NEXTera EN ERGY - NRC

NEXTeraEN ERGY

November 7, 2011

Docket No. 50-443

SBK-L-1 1197

U. S. Nuclear Regulatory CommissionAttention: Document Control DeskOne White Flint North11555 Rockville PikeRockville, MD 20852

Seabrook Station

Request for Alternative to Use ASME Code Case N-716 to ImplementRisk-Informed Inservice Inspection Program

Pursuant to 10 CFR 50.55a(a)(3)(i), NextEra Energy Seabrook, LLC (NextEra) requestsauthorization to implement Risk-Informed / Safety Based Inservice Inspection (RISB ISI)alternative 3AR-1. This alternative will be used in lieu of the existing ASME Section XI CodeCategory B-F, B-J, C-F-I and C-F-2 requirements for examination of Class I and 2 piping welds.This alternative, which is described in Attachment I to this letter, has been developed inaccordance with Code Case N-716, "Alternative Piping Classification and ExaminationRequirements."

NextEra plans to implement the proposed alternative during the third ten-year inserviceinspection interval that began on August 19, 2010. To facilitate the NRC's review, thisalternative contains a template format modeled after previous submittals that the NRC hasapproved. It includes an evaluation of Probabilistic Risk Assessment (PRA) adequacy includinga gap analysis performed against Regulatory Guide 1.200. NextEra requests approval of theRISB ISI Program by August 31, 2012 to facilitate planning for the remainder of the firstinspection period.

ASME Code Case N-716 is founded, in large part, on the RI-ISI process as described in ElectricPower Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, "Revised Risk-Informed Decisionmaking Inservice Inspection of Piping," December 1999, which waspreviously reviewed and approved by the NRC (ADAMS Accession No. ML013470102). As arisk-informed application, this submittal meets the intent and principles of Regulatory Guide1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions onPlant-Specific Changes to the Licensing Basis," and Regulatory Guide 1.178, "An Approach forPlant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping."

et EL

NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

U.S. Nuclear Regulatory ConmnissionSBK-L- 11197

Attachment 2 to this letter contains a commitment regarding fire protection piping segmentsrelated to risk-informed inservice inspection requirements.

If you have any questions regarding this submittal, please contact Mr. Michael O'Keefe,Licensing Manager, at (603) 773-7745.

Sincerely,

NextEra Energy Seabrook, LLC

Paul 0. FreemanSite Vice President

cc: NRC Region I AdministratorG. E. Miller, NRC Project ManagerW. J. Raymond, NRC Resident Inspector

Attachments

Attachment 1

NextEra Energy Seabrook, LLCRequest for Approval of Risk-Informed/Safety Based

Inservice Inspection Alternative for Class 1 and 2 Piping3AR-1, Rev. 0

NEXTERA ENERGY SEABROOK, LLC

PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

3AR-1, REV. 0

--Alternative Provides Acceptable Level of Quality or Safety--

Request for Alternative to Use ASME Code Case N-716 to Implement Risk-Informed InserviceInspection Program

1. ASME Code Components Affected

Code Class: 1 and 2Examination Categories: B-F, B-J, C-F-1 and C-F-2

2. Applicable Code Edition and Addenda

The applicable Code edition and addenda is ASME Section XI, Rules for InserviceInspection of Nuclear Power Plant Components, 2004 Edition. In addition, as required by 10CFR 50.55a, piping ultrasonic examinations are performed per ASME Section XI, 2001Edition, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems.

3. Applicable Code Requirement

For the current inservice inspection (ISI) program at Seabrook, IWB-2200 IWB-2420, IWB-2430, and IWB-2500 provide the examination requirements for Category B-F and CategoryB-J welds. Similarly, IWC-2200, IWC-2420, IWC-2430, and IWC-2500 provide theexamination requirements for Category C-F-I and C-F-2 welds.

4. Reason for Request

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

5. Proposed Alternative And Basis for Use

In lieu of the existing Code requirements, Seabrook proposes to use a RISB process as analternate to the current ISI program for Class 1 and 2 piping. The RIS_B process used in thissubmittal is based upon ASME Code Case N-716, Alternative Piping Classification andExamination Requirements, Section XI, Division 1.

Code Case N-716 is founded, in large part, on the RI-ISI process described in Electric PowerResearch Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk-InformedInservice Inspection Evaluation Procedure, December 1999 (ADAMS Accession No.MLO 13470102) which was previously reviewed and approved by the U.S. NuclearRegulatory Commission (NRC).



3AR-1, REV. 0

In general, a risk-informed program replaces the number and locations of nondestructiveexamination (NDE) inspections based on ASME Code, Section XI requirements with thenumber and locations of these inspections based on the risk-informed guidelines. Theseprocesses result in a program consistent with the concept that, by focusing inspections on themost safety-significant welds, the number of inspections can be reduced while at the sametime 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 UsingProbabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to theLicensing Basis and RG 1.178, An Approach For Plant-Specific Risk-Informed DecisionMaking for Inservice Inspection of Piping. These steps are:

Scope definitionConsequence evaluationDegradation mechanism evaluationPiping segment definitionRisk categorizationInspection/NDE selectionRisk impact assessmentImplementation monitoring and feedback

These same steps were also applied to this RISB process and it is concluded that this RIS_Bprocess alternative also meets the intent and principles of Regulatory Guides 1.174 and1.178.

In general, the methodology in Code Case N-716 replaces a detailed evaluation of the safetysignificance of each pipe segment required by EPRI TR 112657, Rev. B-A with a genericpopulation of high safety-significant segments, supplemented with a rigorous floodinganalysis 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 1.200and ASME RA-Sb-2009, Standard for Probabilistic Risk Assessment for Nuclear PlantApplications.

By using risk-insights to focus examinations on more important locations, while meeting theintent and principles of Regulatory Guides 1.174 and 1.178, this proposed RIS B programwill continue to maintain an acceptable level of quality and safety. Additionally, all pipingcomponents, regardless of risk classification, will continue to receive ASME Code-requiredpressure testing, as part of the current ASME Code, Section XI program. Therefore, approvalfor 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 CFR50.55a(a)(3)(i). A Seabrook specific template for the application of ASME Code Case N-716 attached.



3AR-1, REV. 0

6. Duration of Proposed Alternative

Through the 3rd 10-Year Interval ending August 18, 2020.

7. Precedents

Similar alternatives have been approved for Vogtle Electric Generating Plant, Donald C.Cook 1 and 2, Grand Gulf Nuclear Station, Waterford-3 and North Anna 1 & 2.

8. References

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

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

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

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



3AR-1, REV. 0

TEMPLATE SUBMITTAL

APPLICATION OF ASME CODE CASE N-716

RISK-INFORMED/SAFETY-BASED (RISB)INSER VICE INSPECTION PROGRAM PLAN



3AR-1, REV. 0

Technical Acronyms/Definitions Used in the Template

ACAFWASASEPASMEATWTBERBL-PRACAFTACCCCCCDPCCFCCWCDFCIVClass 2 LSSCLERPCSCVCSDADCDME-CECSCCEOOSFACF&OFLBFTFWHELBHEPHFEHRHRAHSSIEIFIFIVIGSSCILOCA

Alternating CurrentAuxiliary FeedwaterAccident Sequence AnalysisAccident Sequence Evaluation ProgramAmerican Society of Mechanical EngineersAnticipated Transient Without TripBreak Exclusion RegionBase Line PRAComputer-Aided Fault Tree AnalysisPRA abbreviation for Capacity CategoryCrevice CorrosionConditional Core Damage ProbabilityCommon Cause FailureComponent Cooling WaterCore Damage FrequencyContainment Isolation ValveClass 2 Pipe Break in LSS PipingConditional Large Early Release ProbabilityContainment SprayChemical Volume and Control SystemData analysisDirect CurrentDegradation MechanismErosion-CorrosionExternal Chloride Stress Corrosion CrackingEquipment Out of ServiceFlow-Accelerated CorrosionFacts and ObservationsFeedwater Line BreakFault treeFeedwaterHigh Energy Line Break (synonymous with BER)Human Error ProbabilityHuman Failure EventHuman ReliabilityHuman Reliability AnalysisHigh Safety-SignificantInitiating Events AnalysisInternal FloodingInside First Isolation ValveIntergranular Stress Corrosion CrackingIsolable Loss of Coolant Accident



3AR-1, REV. 0

Technical Acronyms/Definitions Used in the Template (cont'd.)IPE Individual Plant EvaluationISEAL Isolable RCP seal injection line LOCALE LERF AnalysisLERF Large Early Release FrequencyLOCA Loss of Coolant AccidentLOSP Loss of Off-Site PowerLSS Low Safety-SignificantMAAP Modular Accident Analysis ProgramMIC Microbiologically-Influenced CorrosionMOV Motor Operated ValveMS Main SteamMU Model UpdateNDE Nondestructive ExaminationNNS Non-Nuclear SafetyNPS Nominal Pipe SizePBF Pressure Boundary FailurePIT PittingPLOCA Potential Loss of Coolant AccidentPOD Probability of DetectionPRA Probabilistic Risk AssessmentPSA Probabilistic Safety AssessmentPWSCC Primary Water SCCQU QuantificationRC Reactor CoolantRCP Reactor Coolant PumpRCPB Reactor Coolant Pressure BoundaryRG Regulatory GuideRHR Residual Heat RemovalRI-BER Risk-Informed Break Exclusion RegionRI-ISI Risk-Informed Inservice InspectionRISB Risk-Informed/Safety Based Inservice InspectionRM Risk ManagementRPV Reactor Pressure VesselSBO Station BlackoutSC Success CriteriaSDC Shutdown CoolingSEAL RCP Seal Injection Line LOCASLB Steam Line BreakSGTR Steam Generator Tube RuptureSSBI Main Steam or Feedwater Break inside the Outer CIVSSBO Main Steam or Feedwater Break Beyond the Outer CIVSSC Systems, Structures, and ComponentsSR Supporting Requirements



3AR-1, REV. 0

Technical Acronyms/Definitions Used in the Template (cont'd.)SW Service WaterSXI Section XISY Systems AnalysisTASCS Thermal Stratification, Cycling, and StripingTGSCC Transgranular Stress Corrosion CrackingTR Technical ReportTT Thermal TransientsVol Volumetric



3AR-1, REV. 0

Table of Contents

1. Introduction

1.1 Relation to NRC Regulatory Guides 1.174 and 1.178

1.2 PSA Quality

2. Proposed Alternative to Current Inservice Inspection Programs

2.1 ASME Section XI

2.2 Augmented Programs

3. Risk-Informed/Safety-Based ISI 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 ISI Plan Change

5. References/Documentation

Attachment A - Seabrook PRA Quality Review



3AR-1, REV. 0

1. INTRODUCTION

NextEra Energy Seabrook, LLC (NextEra) is currently in the third inservice inspection (ISI)interval as defined by the American Society of Mechanical Engineers (ASME) Boiler andPressure Vessel Section XI Code for Inspection Program B. NextEra plans to implement arisk-informed/safety-based inservice inspection (RISB) program in this third ISI interval.The third ISI interval commenced on August 19, 2010.

The ASME Section XI Code of record for the third ISI interval is the 2004 Edition forExamination Category B-F, B-J, C-F-i, and C-F-2 Class I and 2 piping components.

The RISB 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 RI-ISI process as described in Electric Power ResearchInstitute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk-Informed InserviceInspection Evaluation Procedure.

1.1 Relation to NRC Regulatory Guides 1.174 and 1.178

As a risk-informed application, this submittal meets the intent and principles ofRegulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis, and RegulatoryGuide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking InserviceInspection of Piping. Additional information is provided in Section 3.4.2 relative todefense-in-depth.

1.2 Probabilistic Safety Assessment (PSA) Quality

The methodology in Code Case N-716 provides for examination of a genericpopulation of high safety significant (HSS) segments, supplemented with a rigorousinternal flood events risk analysis to identify if any plant-specific HSS segments needto be added. Satisfying the requirement for the plant-specific analysis requiresconfidence that the internal flood events PRA is capable of successfully identifying anysignificant flooding contributors that are not identified in the generic population.

The Seabrook PRA used to support the risk aspects of this ASME Section XI CodeCase N-716 evaluation is a full scope Level I and Level 2 integrated analysis for atpower conditions. The fidelity and technical adequacy of the Seabrook PRA andsupporting documentation are maintained through a formalized process of maintainingand periodically updating the PRA and by performing self-assessments andindependent peer reviews. The latest capability assessment of the Seabrook PRA wasmeasured against the current ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009),as endorsed by NRC Regulatory Guide 1.200 Rev 2.

The PRA model used to support the ASME Code Case N-716 application is SSPSS-2011. The SSPSS-2011 model has been recently updated and contains the latest



3AR-1, REV. 0

upgraded internal flood events risk assessment. The upgraded internal flood eventsmodel has undergone Peer Review.

EPRI Report 1021467, "Nondestructive Evaluation: Probabilistic Risks AssessmentTechnical Adequacy Guidance for Risk-Informed Ifi-Service Inspection Programs" isused to demonstrate the level of the Seabrook PRA technical adequacy needed fordevelopment of the Seabrook risk-informed inservice inspection program.

Attachment A provides a description of the Seabrook PRA Capability status. Based onthe maintenance, update and technical capability evaluations in Appendix A, it isconcluded that the Seabrook PRA has sufficient technical capability and completenessto adequately support application of ASME Code Case N-716.

The following subsections address (a) the internal flood events risk upgrade and (b) theuse of specific screening criteria used in the internal flood events model.

Internal Flood Risk Assessment

Internal flood assessment aspects that are particularly noteworthy to the Code Case N-716 application include the following:

Scope of HSS Pipe Segments/Welds

The internal flood risk assessment is used to determine the scope of high safetysignificant welds.

ASME Code Case N-716, Section 2(a)(5) requires HSS classification for any pipingsegment whose contributions to core damage frequency is greater than I E-06 basedupon a plant-specific probabilistic risk assessment (PRA) of pressure boundary failures(e.g., pipe whip, jet impingement, spray, and inventory losses). All piping classes,including Class 3 or Non-Class piping, are required to be included in this assessment.The PRA quality basis is required to be reviewed to confirm that any such pipingsegments are applicable to the high safety significant categorization of this Case.

Seabrook internal flood events risk analysis identified that 4" and 6" diameter FireProtection piping segments located in the Control Building stairwell contributed greaterthan 1 E-06/yr to the core damage frequency. This piping supplies fire water hosestations located within the Control Building and Diesel Generator Building stairwells.A postulated pipe break in the Control Building stairwell greater than the design basissize has the potential to propagate to the essential switchgear rooms, impactingessential electrical power.

A prudent risk management measure is being taken to essentially eliminate the FireProtection flooding risk from large pipe breaks in the Control Building. The riskreduction measure is installation of a flow orifice in Fire Protection piping located in



3AR-1, REV. 0

the RCA Access Walkway, upstream of the Control Building stairwell. Installing anorifice at this location will limit the postulated maximum Fire Protection break flowrate in the Control Building, and thus reduce the flood risk from the Fire Protectionpipe segments to less than 1 E-06/yr. This would eliminate the Fire Protection pipingfrom the RIS_B scope.

Following installation of the Fire Protection orifice, there will be no piping segmentsthat contribute greater than 1 E-06 to the core damage frequency or greater than I E-07to large early release frequency. No additional high safety significant welds wereidentified via the Seabrook internal flood events risk assessment.

Highly Reliable Actions Used for Screening

The Seabrook internal flood capability category for Supporting Requirements (SRs)IFSN-A14 (IF-C6) and IFSN-A16 (IF-C8) is CC-II. These SRs allow screening offlood areas and/or flood sources if the action can be performed with "high reliability"for the worst flooding initiator. These SRs describe what attributes are needed toconsider the action highly reliable. These attributes are formulated into conservativescreening criteria, which are applied in the Seabrook internal flood risk assessment.

Highly reliable actions apply to most small floods that have no spray impact (sprayimpact would be expected within the first few minutes). For large floods, highlyreliable actions are dependent on the assessment of time available vs time required.The "highly reliable action" screen is only used for flood scenarios with at least 60 minavailable before plant damage. These scenarios must have a control room cue of theflood and must have procedures that provide some direction in flood mitigation. If themitigation requires local action, the area and the path to the area must be accessible.Based on the assessment performed, the total HEP for a highly reliable action is judgedto be less than 2E-4 and the highly reliable action screening criteria applied in theSeabrook flood risk assessment include the following:

a. For small floods (where the cues may be more subtle), highly reliable actionsmust have at least 3 hours for mitigation action.

b. For large floods with generic cues, procedures, and training, highly reliableactions must have at least 2 hours for mitigation action.

c. For large floods with specific cues, procedures, and training, highly reliableactions must have at least 1 hr for mitigation action. The cue, procedure, andtraining are specific for flood scenarios where clear cues point to specific AOPsthat provide direct mitigation actions for the flood.

Based on application of this conservative criteria, no flood scenarios (or pipe segments)have been screened that might otherwise meet or exceed the quantitative CDF andLERF scoping guideline in Code Case N-716.



3AR-1, REV. 0

2. PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS

2.1 ASME Section XI

ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 currentlycontain requirements for the nondestructive examination (NDE) of Class 1 and 2 pipingcomponents.

The alternative RISB Program for piping is described in Code Case N-716. TheRIS_B Program will be substituted for the current program for Class 1 and 2 piping(Examination Categories B-F, B-J, C-F-I and C-F-2) in accordance with 10 CFR50.55a(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.

2.2 Augmented Programs

The impact of the RISB application on the various plant augmented inspectionprograms listed below were considered. This section documents only those plantaugmented inspection programs that address common piping with the RIS_Bapplication scope (e.g., Class 1 and 2 piping).

" The plant augmented inspection program for high-energy line breaks outsidecontainment has not been revised by this application. A separate evaluation andprogram is maintained in accordance with the risk-informed break exclusion regionmethodology (RI-BER) described in EPRI Report 1006937, Extension of EPRI RiskInformed ISI Methodology to Break Exclusion Region Programs.

* A plant augmented inspection program has been implemented in response to NRCBulletin 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 ReactorCoolant System Branch Lines. The thermal fatigue concern addressed was explicitlyconsidered in the application of the RISB process and is subsumed by the RMS_BProgram.

" The plant augmented inspection program for flow accelerated corrosion (FAC) perGL 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, is relied upon tomanage this damage mechanism but is not otherwise affected or changed by theRIS_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) ofunmitigated Alloy 82/182 welds has occurred at pressurized water reactors. ForSeabrook, 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-



3AR-1, REV. 0

end welds and four cold leg nozzle to safe-end welds. The steam generatordissimilar metal welds are not subject to PWSCC because the welds are Alloy52/152, and all of the pressurizer dissimilar metal welds (and the adjacentstainless steel welds) greater than 1" Nominal Pipe Size (NPS) have been overlaidwith Full Structural Weld Overlays (FSWOL). All of the overlaid welds havebeen removed from the risk-informed program and will be examined inaccordance with the requirements set forth in the NRC safety evaluation for theweld overlays.

Seabrook has selected the four RPV hot leg nozzle Alloy 82/182 welds forultrasonic examination for PWSCC within the scope of Code Case N-716. CodeCase N-716 requires examination of these welds every ten years. However, theexamination frequency of butt welds is currently established by NRC rulemaking.The RIS_B Program will not be used to eliminate any welds with overridingregulatory requirements.

Per Code Case N-716 (Table 1, Item No. 1.15, Elements Subject to PrimaryWater Stress Corrosion Cracking (PWSCC), selected butt welds are subject tovolumetric examination. Per Note 3 of Table 1, the examination includesessentially 100% of the examination location. When the required examinationvolume or area cannot be examined due to interference by another component orpart geometry, limited examinations shall be evaluated for acceptability. Areaswith acceptable limited examinations (coverage less or equal to 90%), and theirbases, shall be documented and submitted for relief per the requirements of 10CFR 50.55a(g)(5)(iv).

3. RISK-INFORMED/SAFETY-BASED ISI PROCESS

The process used to develop the RIS_B Program conformed to the methodology described inCode 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:



3AR-1, REV. 0

3.1 Safety Significance Determination

The systems assessed in the RIS B Program are provided in Table 3.1. The piping andinstrumentation diagrams and additional plant information, including the existing plantISI Program were used to define the piping system boundaries. Per Code Case N-716requirements, piping welds are assigned safety-significance categories, which are thenused 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 asprovided in 10 CFR 50.55a(c)(2)(i) and (c)(2)(ii);

(2) Applicable portions of the shutdown cooling pressure boundary function. That is,Class I and 2 welds of systems or portions of systems needed to utilize the normalshutdown cooling flow path either:

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

(b) Other systems or portions of systems from the RPV to the second isolationvalve (i.e., farthest from the RPV) capable of remote closure or to thecontainment 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 outercontainment isolation valve;

(4) Piping within the break exclusion region (BER) greater than 4" NPS for high-energy piping systems as defined by the Owner. Per Code Case N-716, this mayinclude Class 3 or Non-Class piping, but all BER piping at Seabrook isClass 2.

(5) Any piping segment whose contribution to Core Damage Frequency (CDF) isgreater than I E-06 [and per NRC feedback on the Grand Gulf and D. C. CookRISB applications 1 E-07 for Large Early Release Frequency (LERF)] basedupon a plant-specific PSA of pressure boundary failures (e.g., pipe whip, jetimpingement, spray, inventory losses). This may include Class 3 or Non-Classpiping. No piping segments with a contribution to CDF greater than 1 E-06(ME-07 for LERF) were identified.



3AR-1, REV. 0

3.2 Failure Potential Assessment

Failure potential estimates were generated utilizing industry failure history, plant-specificfailure history, and other relevant information. These failure estimates were determinedusing the guidance provided in NRC approved EPRI TR-1 12657 (i.e., the EPRI RI-ISImethodology), with the exception of the deviation discussed below.

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

As previously approved for Seabrook during last interval, a deviation to the EPRIRISB methodology has been implemented in the failure potential assessment. Table3-16 of EPRI TR-l 12657 contains the following criteria for assessing the potential forThermal Stratification, Cycling, and Striping (TASCS). Key attributes for horizontal orslightly sloped piping greater than NPS 1 include:

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

2. The potential exists for leakage flow past a valve, including in-leakage, out-leakageand 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 pipeconnected to header piping containing hot fluid with turbulent flow;

AND

AT > 500F,

AND

Richardson Number > 4 (this value predicts the potential buoyancy of astratified flow)

These criteria, based on meeting a high cycle fatigue endurance limit with the actual ATassumed equal to the greatest potential AT for the transient, will identify locationswhere stratification is likely to occur, but allows for no assessment of severity. Assuch, many locations will be identified as subject to TASCS, where no significantpotential for thermal fatigue exists. The critical attribute missing from the existingmethodology, that would allow consideration of fatigue severity, is a criterion that



3AR-1, REV. 0

addresses the potential for fluid cycling. The impact of this additional consideration onthe existing TASCS susceptibility criteria is presented below.

Turbulent Penetration TASCS

Turbulent penetration is a swirling vertical flow structure in a branch line inducedby high velocity flow in the connected piping. It typically occurs in linesconnected to piping containing hot flowing fluid. In the case of downwardsloping lines that then turn horizontal, significant top-to-bottom cyclic ATs candevelop in the horizontal sections if the horizontal section is less than about 25pipe diameters from the reactor coolant piping. Therefore, TASCS is consideredfor this configuration.

For upward sloping branch lines connected to the hot fluid source that turnhorizontal or in horizontal branch lines, natural convective effects combined witheffects of turbulence penetration will tend to keep the line filled with hot water. Ifthere is in-leakage of cold water, a cold stratified layer of water may be formedand significant top-to-bottom ATs may occur in the horizontal portion of thebranch line. Interaction with the swirling motion from turbulent penetration maycause a periodic axial motion of the cold layer. Therefore, TASCS is consideredfor 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-bottomATs will not occur. Therefore, TASCS is not considered for these no in-leakageconfigurations. Even in fairly long lines, where some heat loss from the outsideof the piping will tend to occur and some fluid stratification may be present, thereis no significant potential for cycling as has been observed for the in-leakage case.The effect of TASCS will not be significant under these conditions and can beneglected.

Low flow TASCS

In some situations, the transient startup of a system (e.g., shutdown coolingsuction 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 rapidlydisplace the cold fluid in stagnant lines, while fluid mixing will occur in thepiping further removed from the hot source and stratified conditions will existonly briefly as the line fills with hot fluid. As such, since the situation is transientin nature, it can be assumed that the criteria for thermal transients (TT) willgovern.

Valve leakage TASCS

Sometimes a very small leakage flow of hot water can occur outward past a valveinto a line that is relatively colder, creating a significant temperature difference.



3AR-1, REV. 0

However, since this is generally a "steady-state" phenomenon with no potentialfor cyclic temperature changes, the effect of TASCS is not significant and can beneglected.

SConvection Heating TASCS

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

In summary, these additional considerations for determining the potential for thermalfatigue as a result of the effects of TASCS provide an allowance for considering cycleseverity. Consideration of cycle severity was used in previous NRC approved RIS_Bprogram submittals for D. C. Cook, Grand Gulf Nuclear Station, Waterford-3, and theVogtle Electric Generating Plant as well as Seabrook during the past interval. Themethodology used in the Seabrook RISB application for assessing TASCS potentialconforms to these updated criteria. Additionally, materials reliability program (MRP)MRP-146 guidance on the subject of TASCS was also incorporated into the SeabrookRIS_B application.

3.3 Element and NDE Selection

Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RISBapplications provided criteria for identifying the number and location of requiredexaminations. Ten percent of the HSS welds shall be selected for examination asfollows:

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

(a) A minimum of 25% of the population identified as susceptible to eachdegradation mechanism and degradation mechanism combination shall beselected.

(b) If the examinations selected above exceed 10% of the total number of HSSwelds, the examinations may be reduced by prorating among eachdegradation mechanism and degradation mechanism combination, to theextent 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 weldpopulation, additional welds shall be selected so that the total numberselected for examination is at least 10%.

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



3AR-1, REV. 0

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

(4) A minimum of 10% of the welds in that portion of the RCPB that lies outsidecontainment (not applicable for Seabrook) shall be selected.

(5) A minimum of 10% of the welds within the break exclusion region (BER) shall beselected. Currently, there are 196 welds at Seabrook in the BER program. TheseBER welds consist of Class 2 welds in the main steam and feedwater systems. ARI-BER program has been implemented for these welds, which also requiredmore than 10% of the population to be examined.

In contrast to a number of traditional RI-ISI program applications, where the percentageof 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 ofthe number of welds and the number selected is provided below, and the results of theselections are presented in Table 3.3. Section 4 of EPRI TR-l 12657 was used asguidance in determining the examination requirements for these locations. Only thoseRIS_B inspection locations that receive a volumetric examination are included.

Class 1 Welds~ll Class 2 Welds•2 • All PipingWelds• 3 )

Total Selected Total Selected Total Selected

750 78 2317 34 3067 112Notes:(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-2 locations. Of the Class 2 piping weldlocations,345 are HSS; the remaining are LSS.

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

3.3.1 Current Examinations

Seabrook is currently using the traditional ASME Section XI inspectionmethodology for ISI examination of Class 2 piping welds per the 2004 Editionof ASME Section XI and for Class I piping welds, the NRC has previouslyapproved an application using EPRI-TR 112657B-A.



3AR-1, REV. 0

3.3.2 Successive Examinations

If indications are detected during RISB ultrasonic examinations, they will beevaluated per IWB-3514 (Class 1) or IWC-3514 (Class 2) to determine theiracceptability. Any unacceptable flaw will be evaluated per the requirements ofASME Code Section XI, IWB-3600 or IWC-3600, as appropriate. As part ofthis evaluation, the degradation mechanism that is responsible for the flaw willbe determined and accounted for in the evaluation. If the flaw is acceptable forcontinued service, successive examinations will be scheduled per Section 6 ofCode Case N-716. If the flaw is found unacceptable for continued operation, itwill be repaired in accordance with IWA-4000, applicable ASME Section XICode Cases, or NRC approved alternatives. The IWB-3600 analyticalevaluation will be submitted to the NRC. Finally, the evaluation will bedocumented in the corrective action program and the Owner submittals requiredby Section XI. Evaluation of indications attributed to PWSCC and successiveexaminations of PWSCC indications will be performed in accordance with NRCrule making.

3.3.3 Scope Expansion

If the nature and type of the flaw is service-induced, then welds subject to thesame type of postulated degradation mechanism will be selected and examinedper Section 6 of Code Case N-716. The evaluation will include whether otherelements in the segment or additional segments are subject to the same rootcause conditions. Additional examinations will be performed on those elementswith the same root cause conditions or degradation mechanisms. The additionalexaminations will include HSS elements up to a number equivalent to thenumber of elements required to be inspected during the current outage. Ifunacceptable flaws or relevant conditions are again found similar to the initialproblem, the remaining elements identified as susceptible will be examinedduring the current outage. No additional examinations need be performed ifthere are no additional elements identified as being susceptible to the same rootcause conditions. The need for extensive root cause analysis beyond thatrequired for the IWB-3600 analytical evaluation will be dependent on practicalconsiderations (i.e., the practicality of performing additional NDE or removingthe flaw for further evaluation during the outage).Scope expansion for flaws characterized as PWSCC will be conducted in

accordance with NRC rule making.

3.3.4 Program Relief Requests

Consistent with previously approved RISB submittals, Seabrook will calculatecoverage and use additional examinations or techniques in the same manner ithas for traditional Section XI examinations. Experience has shown this process



3AR-1, REV. 0

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 forthose cases where greater than 90% coverage is not obtained will be submittedper the requirements of 10 CFR 50.5 5a(g)(5)(iv).

No Seabrook relief requests are being withdrawn due to the RISB application.

3.4 Risk Impact Assessment

The RISB Program development has been conducted in accordance with RegulatoryGuide 1.174 and the requirements of Code Case N-716, and the risk fromimplementation of this program is expected to remain neutral or decrease whencompared to that estimated from current requirements.

This evaluation categorized segments as high safety significant or low safety significantin accordance with Code Case N-716, and then determined what inspection changeswere proposed for each system. The changes included changing the number andlocation of inspections, and in many cases improving the effectiveness of the inspectionto account for the findings of the RISB degradation mechanism assessment. Forexample, examinations of locations subject to thermal fatigue will be conducted on anexpanded 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-1 12657 process forrisk impact analyses, whereby limits are imposed to ensure that the change-in-risk of implementing the RISB Program meets the requirements of RegulatoryGuides 1.174 and 1.178. Section 3.7.2 of EPRI TR- 112657 requires that thecumulative change in CDF and LERF be less than 1E-07 and I E-08 per year persystem, respectively.

For LSS welds, Conditional Core Damage Probability (CCDP)/ConditionalLarge Early Release Probability (CLERP) values of 1 E-4/1E-5 wereconservatively 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 ISI (RI-ISI) methodology. As such, the goal is to determineCCDPs/CLERPs threshold values. For example, the threshold values betweenHigh and Medium consequence categories is 1 E-4 (CCDP)/1E-5 (CLERP) andbetween Medium and Low consequence categories are 1E-6 (CCDP)/1E-7(CLERP) from the EPRI RI-ISI Risk Matrix. Using these threshold valuesstreamlines the change-in-risk evaluation as well as stabilizes the updateprocess. For example, if a CCDP changes from lE-5 to 3E-5 due to an update,it will remain below the I E-4 threshold value; the change-in-risk evaluationwould not require updating.



3AR-1, REV. 0

The updated internal flooding PRA was also reviewed to ensure that there is noLSS Class 2 piping with a CCDP/CLERP greater than 1E-4/1E-5. This reviewidentified some piping in the RHR, CBS, SI and CVCS systems located outsideof containment with a CCDP greater than 1 E-4. As a result, all LSS welds inthese systems are conservatively assigned CCDP/CLERP equal to 5E-4/5E-5.

With respect to assigning failure potentials for LSS piping, the criteria aredefined in Table 3 of Code Case N-716. That is, those locations identified assusceptible to FAC are assigned a high failure potential. Those locationssusceptible to thermal fatigue, erosion-cavitation, corrosion, or stress corrosioncracking are assigned a medium failure potential, unless they have an identifiedpotential for water hammer loads. In such cases, they will be assigned a highfailure potential. Finally, those locations that are identified as not susceptible todegradation are assigned a low failure potential.

In order to streamline the risk impact assessment, a review was conducted thatverified that the LSS piping was not susceptible to water hammer. LSS pipingmay be susceptible to FAC; however, the examination for FAC is performed perthe FAC program. This review was conducted similar to that done for atraditional RI-ISI application. Thus, the high failure potential category is notapplicable to LSS piping. In lieu of conducting a formal degradationmechanism evaluation for all LSS piping (e.g. to determine if thermal fatigue isapplicable), these locations were conservatively assigned to the Medium failurepotential ("Assume Medium" in Table 3.4) for use in the change-in-riskassessment. Experience with previous industry RIS_B applications shows thisto be conservative.

Seabrook has conducted a risk impact analysis per the requirements of Section 5of Code Case N-716 that is consistent with the "Simplified Risk QuantificationMethod" described in Section 3.7 of EPRI TR-1 12657. The analysis estimatesthe net change-in-risk due to the positive and negative influences of adding andremoving locations from the inspection program.

The CCDP and CLERP values used to assess risk impact were estimated basedon pipe break location. Based on these estimated values, a correspondingconsequence rank was assigned per the requirements of EPRI TR- 112657 andupper bound threshold values were used as provided in the table below.Consistent with the EPRI methodology, the upper bound for all break locationsthat fall within the high consequence rank range was based on the highestCCDP value obtained (e.g., Large LOCA CCDP bounds the medium and smallLOCA CCDPs).



3AR-1, REV. 0Estimated Consequence Upper / Lower Bound

Break Location Description of Affected PipingCCDP CLERP Rank CCDP CLERP

LOCA 4E-02 4E-03(U4E0 ()4-3HIGH (U) 4E-02 (U) 4E-03 Unisolable RCPB piping of all sizesThe highest CCDP is Large LOCA and (0.1 margin for CLERP) (L) IE-04 (L) IE-05

ILOCA IE-04 IE-05 (U) IE-04 (U) IE-05 Piping between 1st and 2nd normally open isolation valve

Calculated based on Large LOCA CCDP of 4E-2 and valve fail to MEDIUM inside containment (CS Seal Injection, charging and RCclose probability of <3E-3 (0.1 margin for CLERP) (L) IE-06 (L) IE-07 letdown)

PLOCA I E-04 I E-05

Calculated based on Large LOCA CCDP of 4E-2 and valve rupture (U) I E-04 (U) IE-05 Piping beyond the I st normally closed isolation valve insideprobability of<l E-3 (0.1 margin for CLERP). RHR shutdown MEDIUM containment (CS excess letdown, auxiliary pressurizer spray,cooling suction and return paths are included. The failure of this (L) IE-06 (L) IE-07 RC drains, RHR suction and return, SI injection paths)piping during shutdown (ILOCA) is also bounded by this CCDP.

SLB I E-05 I E-06

Several feedwater and steam line breaks inside containment and (U) I E-04 (U) IE-05outside containment have CCPDs in the mid to high I E-6 range. To MEDIUM (L) I E-06 (L) I E-07 Secondary breaks in the FW and MS systemssimplify the analysis a bounding CCDP for secondary line breaks(SLB) is used (0.1 margin for CLERP)

LSS I E-04 IE-05 MEDIUM (U) IE-04 (U) IE-05 All other Class 2 system piping designated as low safetyEstimated based on upper bound for Medium Consequence (L) I E-06 (L) I E-07 significant in the RCS, FW and MS systems

LSS FD 5E-04 I 5E-05

Based on internal flooding CCDP for class 2 piping in portions of HIGH (U) 5E-04 (U) 5E-05 All other Class 2 system piping designated as low safetyCBS, CS, R-H and SI systems (applies to all welds in these systems (L) IE-04 (L) IE-05 significant in the CBS, CS, RH and SI systemsand 0.1 margin for CLERP applies)

Note: The PRA does not explicitly model potential (PLOCA) and isolable (ILOCA) LOCA events, because such eventsare subsumed by the LOCA initiators in the PRA. That is, the frequency of a LOCA in this limited piping downstream ofthe 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. The upper bound ILOCACCDP of 1E-04 is a bounding value based on MOV demand failure of 1.07E-03 (NUREG/CR-6928). Thus, 4E-02 * 1.1E-03 = 4.4E-05 < 1E-04 CCDP.



3AR-1, REV. 0

The likelihood of pressure boundary failure (PBF) is determined by thepresence of different degradation mechanisms and the rank is based onthe relative failure probability. The basic likelihood of PBF for a pipinglocation with no degradation mechanism present is given as x0 and isexpected to have a value less than I E-08. Piping locations identified asmedium failure potential have a likelihood of 20xo. These PBFlikelihoods are consistent with References 9 and 14 of EPRI TR-l 12657.In addition, the analysis was performed both with and without takingcredit for enhanced inspection effectiveness due to an increased PODfrom application of the RISB approach.

Table 3.4 presents a summary of the RISB Program versus the 1995 Editionthrough 1996 Addenda of ASME Section XI program requirements on a "persystem" basis for the second interval. The presence of FAC was adjusted for inthe quantitative analysis by excluding its impact on the failure potential rank.The exclusion of the impact of FAC on the failure potential rank and thereforein the determination of the change-in-risk, was performed because FAC is adamage mechanism managed by a separate, independent plant augmentedinspection program. The RISB Program credits and relies upon this plantaugmented inspection program to manage this damage mechanism. The plantFAC program will continue to determine where and when examinations shall beperformed. Hence, since the number of FAC examination locations remains thesame "before" and "after" (the implementation of the RISB program) and nodelta exists, there is no need to include the impact of FAC in the performance ofthe risk impact analysis.

As indicated in the following table, this evaluation has demonstrated thatunacceptable risk impacts will not occur from implementation of the RISBProgram, and that the acceptance criteria of Regulatory Guide 1.174 and CodeCase N-716 are satisfied.

With POD Credit Without POD CreditSystem Delta Delta Delta Delta

CDF LERF CDF LERFContainment Spray(CBS) 1.35E-09 1.35E-10 1.35E-09 1.35E-10CVCS (CS) -4.13E-08 -4.13E-09 -2.21E-08 -2.21E-09Feedwater (FW) -4.OOE- 11 -4.OOE-12 -1.60E- 1I -1.60E-12Main Steam (MS) 6.OOE- 1 6.OOE-12 6.OOE-1 1 6.OOE-12Reactor Coolant (RC) -1.03E-07 -1.03E-08 -2.34E-09 -2.34E-10RHR (RH) -2.63E-08 -2.63E-09 -3.89E-09 -3.89E-10Safety Injection (SI) -5.98E-08 -5.98E-09 -1.98E-08 -1.98E-09

Total -2.29E-07 -2.29E-08 -4.68E-08 -4.68E-09



3AR-1, REV. 0

As shown in Table 3.4, new RIS B locations were selected such that the RIS Bselections exceed the Section XI selections for certain categories (Delta columnhas a positive number). To show that the use of a conservative upper boundCCDP/CLERP does not result in an optimistic calculation with regard to meetingthe acceptance criteria, a conservative sensitivity was conducted where the RIS_Bselections were set equal to the Section XI selections (Delta changed frompositive number to zero). The acceptance criteria are met when the number ofRISB selections is not allowed to exceed Section XI.

3.4.2 Defense-in-Depth

The intent of the inspections mandated by 10 CFR 50.55a for piping welds is toidentify conditions such as flaws or indications that may be precursors to leaksor ruptures in a system's pressure boundary. Currently, the process for selectinginspection locations is based upon terminal end locations, structuraldiscontinuities, and stress analysis results. As depicted in ASME White Paper92-01-01 Rev. 1, Evaluation of Inservice Inspection Requirements for Class 1,Category B-J Pressure Retaining Welds, this methodology has been ineffectivein identifying leaks or failures. EPRI TR-1 12657 and Code Case N-716 providea more robust selection process founded on actual service experience withnuclear plant piping failure data.

This process has two key independent ingredients; that is, a determination ofeach location's susceptibility to degradation and secondly, an independentassessment of the consequence of the piping failure. These two ingredientsassure defense-in-depth is maintained. First, by evaluating a location'ssusceptibility to degradation, the likelihood of finding flaws or indications thatmay be precursors to leak or ruptures is increased. Secondly, a genericassessment of high-consequence sites has been determined by Code CaseN-716, supplemented by plant-specific evaluations, thereby requiring aminimum threshold of inspection for important piping whose failure wouldresult in a LOCA or BER break. Finally, Code Case N-716 requires that anypiping on a plant-specific basis that has a contribution to CDF of greater thanI E-06 (or 1 E-07 for LERF) be included in the scope of the application.Seabrook did not identify any such piping.

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

3.5 Implementation

Upon approval of the RISB Program, procedures that comply with the guidelinesdescribed in Code Case N-716 will be prepared to implement and monitor the program.The new program will be implemented during the third ISI interval. No changes to the



3AR-1, REV. 0

Technical Specifications or Updated Final Safety Analysis Report are necessary forprogram implementation.

The applicable aspects of the ASME Code not affected by this change will be retained,such as inspection methods, acceptance guidelines, pressure testing, correctivemeasures, documentation requirements, and quality control requirements. ExistingASME Section XI program implementing procedures will be retained and modified toaddress the RISB process, as appropriate.

3.6 Feedback (Monitoring)

The RISB Program is a living program that is required to be monitored continuouslyfor changes that could impact the basis for which welds are selected for examination.Monitoring encompasses numerous facets, including the review of changes to the plantconfiguration, changes to operations that could affect the degradation assessment, areview of NDE results, a review of site failure information from the corrective actionprogram, and a review of industry failure information from industry operatingexperience (OE). Also included is a review of PRA changes for their impact on theRIS_B program. These reviews provide a feedback loop such that new relevantinformation is obtained that will ensure that the appropriate identification of HSSpiping locations selected for examination is maintained. As a minimum, this reviewwill be conducted on an ASME period basis. In addition, more frequent adjustmentmay be required as directed by NRC Bulletin or Generic Letter requirements, or byindustry 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 andprocedures. 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 developa corrective action plan or plans).

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

E. Implement (complete the work necessary to correct the problem and preventrecurrence).

F. Monitor (through the audit process ensure that the RISB program has beenupdated based on the completed corrective action).

G. Trend (Identify conditions that are significant based on accumulation of similarissues).



3AR-1, REV. 0

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

4. PROPOSED ISI PLAN CHANGE

Seabrook is currently in the first period of the third ISI interval and is using the traditionalASME Section XI inspection methodology for ISI examination of piping welds. At least16% of the ASME Section XI piping examinations will be performed by the end of the firstperiod of the third ISI interval to ensure compliance with the traditional ASME Section XIinspection methodology during the transition to N-716.

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

Alternatively, first period examinations will be completed using the traditional ASME XImethodology. Then, the second and third period examinations will utilize the RIS_Bmethodology. In this case, approximately 1/3 rd of the total number of RISB piping weldsselected for examination will be examined in each of the two remaining periods.

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

A comparison between the RISB Program and the 1995 Edition through 1996 Addenda ofSection XI program requirements for in-scope piping is provided in Table 4.

5. REFERENCES/DOCUMENTATION

EPRI Report 1006937, Extension of EPRI Risk Informed ISI Methodology to Break ExclusionRegion Programs.

EPRI TR-1 12657, Revised Risk-Informed Inservice 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.

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



3AR-1, REV. 0

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

USNRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for AlternativeGG-ISI-002-Implement Risk-Informed ISI based on ASME Code Case N-716, datedSeptember 21, 2007.

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

EPRI Report 1021467 Nondestructive Evaluation: Probabilistic Risk Assessment TechnicalAdequacy Guidance for Risk-Informed In-Service Inspection Programs.

Supporting Onsite Documentation

EPRI Report "ASME Code Case N-716 Evaluation Seabrook Station", July 2011



3AR-1, REV. 0

Table 3.1Code Case N-716 Safety Significance Determination

System Weld N-716 Safety Significance Determination Safety SignificanceCount RCPB SDC PWR: FW BER CDF> IE-6 (1) High Low

CBS 354 "76 V"CS 7

692 €

83 " V50 V" V" V

FW 50

20 V ,39 "

126 V "MS ____ ____

164 _/

285 V V

41 V V VRC 11 " V"

64 V33 V V88 V, V, V"

RH55 V V

290 V

161 / V

S1 66 " V V

369 V555 V V

195 " V "SUMMARY 66 V VRESULTS 83FOR ALLSYSTEMS 50 V v V

146 V V

1972 V

TOTALS 3067

CBS = Containment Building SprayCS = Chemical Volume and Control SystemFW = Main FeedwaterMS = Main SteamRC = Reactor CoolantRH = Residual Heat RemovalSI = Safety Injection

(1) A Fire Protection modification is being made to ensure that there is no piping that exceedsthis criterion



3AR-1, REV. 0

Table 3.2Failure Potential Assessment Summaty

Thermal Localized FlowFatigue Stress Corrosion Cracking Corrosion Sensitive

System(1" TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FACCBSCSFWMSRC "RH V

Notes:1. Systems are described in Table 3.12. A degradation mechanism assessment was not performed on low safety significant piping

segments. This includes the CBS in its entirety, as well as portions of the CS, MS, RC, RH andSI systems.



3AR-1, REV. 0

Table 3.3: Code Case N716 SelectionsWeld Count N716 Selection ConsiderationsSystem SelectionsHSS LSS DMs RCPB RCPB (IFIV) RCPB (OC) BER

CBS 354 None 0CS 6 TT / " 6CS 33 TT / 2CS 4 None V " 0CS 33 None V" 0CS 692 None 0FW 70 None / 13FW 12 TASCS 3FW 71 None 0

FW 39 None 0MS 126 None V 13MS 164 None 0

RC 35 IT T V 7RC 15 TT,TASCS I V 11RC 16 TASCS V V 4RC 209 None V V 12

RC 51 None V 0RC 11 None 0RC 64 None 0RH 6 TASCS V " 5RH 6 None V V 4RH 109 None V 4RH 55 None 5RH 290 None 0

S1 8 TT V V 7S1 8 TT,TASCS V V 3S1 12 TT, IGSCC V 2

S1 6 IGSCC " 3S 26 None V V 6SI 167 None V 2

S1 369 None 049 TT V V 2033 TT 2

23 TT, TASCS V " 14

22 TASCS V V 9Summary 12 TT, IGSCC V 2ResultsRsl 6 IGSCC V 3

All

Systems 245 None V V 22360 None V 6

196 None V 2612 TASCS 3

137 None 5

Totals 1095 1972 None 112

Note: Systems are described in Table 3.1



3AR-1, REV. 0

Table 3.4 Risk Impact Analysis ResultsSafety Break Location Failure Potential Inspections CDF Impact LERF Impact

System Significance DMs Rank SXI RIS B Delta w/POD w/o POD w/POD w/o POD

CBS Total Low Class 2 LSS FD Assume Medium 27 0 -27 1.35E-09 1.35E-09 1.35E-10 1.35E-10

CS High LOCA TT Medium 0 6 6 -4.32E-08 -2.40E-08 -4.32E-09 -2.40E-09

CS High PLOCA/ILOCA TT Medium 0 2 2 -3.60E- I I -2.OOE- I1 -3.60E-12 -2.OOE-12

CS High LOCA None Low 0 0 0 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00

CS High ILOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00

CS Low Class 2 LSS FD Assume Medium 38 0 -38 1.90E-09 1.90E-09 1.90E-10 1.90E-10

CS Total -4.13E-08 -2.21 E-08 -4.13E-09 -2.21 E-09FW High SLB TASCS Medium 0 3 3 -5.40E- I1 -3.OOE- I I -5.40E-12 -3.OOE-12

FW High SLB None Low 41 13 -28 1.40E- I I 1.40E- I I 1.40E-12 1.40E-12

FW Low Class 2 LSS Assume Medium 0 0 0 0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00FW Total -4.OOE-1 1 -1.60E-1 I -4.OOE-12 -1.60E-12

MS High SLB None Low 33 13 -20 1.00E-1 I 1.00E-I I 1,00E-12 1.00E-12MS Low Class 2 LSS Assume Medium 5 0 -5 5.00E-I 1 5.OOE-I I 5,00E-12 5.00E-12

MS Total 6.OOE-11 6.OOE-I1 6.OOE-12 6.OOE-12

RC High LOCA TT Medium 7 7 0 -3.36E-08 0.OOE+00 -3.36E-09 0.00E+00

RC High LOCA TT, TASCS Medium 6 11 5 -6.48E-08 -2.00E-08 -6.48E-09 -2.00E-09

RC High LOCA TASCS Medium 6 4 -2 -1.44E-08 8.00E-09 -1.44E-09 8.00E-10

RC High LOCA None Low 60 12 -48 9.60E-09 9.60E-09 9.60E-10 9.60E-10

RC High PLOCA/ILOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00

RC Low Class 2 LSS Assume Medium 6 0 -6 6.OOE-1 I 6.OOE-I I 6.OOE-12 6.OOE-12

RC Total -1.03E-07 -2.34E-09 -1.03E-08 -2.34E-10

RH High LOCA TASCS Medium 4 5 1 -2.64E-08 -4.OOE-09 -2.64E-09 -4.OOE-10

RH High LOCA None Low 0 4 4 -8.OOE-10 -8.OOE-10 -8.OOE-I 1 -8.OOE-I 1

RH High PLOCA None Low 34 9 -25 1.25E-1 I 1.25E-1 I 1.25E-12 1.25E-12

RH Low Class 2 LSS FD Assume Medium 18 0 -18 9.00E-10 9.OOE-10 9.OOE- 11 9.00E-I 1

RH Total -2.63E-08 -3.89E-09 -2.63E-09 -3.89E-10

SI High LOCA TT Medium 5 7 2 -3.84E-08 -8.OOE-09 -3.84E-09 -8.OOE-10

SI High LOCA TT, TASCS Medium 0 3 3 -2.16E-08 -1.20E-08 -2.16E-09 -1.20E-09

SI High PLOCA TT, IGSCC Medium 0 2 2 -2.OOE-I I -2.OOE-I I -2.OOE-12 -2.OOE-12



3AR-1, REV. 0

Safety Break Location Failure Potential Inspections CDF Impact LERF ImpactSignificance DMs Rank SXI RIS B Delta w[POD w/o POD w/POD w/o POD

SI High PLOCA IGSCC Medium 0 3 3 -3.OOE-1I -3.OOE-I I -3.OOE-12 -3.OOE- 12

SI High LOCA None Low 2 6 4 -8.00E-10 -8.OOE-10 -8.00E-I I -8.00E-I I

SI High PLOCA None Low 12 2 -10 5.00E-12 5.OOE-12 5.00E-13 5.00E-13

SI Low Class 2 LSS FD Assume Medium 20 0 -20 1.00E-09 1.00E-09 1.00E-10 1.00E-10

SI Total -5.98E-08 -1.98E-08 -5.98E-09 -1.98E-09

Grand Total 324 112 -212 -2.29E-07 -4.68E-08 -2.29E-08 -4.68E-09

NotesSystems are described in Table 3.1

1. 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 3.7.1of EPRI TR-1 12657.

2. Only those RISB inspection locations that receive a volumetric examination are included in the count. Locations subjected toVT2 only are not credited in the count for risk impact assessment (there are none for Seabrook).

3. The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium", or "Low" dependingupon potential susceptibly to the various types of degradation. [Note: Low Safety Significant (LSS) locations wereconservatively assumed to be a rank of Medium (i.e., "Assume Medium")

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



3AR-1, REV. 0

Table 4 Inspection Location Selections Comparison

System Safety Significance Break Failure Potential Code Weld Section X! Code Case N716

High Low Location DMs Rank Category Count Vol Surface RISB Other

CBS " Class 2 LSS FD Assume Medium C-F-I 354 27 0 NA

CS " LOCA TT Medium B-J 6 0 4 6 NA

CS " PLOCA/ILOCA TT Medium B-J 33 0 3 2 NA

CS LOCA None Low B-J 4 0 8 0 NA

CS " ILOCA None Low B-J 33 0 0 0 NA

CS Class 2 LSS FD Assume Medium C-F-1 692 38 14 0 NA

FW SLB None Low C-F-2 141 41 5 13 NA

FW _ _ SLB TASCS Medium C-F-2 12 0 0 3 NA

FW " Class 2 LSS Assume Medium C-F-2 39 0 0 0 NA

MS V SLB None Low C-F-2 126 33 10 13 NA

MS " Class 2 LSS Assume Medium C-F-2 164 5 0 0 NA

RC LOCA TT Medium B-J 35 7 7 7 NA

RC _ _ LOCA TT,TASCS Medium B-J 15 6 0 11 NA

RC LOCA TASCS Medium B-i 16 6 0 4 NARC " LOCA None Low B-F, B-J 209 60 30 12 NA

RC " PLOCA/ILOCA None Low B-J,C-F- 1 62 0 I1 0 NA

RC " Class 2 LSS Assume Medium C-F-1 64 6 0 0 NARH V LOCA TASCS Medium B-J 6 4 0 5 NA

RH " LOCA None Low B-J 6 0 0 4 NARHI PLOCA None Low B-J,C-F-1 164 34 0 9 NA

RH __ _Class 2 LSS Assume Medium C-F-i 290 18 1 0 NA

SI LOCA TT Medium B-J 8 5 0 7 NA

SI V/ LOCA TT, TASCS Medium B-I 8 0 2 3 NASI " PLOCA TT, IGSCC Medium B-J 12 0 0 2 NA

SI V PLOCA IGSCC Medium B-I 6 0 0 3 NA

SI V LOCA None Low B-J 26 2 4 6 NA

SI v PLOCA None Low B-I 167 12 30 2 NASI _" Class 2 LSS Assume Medium C-F-1 369 20 8 0 NA



3AR-1, REV. 0

Notes -1. Systems are described in Table 3.12. 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 Bthrough G) in a BWR to be credited toward the 10% requirement. This option is not applicable for the Seabrook RIS_Bapplication. The "Other" column has been retained in this table solely for uniformity purposes with other RISB applicationtemplate submittals and to indicate when RIS_B selections will receive a VT-2 examination (these are not credited in riskimpact assessment), but there are no such cases for Seabrook.

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



3AR-1, REV. 0

Attachment A to Seabrook N-716 Template

Consideration of the Adequacy of

Probabilistic Risk Assessment Model for

Application of Code Case N716



3AR-1, REV. 0

1.0 Introduction

This attachment. summarizes the assessment of the Seabrook PRA capability as measured againstthe current ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009), endorsed by NRC RegGuide1.200 Rev 2. The Seabrook PRA capability assessment is based on industry peer reviews andinternal self assessments, documented in Seabrook Engineering EvaluationEE-1 1-026, Seabrook PRA Capability Assessment (Reference 1). The self-assessment is basedon the latest Seabrook PRA SSPSS-2011 Update, which includes the most recent internal floodevents risk analysis.

In addition, an assessment is made against the modeling supporting requirements asrecommended in EPRI Report 1021467, "Nondestructive Evaluation: Probabilistic RiskAssessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs",Technical Update (Reference 2). In summary, the Seabrook PRA SSPSS-2011 fully meets allthe ASME Supporting Requirements (SR) recommended by EPRI Report 1021467 for Risk-Informed In-Service Inspection Programs.

2.0 Background

2.1 RG1.200 & PRA Standard

The ASME / ANS PRA Standard (ASME/ANS RA-Sa-2009) has eight "parts" with technicalelements, high level requirements (HLRs), and detailed supporting requirements (SRs). Theseparts represent the major classes of hazards included in a PRA: internal events (Part 2), internalflood (Part 3), internal fire (Part 4), seismic events (Part 5), and other external hazard events(Parts 6 to 9). Note, Part 1 is introductory information and does not contain any requirements(except configuration control). Seabrook PRA model maintenance and configuration control isaddressed below in Section 4.0). NRC RegGuide 1.200 Rev 2 endorses the ASME PRAStandard with minor "clarifications." The NRC clarifications are considered in the self-assessment.

The following sections summarize the capability of the Seabrook PRA for the major Standardparts as related to EPRI Report 1021467 recommendations for the RI-ISI applications.

3.0 Assessment of ASME Supporting Requirements (SR) on RI-ISI Application

3.1 Assessment of Part 2 Internal Events on RI-ISI Application

The internal events PRA model was verified through the self-assessment to meet all of therecommended modeling supporting requirements (SRs) identified in EPRI Report 1021467.Therefore, the PRA is fully capable of identifying the internal events and CCDP and CLERPmetrics used for estimating the bounding CCDP/CLERP inputs to the RI-ISI evaluation.



3AR-1, REV. 0

3.2 Assessment of Part 3 Internal Flood Events on RI-ISI Application

A peer review was conducted in 2009 of the Internal Flood PRA, using two industry experts. The peerreview resulted in 26 findings and observations that have been further categorized into significance levelsas follows: 3 "B" level, 23 "C/D" level F&Os. There were no level "A" significance findings. Thesignificance levels are used to assess each finding for its potential impact on the PRA model. Thesignificance levels are defined as follows:

"A" Finding: MAJOR model weakness. Extremely important and necessary to address to ensure thetechnical adequacy of the PSA, the quality of the PSA or the quality of the PSA update process. As notedabove, there were no level "A" internal flood F&Os.

"B" Finding: IMPORTANT plant or model change or model weakness. Important and necessary toaddress but may be deferred until the next PSA update.

"C" Observation: MINOR plant or model change or model error. Considered desirable to maintainmaximum flexibility in PSA applications and consistency in the industry, but not likely to significantly affectresults or insights.

"D" Observation: DOCUMENTATION Change Only. Editorial or minor technical item left to thediscretion of the host utility.

All 26 internal flooding peer review findings have been addressed in the SSPSS-2011 model update. Thethree "B" level findings and associated disposition are summarized in the table below. Other findings hadeither a minor impact on the model or were related to improving documentation.



3AR-1, REV. 0

Seabrook Station Internal Flood Peer Review F&Os, Significance Level BFinding & Observation (Level B) F&O Action Disposition

F&O 4-9 (IFQU-B3) Quantification A check of the data and assumptions used in the internalThe completeness of assumptions and sources of flooding study was performed for reasonableness and forThe ompeteess f asumtios an sorce ofidentification of additional uncertainties. Appendix 12.1 H,uncertainty in the pipe failure data (e.g., error factor, Uncertaties wasrito clariesu Areas ofapplicability of data), failure probability of doors, generic data Uncertainties, was revised to clarify/ensure areas of

and modeling choices needs to be reviewed against other uncertainty and important assumptions are adequately

industry studies. captured and characterized.

F&O 5-2 (IFSO-B3, IFSN-B3) Uncertainty in Flooding Events As mentioned in the disposition for F&O 4-9, Appendix

Appendix 12.1H acknowledges uncertainty in break flow 12.1 H, Uncertainties, was revised to clarify/ensure areas of

rate. Need to expand uncertainty review to discuss other uncertainty and important assumptions are adequately

source related uncertainties such as maintenance-induced captured and characterized. In addition, a sensitivity

events and potential, if any, source pressure or temperature evaluation was performed to conservatively determine the

impacts. Also, discuss potential for breaks or human risk significance of a postulated maximum CW flood event.

induced events greater than assigned (i.e., catastrophic CW The maximum CW break flow was estimated at

expansion joint failure could far exceed 56,000 gpm). approximately 300,000 gpm.

Potential for larger floods can represent key insights. A door failure evaluation was performed to estimate theSpecifically, CW flood rates greater than 56,000 gpm could capacity of the various door configurations at Seabrook.represent a more significant threat to the Essential Doors C102, C101 and C100 provide an interface betweenSwitchgear rooms due to the configuration at Seabrook. the TB and ESWGR-A. The door evaluation indicates that

the capacity of these types of doors loaded against thejam/frame is in excess of any credible flood height in the TB.In addition, other doors in the Turbine Building are expectedto fail at considerably less water height - approximately 10feet (or less) and there is an unlatched door on the east sidenear condensate polishing that opens out. The benefit ofthis door was not credited. Once a flood height of -10 ft orless is achieved, failure of these other doors (which includesthe rollup doors, glass sliding door, misc. double doors) isexpected to vent the flood water to outdoors and result in asteady-state water level in the TB of -4 feet. It is noted thatthis TB flooding scenario is likely to cause a loss of offsitepower or fail non-essential electrical buses, resulting in a tripof the flooding source - the CW pumps long before there ispropagation impact in the essential switchgear rooms.

Based on the above, a conservative flood scenario wasdeveloped as sensitivity case FOTCWS. Based on thissensitivity case, the CDF from a postulated maximum CWbreak event in the TB is approximately 1E-09/yr. Thisscenario is screened from further detailed evaluation usingcriterion QN4a - Specific flood source in a flood area withCDF < -le-9 per yr based on flood-initiated accidentsequences from a specific flood source in the flood area.This assessment is conservative. Realistic modeling wouldeliminate conservatisms and further reduce the impacts.

F&O 5-3 (IFSN-A2) Door Failure Capacity (MC#772) A structural evaluation of typical doors at Seabrook Station

The assessment indicates that there are some "rugged" was performed and documented in a calculation, "Structural

doors capable of withstanding a water-height of 6-7 feet. Evaluation of Door Capacity Under Flooding Loading

These were walked-down for the peer review and they are Conditions". The evaluation was performed for 3 "typical"-

indeed rugged in appearance. However, there is limited type doors including: (1) rugged security door, (2) industrial

basis for door capacity other than "Industry Sources" which 3 hour rated fire door, and (3) double-wide industrial door

include a PWR OG e-mail. The EPRI Flood Guideline says with and without a center locking pin. The evaluationthe following: If there are doors within the boundaries of the addressed the difference in potential failure when each typearea then the following guidance can be applied: of door is loaded against its frame/jamb (stronger door

configuration) verses being loaded against its latch and

Water tight doors should be considered as failing only hinges (weaker door configuration). It is noted that the doorthrough human actions. If the door is alarmed its failure frames at Seabrook are embedded into the adjacentprobability can be considered to be zero. If the door is not concrete and are not supported by installed anchor bolts.alarmed then assume the normal egress failure condition of



3AR-1, REV. 0

Finding & Observation (Level B) F&O Action Dispositiona door opening out of the flood area if the water tight door This represents a much stronger configuration than aopens out of the area. If the water tight door opens into the conventionally installed frame with anchor bolts.area then consider the failure probability to be zero. Door capacity/failure insights from the structural evaluation

Normal egress and fire doors should be considered failed are included in Appendix 12.1A, Methodology. Door failuresafter 3 foot of flood level if the door opens into the area. and the resultant propagation are assessed on an individual

Normal egress and fire doors should be considered failed door/scenario basis. If the scenario's flood water heightafter 1 foot of flood level if the door opens out of the flood does not exceed the door's capacity, the door is notarea. expected to fail, is assumed to remain intact with only gap

leakage contributing to propagation. On the contrary, if the

The 1 and 3 foot EPRI Guideline should be used unless a scenario's flood water height exceeds the door capacity,higher value can be justified. While the doors are clearly door failure is assumed and the resulting propagation is viarugged, some more detailed justification should be the failed (open) door. No credit is given for failure of apresented. barrier to limit the flood consequence without some

assessment of the door failure potential.

The internal flood events PRA model and self-assessment were reviewed and found to meet allof the recommended modeling supporting requirements (SRs) identified in EPRI Report1021467. Therefore, the internal flood events PRA is fully capable of identifying the internalevents metrics used for estimating the bounding CCDP/CLERP inputs to the RI-ISI evaluationand for identifying flood source piping systems those pressure boundary failure would contribute>1E-06/yr CDF and >lE-07 LERF.



3AR-1, REV. 0

3.3 Other Hazard Groups - Internal Fires and External Hazards

3.3.1 Assessment of Part 4 Internal Fire on RI-ISI Application

The internal fire portion of the Seabrook PRA (FPRA) has been updated several times, includingthe most recent 2004 revision of the 1992 IPEEE Report (which was an update of the originalSSPSA-1983). This recent update used then-current methods, but was performed before the mostrecent technical guidance in NUREG/CR-6850. The fire PRA was subject to external review byindustry experts.

The internal fire PRA results are judged to not significantly influence the RI-ISI applicationconclusions. As provided in EPRI Report 1021467, the potential contribution of piping failure tointernal fire risk is judged negligible as the failure probability of piping is insignificant comparedto the failure probability of other systems, structures and components. Fire events are also notlikely 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 thisconclusion remains valid. ISI is an integral part of defense in depth, and the RI-ISI process willmaintain the basic intent of ISI (i.e., identifying and repairing flaws), and thus providereasonable assurance of an ongoing substantive assessment of piping condition. In addition,there are no changes to design basis events and safety margins are maintained.

3.3.2 Assessment of Part 5 Seismic Events on RI-ISI Application

The seismic events portion of the Seabrook PRA (SPRA) has been updated several times,including the most recent 2005 revision of the 1992 IPEEE Report (which was an update of theoriginal SSPSA-1983). This recent update used current methods to address issues related toequipment and operator fragility and the revised hazard spectrum but it did not include an updateto the seismic hazard curve. The seismic PRA was subject to external review by industryexperts.

The seismic events PRA results are judged to not significantly influence the RI-ISI applicationconclusions. As provided in EPRI Report 1021467, well engineered systems and structures areseismically 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 thanthe screening values typically used in seismic assessment and are not considered likely to failduring a seismic event. ISI is not considered in establishing fragilities of such SSCs. Meetingdefense in depth and safety margin principles provides assurance that this conclusion will remainvalid. ISI is an integral part of defense in depth, and the RI-ISI process will maintain the basicintent of ISI (i.e., identifying and repairing flaws), and thus provide reasonable assurance of anongoing substantive assessment of piping condition. In addition, there are no changes to designbasis events and safety margins are maintained.



3AR-1, REV. 0

3.3.3 Assessment of Parts 6 - 10 Other External Events on RI-ISI Application

The analysis of other hazards was performed for the original SSPSA-1983 and was updated forthe 1992 IPEEE Report. This portion of the Seabrook PRA has not been subject to formal selfassessment or peer review.

The PRA assessment of high winds, external floods and other external hazards are judged to notsignificantly influence the RI-ISI application conclusions. As provided in EPRI Report 1021467,the purpose of developing a RI-ISI program is to define an alternative inservice inspectionstrategy for piping systems. Other hazards such as high wind and external floods, are notconsidered in the development of an inservice inspection program for piping. The reasons forthis include: the structural ruggedness of the piping system, location, as relevant systems aretypically inside well engineered structures, and the consequence assessment for internal eventsalready includes the consideration of spatial impacts. In addition, the substantial industryexperience with plants implementing RI-ISI programs has not identified changes based uponinsight from the evaluation of other external hazards. The very small potential impact on thepotential for piping failure of a RI-ISI process, and the approaches to maintaining defense indepth and safety margins, provide confidence in this conclusion.

3.3.4 General Conclusion for Internal Fire Events and External Hazards

Quantification of other hazard groups will not change the conclusions derived from the RI-ISIprocess. As such, EPRI Report 1021467 guidance on meeting Regulatory Guide 1.200, RevisionI and Regulatory Guide 1.174 is sufficient for developing RI-ISI programs. Based onRegulatory Guide 1.174:

* The magnitude of the potential risk impact is not significant,

• Traditional engineering arguments including defense in depth and safety marginare applied, and

• Inclusion of other hazard groups would not affect the decision; that is, they wouldnot alter the results of the comparison with the acceptance guidelines.

4.0 PRA Model Maintenance & Control

Seabrook Station PRA Group instructions define the process of maintaining and updating theSeabrook Station PRA model. The process is consistent with the requirements of theASME/ANS PRA Standard and ensures that the PRA accurately reflects the current SeabrookStation plant design, operation and performance, and that the PRA remains consistent withcurrent risk technology and modeling. A general description of the configuration control processis as follows:



3AR-1, REV. 0

(a) Monitor PRA inputs for new information. This includes monitoring changes to SeabrookStation plant design and operation, monitoring Seabrook Station and industry operatingexperience, and changes in PRA technology and modeling.

(b) Record applicable new information. Applicable new information that has the potential toimpact the PRA model is recorded in the Model Change Database (MCDB). TheseMCDB entries form the content of the next PRA revision. Until close-out, these recordsare pending changes against the PRA model of record.

(c) Assess the significance of new information. The significance of the new information isreviewed with regard to its impact on the PRA model, including cumulative impacts frompending changes. This process identifies the need for a prompt focused PRA revisionverses periodic PRA revision, and the need for PRA upgrade (with Peer Review) versesPRA maintenance.

(d) Perform the PRA revision. The PRA is revised to evaluate the new information andincorporate the model changes identified in the MCDB as appropriate. Control of PRArevisions is provided in PRA Group Instructions. A "periodic" revision to the PRAmodel is performed at least once every three cycles (-4.5 years) to address open items inthe MCDB as well as incorporate any changes in plant design and operations; and toreflect operating experience.

Each model change documented in the MCDB requires an independent technical review.The review of each model change is documented in the "Disposition of Change" fieldwithin the MCDB. The purpose of the independent review is to verify that the modelchange was performed correctly and adequately reflects the plant or data change. Thereview may consist of a point-by-point check or an audit of calculations, analysis anddocumentation. Note that for PRA changes judged to be PRA "upgrades" (newmethodology or significant change in scope or capability), a formal peer review would berequired in addition to the independent technical review.

The PRA model documentation is updated as applicable for each update. The SeabrookStation PRA documentation consists of three levels (or tiers). Tier 1 is an ExecutiveSummary, a high level report appropriate for plant management. Tier 2 is thecomprehensive documentation of the model at a level adequate for an external reviewerto understand the basis for the risk from Seabrook Station. Tier 2 consists of the detailedsystems notebooks, data notebooks, and RISKMAN model file reports for event tree rulesand master frequency file. Tier 3 consists of spreadsheets, data bases, and other detailedcalculations and reports as well as the RISKMAN computer model itself. This level isadequate for an external reviewer to be able to reproduce any of the risk results. Tier 2and 3 comprise the controlled risk model.

(e) Control of computer codes and models. Control of computer codes and models isprovided in PRA Group instructions. These instructions provide guidance formaintaining the computer codes that form the basis of the Seabrook Station PRA and



3AR-1, REV. 0

risk-informed applications for both vendor-provided software and in-house software.The PRA computer codes are controlled and maintained to meet requirements of theNextEra Energy corporate Software Quality Assurance Program including: classificationof PRA software, identification of associated SQA requirements, and control of computercode configuration.

Seabrook Station PRA staff members have many years of plant engineering, operations and PRAexperience. PRA qualification is performed as part of the Engineering Support PersonnelTraining Program (ESP) for the duty area of Risk Management Engineer/Analyst Engineering.

5.0 References

1. Seabrook Station Engineering Evaluation EE- 11-026, Seabrook PRA CapabilityAssessment, October 2011, Revision 0.

2. EPRI Report 1021467, "Nondestructive Evaluation: Probabilistic Risks AssessmentTechnical Adequacy Guidance for Risk-Informed In-Service Inspection Programs",Technical Update, July 2010

Attachment 2

Regulatory Commitment

Regulatory Commitment

The following table identifies those actions committed to by NextEra Energy Seabrook, LLC inthis document. Any other statements in this submittal are provided for infornation purposes andare not considered to be regulatory commitments. Please direct questions regardingcommitments to Mr. Michael O'Keefe, Licensing Manager.

Regulatory Commitment Due Date / Event

NextEra internal flood risk assessment SSPSS-201 I identified that4" and 6" diameter Fire Protection piping segments located in theControl Building stairwell contributed greater than 1 E-06/yr to thecore damage frequency. Therefore, NextEra is committing to aprudent risk management measure to reduce the Fire Protectionflooding risk in the Control Building. The modification will limitthe postulated maximum fire protection break flow rate in theControl Building, and thus reduce the flood risk from the FireProtection pipe segments to less than 1E-06/yr. These fireprotection piping segments are, therefore, not included in theRISB scope. This modification will be completed prior toimplementation of risk-informed inservice inspections.

prior to implementation of risk-informed inservice inspections

NEXTera EN ERGY - NRC

Documents

Transcript of NEXTera EN ERGY - NRC