12 PPSDC TASCS Medium 2

0

-2 1.20E-11 2.00E-11 1.20E-12 2.00E-12 PLOCA IT Medium 10 10 0

-1.20E-10 O.OOE+OO

-1.20E-11 O.OOE+OO PPLOCA IT Medium B

5

-3

-4.20E-11 3.00E-11

-4.20E-12 3.00E-12 PLOCA IT,IGSCC Medium 2

1

-1 1.00E-11 1.00E-11 1.00E-12 1.00E-12 PLOCA None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO PSDC None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO PPSDC None Low 4

0

-4 2.00E-12 2.00E-12 2.00E-13 2.00E-13 LSS AHA Assume Medium 23 0

-23 4.S0E-09 4.S0E-09 4.S0E-10 4.S0E-10 4.40E-09 4.63E-09 4.40E-10 4.63E.10 i

E1-33

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table 3.4-1a: FNP-1Risk Impact Analysis Results Safety Break Failure Potential Inspections CDF Impact LERF Impact System (1)

Significance Location (5)

DMs Rank (4)

SXI (2)

RIS_B (3)

Delta wlPOD wlo POD wlPOD wlo POD SI High PLOCA IGSCC Medium 3

2

-1 1.00E-11 1.00E-11 1.00E-12 1.00E-12 SI High PLOCA n

Medium 2

3 1

-4.20E-11

-1.00E-11

-4.20E-12

-1.00E-12

• SI High PLOCA n,IGSCC Medium 1

1 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO SI High PLOCA None Low 2

16 14

-7.00E-12

-7.00E-12

-7.00E-13

-7.00E-13 I SI Low LSS Assume Medium 36 0

-36 3.60E-10 3.60E-10 3.60E-11 3.60E-11 SI Total 3.21E-10 3.53E-10 3.21E-11 3.53E-11 i

AFWTotal Low LSS Assume Medium 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO I

CS Total Low LSS Assume Medium 12 0

-12 1.20E-10 1.20E-10 1.20E-11 1.20E-11 Grand Total 236 106

-130 3.14E-09 1.27E-08 3.14E-10 1.27E-09 I

Notes

1. Systems are described in Table 3.1 a (Unit 1) and Table 3.1 b (Unit 2).

2. Only those ASME Section XI Code inspection locations that received a volumetric 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-112657.

3. Only those RIS_B inspection locations that receive a volumetric examination are included in the count. Locations subjected to VT2 only are not credited in the count for risk impact assessment (there are none for Farley).

4. The failure potential rank for high safety significant (HSS) locations is 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")

5. The "LSS" designation is used to identify those Code Class 2 locations that are not HSS because they do not meet any of the five HSS criteria of Section 2(a) of N-716 (e.g., not part of the BER scope).

E1-34

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-AL T -12, VERSION 2.0

~~~~~

Table 3.4-1 b: FNP-2 Risk Impact Analysis Results

~~~~~~-~~~-

Failure Potential Inspections CDF Impact LERF Impact Safety Break System (1)

Significance Location (5)

DMs Rank (4)

SXI (2)

RIS_B (3)

Delta wlPOD wlo POD wlPOD w/o POD CVCS High LOCA IT Medium 0

3 3

-1.62E-09

-9.00E-10

-1.62E-10

-9.00E-11 CVCS High PLOCA IT Medium 0

1 1

-1.80E-11

-1.00E-ll

-1.80E-12

-1.00E-12

~~~-~~~

~ ~~~~ ~~~~~

~~~-~~~

CVCS High I LOCA IT Medium 0

2 2

-3.60E-11

-2.00E-11

-3.60E-12

-2.00E-12 CVCS High LOCA lASCS,IT Medium 0

3 3

-1.62E-09

-9.00E-l0

-1.62E-10

-9.00E-11 CVCS High PLOCA None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO CVCS High SEAL None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO CVCS High ISEAL None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO CVCS Low LSS Assume Medium 25 0

-25 5.00E-09 5.00E-09 5.00E-10 5.00E-l0 CVCSTotal 1.71E-09 3.17E-09 1.71E-10 3.17E-10 FW Total High SLB None Low 8

9 1

-5.00E-13

-5.00E-13

-5.00E-14

-5.00E-14 MS High SLB None Low 7

14 7

-3.50E-12

-3.50E-12

-3.50E-13

-3.50E-13 MS Low LSS Assume Medium 4

0

-4 4.00E-11 4.00E-11 4.00E-12 4.00E-12 MS Total 3.65E-11 3.65E-11 3.65E-12 3.65E-12 RC High LOCA PWSCC Medium 3

3 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RC High LOCA lASCS Medium 13 5

-S

-3.60E-10 2.40E-09

-3.60E-11 2.40E-10 RC High LOCA lASCS, IT Medium 15 6

-9

-5.40E-l0 2.70E-09

-5.40E-11 2.70E-10 RC High LOCA IT Medium 5

5 0

-1.S0E-09 O.OOE+OO

-1.80E-10 O.OOE+OO RC High LOCA None Low 63 11

-52 7.S0E*10 7.S0E-10 7.S0E-11 7.80E-ll RC High PLOCA None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RC Total

-1.92E-09 5.88E-09

-1.92E-10 5.88E-10 RHR High PSDC lASCS Medium 0

2 2

-3.60E-11

-2.00E-11

-3.60E-12

-2.00E-12 RHR High PLOCA IT Medium 5

9 4

-1.32E-10

-4.00E-11

-1.32E-l1

-4.00E-12 RHR High PPLOCA IT Medium 10 6

-4

-4.S0E-11 4.00E-l1

-4.80E-12 4.00E-12 RHR High PLOCA IT,IGSCC Medium 4

2

-2 2.00E-11 2.00E-11 2.00E-12 2.00E-12 RHR High PLOCA None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RHR High PSDC None Low 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RHR High PPSDC None Low 3

0

-3 1.50E-12 1.50E-12 1.50E-13 1.50E-13 RHR Low LSS RHR Assume Medium 25 0

-25 5.00E-09 5.00E*09 5.00E-l0 5.00E-10 RHRTotal 4.81E-09 5.00E-09 4.81E-10 5.00E-10

~~~~~~~

~ ~ ~~~~-~~ ~ ~

E1-35

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table 3.4-1b: FNP-2 Risk Impact Analysis Results Failure Potential Inspections CDF Impact LERF Impact Safety Break System (1)

Significance Location (5)

OMs Rank (4)

SXI (2)

RIS_B (3)

Delta wIPOD w/o POD wIPOD w/o POD 81 High PLOCA IG8CC Medium 1

2 1

-l.OOE-ll

-l.OOE-11

-l.OOE-12

-I.OOE-12 81 High PLOCA n

Medium 0

1 1

-1.80E-ll

-l.OOE-II

-1.80E-12

-I.OOE-12 81 High PLOCA n,IG8CC Medium 0

1 1

-J.OOE-II

-I.OOE-ll

-I.OOE-12

-I.OOE-12 81 High PLOCA None Low 5

17 12

-6.00E-12

-6.00E-12

-6.00E-I3

-6.00E-13 81 Low L88 Assume Medium 34 0

-34 3.40E-tO 3AOE-tO 3.40E-ll 3.40E-ll SI Total 2.96E*IO 3.04E*IO 2.96E*ll 3.04E*ll AFWTotal Low L88 Assume Medium 0

0 0

O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO 1----

CS Total Low L8S Assume Medium 13 0

-13 l.30E-IO 1.30E-IO 1.30E*ll 1.30E-ll Grand Total 243 102

-141 5.05E-09 1.45E-08 5.05E-10 1.45E-09 Notes

1. Systems are described in Table 3.1a (Unit 1) and Table 3.1b (Unit 2).

2. Only those ASME Section XI Code inspection locations that received a volumetric 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-112657.

3. Only those RIS_B inspection locations that receive a volumetric examination are included in the count. Locations subjected to VT2 only are not credited in the count for risk impact assessment (there are none for Farley).

4. 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")

5. The "LSS" designation in Table 3.4-1a (Unit 1) and Table 3.4-1b (Unit 2) is used to identify those Code Class 2 locations that are not HSS because they do not meet any of the five HSS criteria of Section 2(a) of N-716 (e.g., not part of the BER scope).

E1-36

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a}(3}(i)

FNP-ISI-ALT-12, VERSION 2.0 Table 4a: FNP-1 Inspection Location Selections Comparison safety Code Case N716 Failure Potential Code Section XI System Significance Weld (2)

Break Location Category (1)

Rank Count High Low OMs (4)

Vol Surface RIS_B Other (3)

CVCS

~

LOCA IT Medium B..J 4

0 0

4 NA CVCS

~

PLOCA IT Medium B-J 7

0 0

1 NA CVCS

~

ILOCA IT Medium B-J 11 0

1 2

NA CVCS

~

LOCA None Low B..J 1

0 0

1 NA CVCS

~

PLOCA None Low B-J 30 0

2 0

NA CVCS

~

SEAL None Low B..J 19 0

2 2

NA CVCS

~

ISEAL None Low B-J 22 0

12 0

NA CVCS

~

LSS N/A Assume Medium C-F-1 312 22 1

0 NA

~

FW SLB None Low C-F-2 81 12 1

9 NA

~

MS SLB None Low C-F-2, HELB 151 0

0 16 NA MS

~

LSS N/A Assume Medium C-F-2 46 2

0 0

NA RC

~

LOCA PWSCC Medium B-F 3

3 0

3 NA RC

~

LOCA TASCS Medium B-J 19 16 0

5 NA RC

~

LOCA TASCS, IT Medium B-J 23 9

5 8

NA RC

~

LOCA IT Medium B-J 17 6

2 6

NA

~

RC LOCA None Low B-F, B-J 212 60 20 7

NA RC

~

PLOCA None Low B-J 10 0

3 0

NA RHR

~

PSDC TASCS Medium B..J 10 1

0 4

NA RHR

~

PPSDC TASCS Medium C-F-1 5

2 0

0 NA

~

RHR PLOCA IT Medium B-J 77 10 0

10 NA

~

RHR PPLOCA IT Medium C-F-1 45 8

0 5

NA RHR

~

PLOCA IT,IGSCC Medium B..J 4

2 0

1 NA

~

RHR PLOCA None Low B..J 17 0

9 0

NA RHR

~

PSDC None Low B..J 8

0 0

0 NA

~

RHR PPSDC None Low C-F-1 25 4

0 0

NA RHR

~

LSS RHR NlA Assume Medium C-F-1 363 23 0

0 NA SI PLOCA IGSCC Medium B..J 8

3 0

2 NA

~

SI

~

PLOCA IT Medium B-J 12 2

0 3

NA E1-37

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0

~"--

~

Table 4a: FNP-1 Inspection Location Selections Comparison,

Safety Code Case N716 Failure Potential Code Section XI System Significance Weld (2)

Break Location Category (1)

Rank Count High Low OMs (4)

Vol Surface RIS_B Other (3)

SI 01' PLOCA TT,IGSCC Medium 8-J 4

1 2

1 NA SI 01' PLOCA None Low 8-J 191 2

30 16 NA SI 01' LSS N/A Assume Medium C-F-1 752 36 29 0

NA AFW 01' LSS N/A Assume Medium C-F-2 12 0

0 0

NA 01' LSS N/A Assume Medium C-F-1 132 12 0

0 NA

,.- C~.. -

1..

Notes

1. Systems are described in Table 3.1a (Unit 1) and Table 3.1 b (Unit 2).

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. This option is not applicable for the FNP RIS_B application. The "Other" column has been retained in this table solely for uniformity purposes with other RIS_B application template submittals and to indicate when RIS_B selections will receive a VT-2 examination (these are not credited in risk impact assessment), but there are no such cases for Farley.

3. The failure potential rank for high safety significant (HSS) locations is 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 (Le., "Assume Medium").

4. Category "HELB" refers to non safety-related high energy line break piping (also called break exclusion region piping).

E1-38

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a}(3}(i}

FNP-ISI-ALT-12, VERSION 2.0 Table 4b: FNP-2 Inspection Location Selection Comparison System (1)

Safety Significance High Low Break Location Failure Potential Rank OMs (3)

Code Category (4)

Weld Count Section XI Vol Surface Code Case N716 Other RIS_B (2)

CVCS

./

LOCA n

Medium 8-J 3

0 1

3 NA CVCS

./

PLOCA n

Medium 8-J 7

0 0

1 NA CVCS

./

I LOCA n

Medium 8-J 11 0

0 2

NA CVCS

./

LOCA TASCS, n Medium 8-J 3

0 1

3 NA CVCS

./

PLOCA None Low 8-J 26 0

5 0

NA CVCS

./

SEAL None Low 8-J 23 0

6 0

NA CVCS

./

ISEAL None Low 8-J 10 0

0 0

NA CVCS

./

LSS NlA Assume Medium C*F-1 297 25 2

0 NA FW

./

SL8 None Low C-F-2 81 8

1 9

NA MS

./

SLB None Low C*F-2, HEL8 138 7

0 14 NA MS

./

LSS NlA Assume Medium C-F-2 51 4

0 0

NA RC

./

LOCA PWSCC Medium 8-F 3

3 0

3 NA RC

./

LOCA TASCS Medium 8-J 19 13 0

5 NA RC

./

LOCA TASCS, n Medium B*J 21 15 3

6 NA RC

./

LOCA n

Medium 8-J 18 5

1 5

NA RC

./

LOCA None Low B-F,8-J 220 63 12 11 NA RC

./

PLOCA None Low 8-J 10 0

5 0

NA RHR

./

PSDC TASCS Medium 8-J 6

0 0

2 NA RHR

./

PLOCA n

Medium 8-J 69 5

1 9

NA RHR

./

PPLOCA n

Medium C-F-1 48 10 0

6 NA RHR

./

PLOCA n,IGSCC Medium 8-J 6

4 0

2 NA RHR

./

PLOCA None Low B-J 16 0

8 0

NA RHR

./

PSDC None Low 8-J 20 0

0 0

NA RHR

./

PPSDC None Low C-F-1 25 3

0 0

NA RHR

./

LSS RHR NlA Assume Medium C*F*1 361 25 0

0 NA SI

./

PLOCA IGSCC Medium 8-J 8

1 0

2 NA SI

./

PLOCA n

Medium B-J 1

0 1

1 NA SI

./

PLOCA n,IGSCC Medium 8-J 2

0 0

1 NA E1-39

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-AL T-12, VERSION 2.0

-~~". ------

Table 4b: FNP-2 Inspection Location Selection Comparison I

safety Significance Failure Potential Code Section XI Code Case N716 System Break Weld I

Category (1)

Location Count I

Rank Other High Low OMs (4)

Vol Surface RIS_B (3)

(2)

SI

./

PLOCA None Low B-J 190 5

32 17 NA SI

./

LSS N/A Assume Medium C-F-1 576 34 18 0

NA AFW

./

LSS N/A Assume Medium C-F-2 9

0 0

0 NA I

CS

./

LSS N/A Assume Medium C-F-1 141 13 0

0 NA

---.J Notes

1. Systems are described in Table 3.1 a (Unit 1) and Table 3.1 b (Unit 2).

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. This option is not applicable for the Farley RIS_B application. The "Other" column has been retained in this table solely for uniformity purposes with other RIS_B application template submittals and to indicate when RIS_B selections will receive a VT-2 examination (these are not credited in risk impact assessment), but there are no such cases for Farley.

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 (Le., "Assume Medium").

4. Category "HELB" refers to non safety-related high energy line break piping (also called break exclusion region piping).

E1-40

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Attachment A to FNP N-716 Template Consideration of the Adequacy of Probabilistic Risk Assessment Model for Application of Code Case N716 A-1

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(I)

FNP-ISI-ALT-12, VERSION 2.0 Introduction SNC employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating SNC nuclear generation sites. This approach includes both a proceduralized PRA maintenance and an update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the FNP PRA.

PRA Maintenance and Update The SNC risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated units. The SNC risk management process also delineates the responsibilities and guidelines for updating the full power internal events PRA models at aU operating SNC nuclear generation sites. The overall SNC risk management program defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the FNP PRA model has been updated according to the requirements defined in the SNC risk management process:

Pertinent modifications to the physical plant (i.e. those potentially affecting the Base Line PRA (BL-PRA) models, calculated core damage frequencies (CDFs), or large early release frequencies (LERFs) to a significant degree) shall be reviewed to determine the scope and necessity of a revision to the baseline model within six months following either a periodic refueling outage on Unit 1 or a specific major plant modification occurring outside a refueling outage. The BL-PRAs should be updated as necessary in accordance with a schedule approved by the PRA Manager following the scoping review. Upon completion of the lead Unit's BL-PRA, the other Unit's BL-PRA will be regenerated by modification of the updated BL-PRAs to account for Unit differences which significantly impact the results.

Pertinent modifications to plant procedures and Technical Specifications shall be reviewed annually for changes which are of statistical significance to the results of the BL-PRA and those changes documented. Reliability data, failure data, initiating events frequency data, human reliability data, and other such PRA inputs shall be reviewed approximately every three years for statistical significance to the results of the BL-PRAs. Following the tri-annual review, the BL-PRAs shall be updated to account for the significant changes to these two categories of PRA inputs in accordance with an approved schedule.

BL-PRAs shall be updated to reflect germane changes in methodology, phenomenology, and regulation as judged to be prudent or as required by regulation.

A-2

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 In addition to these activities, SNC risk management procedures [2, 3, 4, 5, and 6] provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes:

• Documentation of the PRA model, PRA products, and bases documents.

• The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.

• Guidelines for updating the full power, internal events PRA modelS for SNC nuclear generation sites.

• Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modi'fications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (1 OCFR50.65 (a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximate 3-year cycle; however, longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant.

Table A-1 provides historical details of the FNP PRA model changes/revisions from the Individual Plant Evaluation (IPE) to the current model.

Consistency with Applicable ASME PRA Standard Requirements Previous Peer Reviews and Self Assessments for the Farley PRA Model Several assessments of the technical capability have been made for the Farley Nuclear Plant PRA models. These assessments are shown as follows and are further discussed in the paragraphs below.

• An independent PRA peer review was conducted under the auspices of the Westinghouse Owners Group (WOG) in 2001 [7], following the Industry PRA Peer Review process [8J. This peer review included an assessment of the PRA model maintenance and update process.

In 2005, a gap analysis [9] was performed against the available version of the ASME PRA Standard [10] and Regulatory Guide 1.200.

RG 1.200 PRA Peer Review for the Farley PRA Model against the ASME PRA Standard Requirements A complete Peer Review of the Farley Nuclear Power Plant Probabilistic Risk Assessment (PRA) model against the requirements of Section 2 of the ASME/American Nuclear Society (ANS) Combined PRA Standard [11] and the guidance provided by A-3

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-AL T -12, VERSION 2.0 Regulatory Guide (RG) 1.200, Revision 2 [12] (for Internal Events only) was completed in March 2010 for the Unit 1 model. This peer review was performed using the process defined in Nuclear Energy Institute (NEI) 05-04 [13].

The final report is under final review. A summary of the results of the Peer Review for the Internal Events is described below:

The Combined PRA Standard [11] contains a total 326 numbered supporting requirements (SR) in fourteen technical elements and the configuration control element.

Of the 326 SRs, eight were determined to be not applicable to the Farley PRA. These are: AS-B4, SY -A9, HR-C3, HR-D5, DA-C5, DA-C8, DA-C15, and DA-D2. Therefore, 318 SRs apply to Farley. Among the 318 applicable SRs, 92% of the SRs met Capability Category II or higher, as shown below.

Capability Category (CC)

NO.ofSRs

% of Total Applicable SRs Met CCII 30 9%

Met CCIII 12 4%

Met CCI/II 13 4%

Met CCIIIIII 24 7%

Met CCIIIIIIII 213 65%

Met CCI but not CCII 9

3%

Did not meet either CCI or CCII ("not-mer' SRs) 17 5%

SR Not Applicable 8

3%

Total 326 100%

Resolution of Findings from the RG 1.200, Revision 2 PRA Peer Review Table A-2 lists the 17 CDF related "not-mer' SRs and five CDF related SRs that met CCI requirements but not CCII requirements. The associated Facts and Observations (F&Os) and the resolution of each F&O are shown in Table A-2 for each of the 22 related SRs. All 22 SRs have been closed as shown in the "Status of Resolution" column in Table A-2.

Table A-3lists four LERF related SRs that met CCI requirements but not CCII requirements. A summary of the Peer Team assessment for each SR is provided. None of the four SRs have been closed on the basis that not meeting CCII requirements has a minimal impact on the risk-insights provided for this RIS_B application. The implementation of the assessment would result in a conservative estimation of the LERF figure of merit which would be conservative for this RIS_B application.

The following should be noted:

• All the F&Os associated with the "met" CCII SRs were reviewed to assess their potential impact on this application. One F&O, DA-C14-01, was judged to need discussion for this RIS_B application. This F&O observes that, although the generic data had been updated (using NUREG/CR-6928 [15]) and the updated data was further updated by the plant specific data, the utilized plant specific data was not the I

J i

A-4

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-AL T -12, VERSION 2.0 most up-to-date data (plant specific data covered from 1984 to 2001). This F&O will be closed by executing the recommended action in the F&O. However, the PRA results provided in support of this RIS_B application is not based on a model where the F&O is closed. It is judged that the impact on risk insights for this RIS_B application is minimal because 1) aU Data (DA) element SRs have been met at cell or higher and 2) the eDF and LERF figures of merit are not expected to substantially change by executing the F&O recommendation.

• The Unit 1 internal flooding eDF and LERF results are used for both units. Although, there is no independent Unit 2 internal flooding model, it is judged that for this RIS_B application, the Unit 1 model is adequate to represent Unit 2 flooding risk based on the following:

To ensure applicability, the characteristics of the Unit 2 flood areas were compared to Unit 1. It is recognized that the most important plant features that may affect the internal flooding risk involve the locations and sizes of various flood sources, the locations of the accident mitigation equipment, and the plant configurations that may affect flood propagation paths. The major plant differences were evaluated based on information collected from the Unit 1 plant walk-down and a subsequent review of the Unit 2 plant design information related to flooding. This review did not identify any major differences that would adversely impact the flooding analysis results developed for Unit 1.

After the Unit 2 internal flooding model is completed, if there are any segments identified with a contribution to eDF greater than 1 E-06 (1 E-07 for LERF) they will be added to the High Safety-Significant scope during the subsequent periodic update.

Conclusion The resolutions to the March 2010 PWROG peer review results shown in Table A-2 demonstrate that while there are some gaps related to documentation, the technical adequacy of the Farley PRA model is robust and the model is suitable for use in support of this RIS_B Inservice Inspection application.

A-5

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-AL T -12, VERSION 2.0 Attachment A References

1. Generation and Maintenance of Probabilistic Risk Assessment Models and Associated Updates, NL-PRA-001 Version 3.0, SNC, 2008.

2. Col/ection, Evaluation, and Documentation of Baseline PRA Update Information, NL PRA-002 Version 2.0, SNC, 2008.

3. Structures, Systems, and Components Risk Significance Evaluation Procedure for Maintenance Rule, NL-PRA-003 Version 2.0, SNC, 2008.

4. PRA Calculations - Preparation and Revision, NL-PRA-008 Version 4.0, SNC, 2009.

5. PRA Calculation Administration, NL-PRA-009 Version 2.0, SNC, 2008.

6. PRA Software and Application Control, NL-PRA-010 Version 2.0, SNC, 2008.

7. Westinghouse Owners Group Peer Review Final Report, Westinghouse, 2002.

8. Probabilistic Risk Assessment (PRA) Peer Review Process Guidance, NEI-00-02, Rev.

A3,2000.

9. Gap Analysis of the Farley PRA, ERIN, 2005.

10. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-S-2002, 2002 and Addenda to Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sa-2003, 2003.

11. ASMEI ANS RA-Sa-2009, Addenda to ASMEIANS RA-S-2008 Standard for Level 1/

Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, American SOCiety of Mechanical Engineers, 2009.

12. U.S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 2, 2008.

13. Process for Performing Internal Events PRA Peer Reviews Using the ASMEIANS PRA Standard, NEI 05-04, Revision 2, 2008.

14. Farley Nuclear Plant Internal Flooding Probabilistic Risk Assessment, ASS 1762171-R 001, Revision 3, ASS Consulting for SNC, 2009.

15. Industry-Average Performance for Components and Initiating Events at U.S.

Commercial Nuclear Power Plants, NUREGICR 6928, Idaho National Laboratory for the US NRC, 2007.

A-6

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table A-1 History of FNP Units 1 and 2 PRA Model Changes CDF and LERF by Revision i

Unit 1 (Unit 2) CDF Unit 1 (Unit 2) LERF Revision Major Changes from the Previous Revision per Reactor Year per Reactor Year I

1.30E-04 4.47E-07 N/A o(lPE) i 1 (12/1997) 7.63E-05 (7.49E-05) 6.29E-07 (6.29E-07)

Converted the model from a large event tree to a linked fault tree using CAFT A.

• Developed unit-specific models for Unit 1 and Unit 2 to support EOOS.

• Incorporated plant design changes that were completed since the IPE.

2 (05/1998) 8.72E-05 (8.65E-05) 5.50E-07 (5.50E-07)

Revised RCP seal LOCA modeling.

• Revised SSO modeling.

• Revised ATWS modeling to ensure proper application of UET.

• Changed the mission time for AFW to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for general transient initiating events.

• Refined the modeling of swing components to ensure all failure modes are addressed where train re-alignment is credited.

• Revised LERF modeling to use the LERF definition developed by the WaG Risk Based Technologies Working Group.

• Incorporated plant design changes that were completed since the previous revision.

A-7

I i

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table A-1 History of FNP Units 1 and 2 PRA Model Changes CDF and LERF by Revision Revision 3 (OS/1999) 4 (05/2000)

S (11/2001)

Unit 1 (Unit 2) CDF per Reactor Year B.52E-05 (BA5E-05) 5.S7E-05 (B.91 E-OS) 3.SBE-05 (S.S1 E-OS)

Unit 1 (Unit 2) LERF per Reactor Year 4.S0E-07 (4.50E-07) 4A7E-07 (4.53E-07) 4.19E-07 (4.2BE-07)

Major Changes from the Previous Revision

• Updated component reliability data to include plant experience through 12131/97.

• Updated initiating event frequencies using NUREG/CR-5750 generic data and plant experience through 12/31197.

• Incorporated design changes for the instrument air system.

• Expanded modeling of the service water intake structure and turbine building DC systems to include alternate battery chargers and battery banks to support EOOS assessments.

• Revised SBO modeling to include SBO sequences in the fault tree rather than adding offsite power recovery during post processing.

• Revised the A TWT modeling to ensure that the proper success criteria for AFW are applied to the various cases.

• Added Very Small LOCA event tree.

• Revised HRA for events where procedures had changed.

• Updated the flooding analysis for the Service Water Intake Structure and CCW pump/heat exchanger rooms.

• Added the System Model for emergency air compressors for atmospheric relief valves and AFW pumps.

• Added Unit 2 Service Water lube and cooling booster pumps.

• Incorporated plant design changes that were completed since the previous revision.

• Revised the model to address WOG Peer Review comments.

• Incorporated plant design changes that were completed since the previous revision.

A-8 I

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-AL T -12, VERSION 2.0 Table A-1 History of FNP Units 1 and 2 PRA Model Changes CDF and LERF by Revision Unit 1 (Unit 2) CDF Unit 1 (Unit 2) LERF Revision Major Changes from the Previous Revision per Reactor Year per Reactor Year 6 (03/2005) 3.79E-05 (3.32E-05) 4.94E-07 (4.92E-07)

• Incorporated plant design changes through December 2004.

• Revised the SGTR Event Tree.

• Updated the CCF Analysis.

• Updated the HRA.

• Updated component reliability, unavailability and initiating event data with plant experience through December 2001.

7 (06/2006) 2.35E-05 (2.03E-05) 5.11 E-07 (5.06E-07)

• Revised SW success criteria for diesel generator support.

• Revised the LERF model to incorporate induced SGTR.

• Revised modeling of maintenance on CCW and Charging pumps to incorporate the current plant practice of minimizing at-power train maintenance outages utilizing swing pumps.

• Revised the pipe rupture frequencies for internal flooding per EPRI TR-1013141.

• Revised the modeling of SW Pump 20 to reflect design changes completed following Revision 6.

• Revised the event tree for Secondary Side I

Break initiating events.

I A-9

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(I)

FNP-ISI-ALT-12, VERSION 2.0 Table A-1 History of FNP Units 1 and 2 PRA Model Changes CDF and LERF by Revision Unit 1 (Unit 2) CDF Unit 1 (Unit 2) LERF Revision Major Changes from the Previous Revision per Reactor Year per Reactor Year 8 (06/2008) 1.87E-05(1.54E-05) 5.05E-07(5.00E-07)

• Incorporated removal of the final Unit 2 SW Booster Pump.

• Revised requirements for Unit 2 SW Pumps such that the cyclone separator is no longer required.

• Revised operating alignment in which RCP Seal Injection and RCP Thermal Barrier Cooling are supplied by opposite trains.

This eliminates a single train CCW, SW, or Electrical Bus initiating event for causing a total loss of RCP Seal cooling.

• 20% CDF reduction for Unit 1 and 24%

CDF reduction for Unit 2.

9 (10/2010) 2.28E-05 (Unit 2 in 1.40E-07 (Unit 2 in Upgraded per RG 1.200 Rev 2.

progress) progress)

• Incorporated RCP Shutdown Seals.

• Incorporated Internal Flooding.

• Restructured Event Trees.

A-10

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table A-2 Resolution of the CDF Related Farley PRA Peer Review F&Os for "Not Met" Capability Category II SRs Peer Review Technical F&O#

Finding Resolution The Status of Resolution by SNC Element IE-AS IE-A5-01 The system notebooks look at the Add a systematic review A systematic review of the Farley safety and (SR CCI) impact of the identified initiators on that of the safety and non non-safety systems was performed that metCCI system. However, a system by system safety systems that resulted in the development of a Table C-1 review might identify additional plant could cause a plant "Farley Initiating Event Identification Analysis" specific initiators, particularly associated scram to verify that no which is documented as part of the Farley with transformers, buses, etc.

additional initiators are Initiating Event Notebook. This table lists each needed.

Farley system ordered by a system group identifier, system 10, system description, impact of system loss and treatment of system loss in Farley PRA. The treatment of system loss" addressed specifically whether the loss of a system would result in an initiating event and how the initiating event was grouped. This finding is considered closed pending incorporation into Initiating Events notebook.

IE-A9 IE-A9-01 A plant-specific review of potential Review significant non-A search was performed using the Condition (SR CCI) precursor events, such as intake scram events at the Reports database for significant non-scram met CCI structure clogging and others has not plant to determine if any events. A comparison of the results was made been performed for Farley.

precursors exist.

to Farley's initiating events list. No new initiating event precursors to plant trips were found. This finding is considered closed pending incorporation into the Initiating Events notebook.

A-11

I ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-AL T -12, VERSION 2.0

---

~~

Peer Review Technical Element IE-B3 (SR CCI) met CCI IE-C5 (SR CC-I/II) not met AS-C2 (SR CC-I/IIIIII) not met Table A-2 Resolution of the CDF Related Farley PRA Peer Review F&Os for "Not Met" Capability Category II SRs F&O#

Finding Resolution The Status of Resolution by SNC j

IE-B1-01 Several cases were noted where Include the impact of the Table C-1 "Farley Initiating Event Identification grouping in the initiating event initiator on the PSA Analysis" was created which is documented in document is unclear or incorrect.

systems in the model.

the Farley Initiating Event Notebook. This table Therefore, additional documentation is lists each Farley system ordered by a system needed to verify that the event grouping group identifier, system 10, system description, is clear and can be easily traced to the impact of system loss and treatment of system plant impact.

loss in Farley PRA. The treatment of "system loss" addressed specifically whether the loss of a system would result in an initiating event and how the initiating event was grouped. This finding is considered closed pending incorporation into Initiating Events notebook.

IE-C5-01 Farley did not weigh the initiating event Modify the initiating The adjustment has been done as part of the frequencies by the fraction of time the event frequency to quantification. This finding is considered plant is at power.

address plant closed pending incorporation into Initiating availability.

Events notebook.

AS-C2-02 The Farley Accident Sequence Add initiating events The Accident Sequence notebook was revised notebook provides discussions of the

%LOSPF and %LOSPG to correctly reference the loss of bus initiating examples indicated in the SA. Improve to Table 2.6-1. Correct events. The descriptions of the %LOSSACF the level of documentation.

the descriptions of and %LOSSACG events in Table 2.6-4 were initiating events not changed because they are correct.

%LOSSACF and Instead, the descriptions for those events were

%LOSSACG in Table corrected in Table 2.6-1 and events %LOSPF 2.6-4.

and %LOSPG were added to Table 2.6-1.

Documentation was revised. This finding is considered closed.

--~

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ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table A-2 Resolution of the CDF Related Farley PRA Peer Review F&Os for "Not Met" Capability Category II SRs Peer Review Technical F&O#

Element SY-A6 SY-A9-01 (SR CC-I/II/III) not met SY-C1 SY-C1-01 (SR CC-I/II/III) not met HR-D2 HR-D2-01 (SR CCI)

HR-D2-02 metCCI HR-G1 HR-G1-01 (SR CCI) metCCI Finding The system boundary as defined in the system notebook does not match up to the fault tree. Examples include room cooling is defined in the notebook as system dependency but in the model room cooling is included as part of the system designation.

The system notebooks documentation on test and maintenance for several systems is incorrect and references old or incorrect documents.

Detailed Human Failure Event (HFE) assessments are used for events that are not shown to be directly applicable to the analysis performed. Also, the screening values used for pre-Human Reliability Analyses (HRAs) are significantly lower than the ASEP values without justification of the values used.

In general detailed analysis is done for most post HRA events. However, the most important HRAs showing up in the cutset have not been performed on a detailed analysis.

Resolution Review the component boundary definitions to ensure that they are sufficiently detailed to identify exactly what is included within each component and that are consistent from the model to the system notebooks.

Correct the system notebook's references for test and maintenance information.

Perform detailed analysis on all events to verify the applicability used or use screening values for those events not explicitly analyzed with a detailed analysis.

Develop HRAs for the referenced 2 events and include in the HRA calculation.

The Status of Resolution by SNC The system notebooks were reviewed and modified as needed to reflect the boundary of the system as shown in the model. The support system sections were reviewed and corrected as needed to reflect the support systems as modeled. This finding is considered closed pending final updates to the system notebooks.

This is a documentation issue. The references were corrected. This finding is considered closed pending final updates to the system notebooks.

A revision to Table 8-2 of the HRA notebook has been prepared providing a more detailed explanation of the approach used. The pre-initiator approach relies on detailed THERP assessments that are mapped to similar HFEs.

HR-D2 does not preclude using detailed THERP analyses for all HFEs. These findings are considered closed pending update of the HRA notebook.

Included the events in the HRA calculator file using the values found in NUREG CR5500 and WCAP-15831. The finding is considered closed pending update of the HRA notebook.

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ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0

Table A-2 Resolution of the CDF Related Farley PRA Peer Review F&Os for "Not Met" Capability Category II SRs Peer Review Technical F&O#

Element HR-G7 HR-G7-01 aU-A5 HR-G7-02 aU-C2 (SR CCIIII/II) not met HR-13 HR-13-01 (SR CCIIII/II) not met IFEV-B3 IFEV-B3-01 IFSO-B3 IFPP-B3 (SR CCIIII/II) not met IFPP-B2 I FPP-B2-02 (SR CCIIII/II)

IFPP-B2-03 not met Finding The multiple human action analysis described in Appendix C does not appear to be used in the quantification.

Attachment C to the HRA notebook performs the dependency assessment, but the dependency factors are based upon 2004 HRA values. The multiplication factors in the rule file are to be based upon current HRA. The top HRA cutset combinations in the au notebook are not addressed in the HRA dependency analysis.

Sources of uncertainty are not included in the HRA calculation similar to other Farley documentation.

The Farley PRA flooding analysis indicates that sources of uncertainty were not documented because of the low contribution to CDF and LERF from flooding. Although this is true, the SR requires that a discussion of uncertainty be provided.

Internal flooding notebook provides the process and selection result of flood areas partitioning. However, there is no description about the reason for eliminating areas from further analysis, except containment.

Resolution Explicitly evaluate the top HRA combinations in the dependency analysis. Update the HRA dependence evaluation to be consistent with industry practices.

Include a source of uncertainty in the HRA calculation.

Include a discussion of uncertainty and assumptions related to internal flooding issues including partitioning, initiating events, and flood sources.

Add information about the screened/eliminated areas and buildings in terms of internal flooding analysis.

The Status of Resolution by SNC An HRA Dependency Analysis was conducted and incorporated into the Revision 9 model quantification. This analysis will be incorporated into the HRA notebook as Appendix C. The finding is considered closed pending update of the HRA notebook.

A document was created to address HRA Uncertainty for the Farley model. It can be found as Attachment F in the HRA notebook.

The finding is considered closed pending update of the HRA notebook.

New text concerning uncertainty and assum ptions has been incorporated into the appropriate sections of the Flooding notebook.

The finding is considered closed pending update of the Flooding notebook.

New text concerning screened/eliminated areas and buildings has been incorporated into the appropriate sections of the Flooding notebook.

The finding is considered closed pending update of the Flooding notebook.

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ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0

-~~~" ------

Peer Review Technical Element IFQU-A7 (SR CCIIII/II) not met IFSN-A4 (SR CCI/IIIII) not met IFSN-B3 (SR CCIIII/II) not met QU-F1 (SR CCIIII/II) not met MU-B4 (SR CCI/II/II) not met Table A-2 Resolution of the CDF Related Farley PRA Peer Review F&Os for "Not Met" Capability Category II SRs F&O#

Finding Resolution The Status of Resolution by SNC IFQU-A7-01 Quantification of flooding event does Perform and provide An HRA Dependency Analysis was conducted not perform uncertainty analysis and uncertainty analysis and and incorporated into the Revision 9 model dependency analysis.

dependency analysis, quantification. This analysis will be even though the flood incorporated into the HRA notebook as risk is not significant.

Appendix C. The finding is considered closed pending update of the HRA notebook.

IFSN-A4-01 In the IF Notebook, there was extensive Add a table that New text has been incorporated into the discussion with respect to treatment of explicitly includes drain appropriate sections of the Flooding notebook.

drains, there was explicit evidence that capacities.

The finding is considered closed pending drains were considered as propagation update of the Flooding notebook.

paths for several flood scenarios.

However, no explicit estimation of drain capacities could be found.

IFSN-B3-01 There is no description about Include a section in the New text concerning uncertainty and uncertainty.

IF Notebook to discuss assumptions has been incorporated into the the IF assumptions and appropriate sections of the Flooding notebook.

sources of uncertainty.

The finding is considered closed pending update of the Flooding notebook.

QU-F1-01 The mutually exclusive logic was Update the Documentation reference has been updated.

generated by the procedure FNP-O-documentation to reflect The finding is considered closed pending ACP-52.1 but was not documented in the actual references.

update of the Quantification notebook.

the quantification notebook.

MU-B4-01 There is no reference to a peer review Revise either NL-PRA-These procedures are under revision.

for upgrades. Did not find a section 001 or NL-PRA-002 to which addressed upgrades (not explicitly require a peer updates) to the PRA specifically change review for PRA in software used.

upgrades A-15 i

ENCLOSURE 1 FARLEY NUCLEAR PLANT, UNITS 1 AND 2 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR SO.SSa(a)(3)(i)

FNP-ISI-ALT-12, VERSION 2.0 Table A-3 LERF Related Farley PRA Peer Review F&Os for Not-Met CAT II SRs Review Element Summary of Assessment LE-C2 (SR CCI) met CCI The Farley PRA LERF model relies largely on human error probabilities taken from the WCAP-16341-P.

are generic rather than plant-specific, they were derived as conservative estimates.

LE-C9 (SR CCI)

No credit is taken for either equipment operation or human actions in adverse environments.

met CCI Because the WCAP HEPs LE-C11 No credit was taken in the Farley PRA for equipment or operator actions impacted by containment failure. The WCAP-16341-P (SR CCI) methodology conservatively does not credit containment sprays for fission product scrubbing or pressure suppression for the met CCI containment failure.

LE-C12 (SR CCI) metCCI The LERF frequency calculated in the Farley PRA is so low that no review was performed to reduce LERF based on engineering analysis to support equipment operation or operator action after containment failure.

A-16