SBK-L-11197, Request for Alternative to Use ASME Code Case N-716 to Implement Risk-Informed Inservice Inspection Program

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

ML11319A016
Person / Time
Site:	Seabrook
Issue date:	11/07/2011
From:	Freeman P NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-11197
Download: ML11319A016 (48)

Text

NEXTera EN ERGY November 7, 2011 Docket No. 50-443 SBK-L-1 1197 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 Seabrook Station Request for Alternative to Use ASME Code Case N-716 to Implement Risk-Informed Inservice Inspection Program Pursuant to 10 CFR 50.55a(a)(3)(i), NextEra Energy Seabrook, LLC (NextEra) requests authorization 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 Code Category 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 in accordance with Code Case N-716, "Alternative Piping Classification and Examination Requirements."

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

ASME Code Case N-716 is founded, in large part, on the RI-ISI process as described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, "Revised Risk-Informed Decisionmaking Inservice Inspection of Piping," December 1999, which was previously reviewed and approved by the NRC (ADAMS Accession No. ML013470102). 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."

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

U.S. Nuclear Regulatory Conmnission SBK-L-11197 Page 2 to this letter contains a commitment regarding fire protection piping segments related 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. Freeman Site Vice President cc:

NRC Region I Administrator G. E. Miller, NRC Project Manager W. J. Raymond, NRC Resident Inspector Attachments NextEra Energy Seabrook, LLC Request for Approval of Risk-Informed/Safety Based Inservice Inspection Alternative for Class 1 and 2 Piping 3AR-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 Inservice Inspection Program

1. ASME Code Components Affected Code Class:

1 and 2 Examination 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 Inservice Inspection of Nuclear Power Plant Components, 2004 Edition. 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.

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 Category B-J welds. Similarly, IWC-2200, IWC-2420, IWC-2430, and IWC-2500 provide the examination 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 an alternate to the current ISI 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. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure, December 1999 (ADAMS Accession No. MLO 13470102) which was previously reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC).

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 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 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 RISB 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 ASME RA-Sb-2009, Standard for Probabilistic Risk Assessment for Nuclear Plant Applications.

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 Seabrook specific template for the application of ASME Code Case N-716 attached.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 TEMPLATE SUBMITTAL APPLICATION OF ASME CODE CASE N-716 RISK-INFORMED/SAFETY-BASED (RISB)

INSER VICE INSPECTION PROGRAM PLAN

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Technical Acronyms/Definitions Used in the Template 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 DM E-C ECSCC EOOS FAC F&O FLB FT FW HELB HEP HFE HR HRA HSS IE IF IFIV IGSSC ILOCA Alternating Current Auxiliary Feedwater Accident Sequence Analysis Accident Sequence Evaluation Program American Society of Mechanical Engineers Anticipated Transient Without Trip Break Exclusion Region 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Technical Acronyms/Definitions Used in the Template (cont'd.)

IPE Individual Plant Evaluation ISEAL Isolable RCP seal injection line LOCA LE LERF Analysis LERF Large Early Release Frequency LOCA Loss of Coolant Accident LOSP Loss of Off-Site Power LSS Low Safety-Significant MAAP Modular Accident Analysis Program MIC Microbiologically-Influenced Corrosion MOV Motor Operated Valve MS Main Steam MU Model Update NDE Nondestructive Examination NNS Non-Nuclear Safety NPS Nominal Pipe Size PBF Pressure Boundary Failure PIT Pitting PLOCA Potential Loss of Coolant Accident POD Probability of Detection PRA Probabilistic Risk Assessment PSA Probabilistic Safety Assessment PWSCC Primary Water SCC QU Quantification RC Reactor Coolant RCP Reactor Coolant Pump RCPB Reactor Coolant Pressure Boundary RG Regulatory Guide RHR Residual Heat Removal RI-BER Risk-Informed Break Exclusion Region RI-ISI Risk-Informed Inservice Inspection RISB Risk-Informed/Safety Based Inservice Inspection RM Risk Management RPV Reactor Pressure Vessel SBO Station Blackout SC Success Criteria SDC Shutdown Cooling SEAL RCP Seal Injection Line LOCA SLB Steam Line Break SGTR Steam Generator Tube Rupture SSBI Main Steam or Feedwater Break inside the Outer CIV SSBO Main Steam or Feedwater Break Beyond the Outer CIV SSC Systems, Structures, and Components SR Supporting Requirements

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Technical Acronyms/Definitions Used in the Template (cont'd.)

SW Service Water SXI Section XI SY Systems Analysis TASCS Thermal Stratification, Cycling, and Striping TGSCC Transgranular Stress Corrosion Cracking TR Technical Report TT Thermal Transients Vol Volumetric

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 and Pressure Vessel Section XI Code for Inspection Program B. NextEra plans to implement a risk-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 for Examination 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 Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk-Informed Inservice Inspection 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 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 Safety Assessment (PSA) Quality The methodology in Code Case N-716 provides for examination of a generic population of high safety significant (HSS) segments, supplemented with a rigorous internal flood events risk analysis to identify if any plant-specific HSS segments need to be added. Satisfying the requirement for the plant-specific analysis requires confidence that the internal flood events PRA is capable of successfully identifying any significant flooding contributors that are not identified in the generic population.

The Seabrook PRA used to support the risk aspects of this ASME Section XI Code Case N-716 evaluation is a full scope Level I and Level 2 integrated analysis for at power conditions. The fidelity and technical adequacy of the Seabrook PRA and supporting documentation are maintained through a formalized process of maintaining and periodically updating the PRA and by performing self-assessments and independent peer reviews. The latest capability assessment of the Seabrook PRA was measured 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 upgraded internal flood events risk assessment. The upgraded internal flood events model has undergone Peer Review.

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

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

The following subsections address (a) the internal flood events risk upgrade and (b) the use 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 safety significant welds.

ASME Code Case N-716, Section 2(a)(5) requires HSS classification for any piping segment whose contributions to core damage frequency is greater than I E-06 based upon 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 piping segments are applicable to the high safety significant categorization of this Case.

Seabrook internal flood events risk analysis identified that 4" and 6" diameter Fire Protection piping segments located in the Control Building stairwell contributed greater than 1 E-06/yr to the core damage frequency. This piping supplies fire water hose stations located within the Control Building and Diesel Generator Building stairwells.

A postulated pipe break in the Control Building stairwell greater than the design basis size has the potential to propagate to the essential switchgear rooms, impacting essential electrical power.

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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 the RCA Access Walkway, upstream of the Control Building stairwell. Installing an orifice at this location will limit the postulated maximum Fire Protection break flow rate in the Control Building, and thus reduce the flood risk from the Fire Protection pipe segments to less than 1 E-06/yr. This would eliminate the Fire Protection piping from the RIS_B scope.

Following installation of the Fire Protection orifice, there will be no piping segments that contribute greater than 1 E-06 to the core damage frequency or greater than I E-07 to large early release frequency. No additional high safety significant welds were identified 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 of flood 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 to consider the action highly reliable. These attributes are formulated into conservative screening criteria, which are applied in the Seabrook internal flood risk assessment.

Highly reliable actions apply to most small floods that have no spray impact (spray impact would be expected within the first few minutes). For large floods, highly reliable 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 min available before plant damage. These scenarios must have a control room cue of the flood and must have procedures that provide some direction in flood mitigation. If the mitigation 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 judged to be less than 2E-4 and the highly reliable action screening criteria applied in the Seabrook flood risk assessment include the following:

a. For small floods (where the cues may be more subtle), highly reliable actions must have at least 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for mitigation action.

b. For large floods with generic cues, procedures, and training, highly reliable actions must have at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for mitigation action.

c. For large floods with specific cues, procedures, and training, highly reliable actions must have at least 1 hr for mitigation action. The cue, procedure, and training are specific for flood scenarios where clear cues point to specific AOPs that 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 and LERF scoping guideline in Code Case N-716.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components.

The alternative RISB 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-I 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.

2.2 Augmented Programs The impact of the RISB 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 has not been revised by this application. A separate evaluation and program is maintained in accordance with the risk-informed break exclusion region methodology (RI-BER) described in EPRI Report 1006937, Extension of EPRI Risk Informed ISI Methodology to Break Exclusion Region Programs.

• A plant augmented inspection program has been implemented 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 RISB process and is subsumed by the RMS_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 82/182 welds has occurred at pressurized water reactors. For Seabrook, 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-

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 end welds and four 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 (and the adjacent stainless steel 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.

Seabrook has selected the four RPV hot leg nozzle Alloy 82/182 welds for ultrasonic examination for PWSCC within the scope of Code Case N-716. Code Case N-716 requires examination of these welds every ten years. However, the examination frequency of butt welds is currently established by NRC rulemaking.

The RIS_B Program will not be used to eliminate any welds with overriding 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 ISI 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:

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 and instrumentation diagrams and additional plant information, including the existing plant ISI 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);

(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 normal shutdown cooling flow path either:

(a) As part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or (b) Other systems or portions of systems from the RPV to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; (3)

That portion of the Class 2 feedwater system [> 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 (BER) 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, but all BER piping at Seabrook is Class 2.

(5)

Any piping segment whose contribution to Core Damage Frequency (CDF) is greater than I E-06 [and per NRC feedback on the Grand Gulf and D. C. Cook RISB 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. No piping segments with a contribution to CDF greater than 1 E-06 (ME-07 for LERF) were identified.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 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-1 12657 (i.e., 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.

As previously approved for Seabrook during last interval, a deviation to the EPRI RISB methodology has been implemented in the failure potential assessment. Table 3-16 of EPRI TR-l 12657 contains the following criteria for assessing the potential for Thermal Stratification, Cycling, and Striping (TASCS). 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 fluid with turbulent flow; AND AT > 500F, AND Richardson Number > 4 (this value predicts the potential buoyancy of a stratified flow)

These criteria, based on meeting a high cycle fatigue endurance limit with the actual AT assumed equal to the greatest potential AT 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 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, significant top-to-bottom cyclic ATs 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.

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 TASCS 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 (TT) 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.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 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.

SConvection 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 TASCS 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 as well as Seabrook during the past interval. The methodology used in the Seabrook RISB application for assessing TASCS potential conforms to these updated criteria. Additionally, materials reliability program (MRP)

MRP-146 guidance on the subject of TASCS was also incorporated into the Seabrook RIS_B application.

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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 (3)

For the RCPB, at least two-thirds of the examinations shall be located between the inside first isolation valve (IFIV) (i.e., 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 Seabrook) shall be selected.

(5)

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

In contrast to a number of traditional RI-ISI 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 selections are presented in Table 3.3. Section 4 of EPRI TR-l 12657 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.

Class 1 Welds~ll Class 2 Welds*2*

All Piping Welds*3 )

Total Selected Total Selected Total Selected 750 78 2317 34 3067 112 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-2 locations. Of the Class 2 piping weld locations, 345 are HSS; the remaining are LSS.

(3) 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 RISB Program.

3.3.1 Current Examinations Seabrook is currently using the traditional ASME Section XI inspection methodology for ISI examination of Class 2 piping welds per the 2004 Edition of ASME Section XI and for Class I piping welds, the NRC has previously approved an application using EPRI-TR 112657B-A.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 3.3.2 Successive Examinations If indications are detected during RISB 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 NRC rule making.

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 (i.e., 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 NRC rule making.

3.3.4 Program Relief Requests Consistent with previously approved RISB submittals, Seabrook 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 for those cases where greater than 90% coverage is not obtained will be submitted per 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 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 RISB degradation mechanism assessment. For example, examinations of locations subject to thermal fatigue will be 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-1 12657 process for risk impact analyses, whereby limits are imposed to ensure that the change-in-risk of implementing the RISB Program meets the requirements of Regulatory Guides 1.174 and 1.178. Section 3.7.2 of EPRI TR-112657 requires that the cumulative change in CDF and LERF be less than 1E-07 and I E-08 per year per system, respectively.

For LSS welds, Conditional Core Damage Probability (CCDP)/Conditional Large Early Release Probability (CLERP) values of 1 E-4/1E-5 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 ISI (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)/1E-5 (CLERP) and between Medium and Low consequence categories are 1E-6 (CCDP)/1E-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 lE-5 to 3E-5 due to an update, it will remain below the I E-4 threshold value; the change-in-risk evaluation would not require updating.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 The updated internal flooding PRA was also reviewed to ensure that there is no LSS Class 2 piping with a CCDP/CLERP greater than 1E-4/1E-5. This review identified some piping in the RHR, CBS, SI and CVCS systems located outside of containment with a CCDP greater than 1 E-4. As a result, all LSS welds in these systems are conservatively assigned CCDP/CLERP equal to 5E-4/5E-5.

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 potential ("Assume Medium" in Table 3.4) for use in the change-in-risk assessment. Experience with previous industry RIS_B applications shows this to be conservative.

Seabrook 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-1 12657. 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., Large LOCA CCDP bounds the medium and small LOCA CCDPs).

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Estimated Consequence Upper / Lower Bound Break Location Description of Affected Piping CCDP CLERP Rank CCDP CLERP LOCA 4E-02 4E-03(U4E0

()4-3 HIGH (U) 4E-02 (U) 4E-03 Unisolable RCPB piping of all sizes The 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 RC close 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 inside probability 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-05 outside 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 systems simplify 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 safety Estimated 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 safety CBS, 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 systems and 0.1 margin for CLERP applies)

Note: The PRA does not explicitly model potential (PLOCA) and isolable (ILOCA) 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. The upper bound ILOCA CCDP 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.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 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 x0 and is expected to have a value less than I 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-l 12657.

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 RISB approach.

Table 3.4 presents a summary of the RISB Program versus the 1995 Edition through 1996 Addenda of ASME Section XI program requirements on a "per system" basis for the second interval. The presence of FAC was adjusted for in the quantitative analysis by excluding its impact on the failure potential rank.

The exclusion of the impact of FAC on the failure potential rank and therefore in the determination of the change-in-risk, was performed because FAC is a damage mechanism managed by a separate, independent plant augmented inspection program. The RISB Program credits and relies upon this plant augmented inspection program to manage this damage mechanism. The plant FAC program will continue to determine where and when examinations shall be performed. Hence, since the number of FAC examination locations remains the same "before" and "after" (the implementation of the RISB program) and no delta exists, there is no need to include the impact of FAC in the performance of the risk impact analysis.

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

With POD Credit Without POD Credit System Delta Delta Delta Delta CDF LERF CDF LERF Containment Spray (CBS) 1.35E-09 1.35E-10 1.35E-09 1.35E-10 CVCS (CS)

-4.13E-08

-4.13E-09

-2.21E-08

-2.21E-09 Feedwater (FW)

-4.OOE-11

-4.OOE-12

-1.60E-1I

-1.60E-12 Main Steam (MS) 6.OOE-1 6.OOE-12 6.OOE-1 1 6.OOE-12 Reactor Coolant (RC)

-1.03E-07

-1.03E-08

-2.34E-09

-2.34E-10 RHR (RH)

-2.63E-08

-2.63E-09

-3.89E-09

-3.89E-10 Safety 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 As shown in Table 3.4, new RIS B locations were selected such that the RIS B selections exceed the Section XI selections for certain 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 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 RISB 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 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 1, Category B-J Pressure Retaining Welds, this methodology has been ineffective in identifying leaks or failures. EPRI TR-1 12657 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 I 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 be pressure tested in accordance with the Code, regardless of its safety significance.

3.5 Implementation Upon approval of the RISB 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 third ISI interval. No changes to the

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 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 RISB process, as appropriate.

3.6 Feedback (Monitoring)

The RISB 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 NDE results, a review of site failure information from the corrective action program, and a review of industry failure information from industry operating experience (OE). 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).

E.

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

F.

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

G.

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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 For preservice examinations, Seabrook 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 ISI PLAN CHANGE Seabrook is currently in the first period of the third ISI interval and is using the traditional ASME Section XI inspection methodology for ISI 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 third ISI interval to ensure compliance with the traditional ASME Section XI inspection methodology during the transition to N-716.

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

Alternatively, first period examinations will be completed using the traditional ASME XI methodology. Then, the second and third period examinations will utilize the RIS_B methodology. In this case, approximately 1/3rd of the total number of RISB piping welds selected 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 any PWSCC examination requirements for the Alloy 82/182 examinations.

A comparison between the RISB Program and the 1995 Edition through 1996 Addenda of Section 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 Exclusion Region 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 Decisionmaking Inservice Inspection of Piping.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 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-Implement Risk-Informed ISI 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 ISI program for Class 1 and 2 Piping Welds, dated September 28, 2007.

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

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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Table 3.1 Code Case N-716 Safety Significance Determination System Weld N-716 Safety Significance Determination Safety Significance Count RCPB SDC PWR: FW BER CDF> IE-6 (1)

High Low CBS 354 76 V"

CS 7

692

83 V

50 V"

V" V

FW 50 20 V

39 126 V

MS 164

_/

285 V

V 41 V

V V

RC 11 V"

64 V

33 V

V 88 V,

V, V"

RH 55 V

V 290 V

161

/

V S1 66 V

V 369 V

555 V

V 195 V

SUMMARY

66 V

V RESULTS 83 FOR ALL SYSTEMS 50 V

v V

146 V

V 1972 V

TOTALS 3067 CBS = Containment Building Spray CS = Chemical Volume and Control System FW = Main Feedwater MS = Main Steam RC = Reactor Coolant RH = Residual Heat Removal SI = Safety Injection (1) A Fire Protection modification is being made to ensure that there is no piping that exceeds this criterion

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Table 3.2 Failure Potential Assessment Summaty Thermal Localized Flow Fatigue Stress Corrosion Cracking Corrosion Sensitive System(1" TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC CBS CS FW MS RC RH V

Notes:

1. Systems are described in Table 3.1

2. 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 and SI systems.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Table 3.3: Code Case N716 Selections Weld Count N716 Selection Considerations System Selections HSS LSS DMs RCPB RCPB (IFIV)

RCPB (OC)

BER CBS 354 None 0

CS 6

TT

/

6 CS 33 TT

/

2 CS 4

None V

0 CS 33 None V"

0 CS 692 None 0

FW 70 None

/

13 FW 12 TASCS 3

FW 71 None 0

FW 39 None 0

MS 126 None V

13 MS 164 None 0

RC 35 IT T V

7 RC 15 TT,TASCS I

V 11 RC 16 TASCS V

V 4

RC 209 None V

V 12 RC 51 None V

0 RC 11 None 0

RC 64 None 0

RH 6

TASCS V

5 RH 6

None V

V 4

RH 109 None V

4 RH 55 None 5

RH 290 None 0

S1 8

TT V

V 7

S1 8

TT,TASCS V

V 3

S1 12 TT, IGSCC V

2 S1 6

IGSCC 3

S 26 None V

V 6

SI 167 None V

2 S1 369 None 0

49 TT V

V 20 33 TT 2

23 TT, TASCS V

14 22 TASCS V

V 9

Summary 12 TT, IGSCC V

2 Results Rsl 6

IGSCC V

3 All Systems 245 None V

V 22 360 None V

6 196 None V

26 12 TASCS 3

137 None 5

Totals 1095 1972 None 112 Note: Systems are described in Table 3.1

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Table 3.4 Risk Impact Analysis Results Safety 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-09 FW 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+00 FW 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-12 MS 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Safety Break Location Failure Potential Inspections CDF Impact LERF Impact Significance 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 Notes Systems 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 3.7.1 of EPRI TR-1 12657.

2. Only those RISB 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 Seabrook).

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 (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 the five HSS criteria of Section 2(a) of N-716 (e.g., not part of the BER scope).

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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

NA RC 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

NA RH V

LOCA TASCS Medium B-J 6

4 0

5 NA RH LOCA None Low B-J 6

0 0

4 NA RHI 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 NA SI 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

NA SI Class 2 LSS Assume Medium C-F-1 369 20 8

0 NA

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Notes -

1. Systems are described in Table 3.1

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 Seabrook RIS_B application. The "Other" column has been retained in this table solely for uniformity purposes with other RISB 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 Seabrook.

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 (i.e., "Assume Medium").

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 1.0 Introduction This attachment. summarizes the assessment of the Seabrook PRA capability as measured against the current ASME/ANS PRA Standard (ASME/ANS RA-Sa-2009), endorsed by NRC RegGuide 1.200 Rev 2. The Seabrook PRA capability assessment is based on industry peer reviews and internal self assessments, documented in Seabrook Engineering Evaluation EE-1 1-026, Seabrook PRA Capability Assessment (Reference 1). The self-assessment is based on the latest Seabrook PRA SSPSS-2011 Update, which includes the most recent internal flood events risk analysis.

In addition, an assessment is made against the modeling supporting requirements as recommended in EPRI Report 1021467, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs",

Technical Update (Reference 2). In summary, the Seabrook PRA SSPSS-2011 fully meets all the 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 technical elements, high level requirements (HLRs), and detailed supporting requirements (SRs). These parts represent the major classes of hazards included in a PRA: internal events (Part 2), internal flood (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 is addressed below in Section 4.0). NRC RegGuide 1.200 Rev 2 endorses the ASME PRA Standard 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 Standard parts 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 the recommended modeling supporting requirements (SRs) identified in EPRI Report 1021467.

Therefore, the PRA is fully capable of identifying the internal events and CCDP and CLERP metrics used for estimating the bounding CCDP/CLERP inputs to the RI-ISI evaluation.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 peer review resulted in 26 findings and observations that have been further categorized into significance levels as follows: 3 "B" level, 23 "C/D" level F&Os. There were no level "A" significance findings. The significance levels are used to assess each finding for its potential impact on the PRA model. The significance levels are defined as follows:

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

"B" Finding:

IMPORTANT plant or model change or model weakness. Important and necessary to address but may be deferred until the next PSA update.

"C" Observation:

MINOR plant or model change or model error. Considered desirable to maintain maximum flexibility in PSA applications and consistency in the industry, but not likely to significantly affect results or insights.

"D" Observation:

DOCUMENTATION Change Only. Editorial or minor technical item left to the discretion of the host utility.

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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Seabrook Station Internal Flood Peer Review F&Os, Significance Level B Finding & Observation (Level B)

F&O Action Disposition F&O 4-9 (IFQU-B3) Quantification A check of the data and assumptions used in the internal The completeness of assumptions and sources of flooding study was performed for reasonableness and for The 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 of applicability 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 the Specifically, 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 between Switchgear 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 the jam/frame is in excess of any credible flood height in the TB.

In addition, other doors in the Turbine Building are expected to fail at considerably less water height - approximately 10 feet (or less) and there is an unlatched door on the east side near condensate polishing that opens out. The benefit of this door was not credited. Once a flood height of -10 ft or less is achieved, failure of these other doors (which includes the rollup doors, glass sliding door, misc. double doors) is expected to vent the flood water to outdoors and result in a steady-state water level in the TB of -4 feet. It is noted that this TB flooding scenario is likely to cause a loss of offsite power or fail non-essential electrical buses, resulting in a trip of the flooding source - the CW pumps long before there is propagation impact in the essential switchgear rooms.

Based on the above, a conservative flood scenario was developed as sensitivity case FOTCWS. Based on this sensitivity case, the CDF from a postulated maximum CW break event in the TB is approximately 1E-09/yr. This scenario is screened from further detailed evaluation using criterion QN4a - Specific flood source in a flood area with CDF < -le-9 per yr based on flood-initiated accident sequences from a specific flood source in the flood area.

This assessment is conservative. Realistic modeling would eliminate 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 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 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 evaluation the following: If there are doors within the boundaries of the addressed the difference in potential failure when each type area 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 door through human actions. If the door is alarmed its failure frames at Seabrook are embedded into the adjacent probability 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

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 Finding & Observation (Level B)

F&O Action Disposition a door opening out of the flood area if the water tight door This represents a much stronger configuration than a opens 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 failures after 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 height after 1 foot of flood level if the door opens out of the flood does not exceed the door's capacity, the door is not area.

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 via rugged, some more detailed justification should be the failed (open) door. No credit is given for failure of a presented.

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 all of the recommended modeling supporting requirements (SRs) identified in EPRI Report 1021467. Therefore, the internal flood events PRA is fully capable of identifying the internal events metrics used for estimating the bounding CCDP/CLERP inputs to the RI-ISI evaluation and for identifying flood source piping systems those pressure boundary failure would contribute

>1E-06/yr CDF and >lE-07 LERF.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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, including the most recent 2004 revision of the 1992 IPEEE Report (which was an update of the original SSPSA-1983). This recent update used then-current methods, but was performed before the most recent technical guidance in NUREG/CR-6850. The fire PRA was subject to external review by industry experts.

The internal fire PRA results are judged to not significantly influence the RI-ISI application conclusions. As provided in EPRI Report 1021467, the potential contribution of piping failure to internal fire risk is judged negligible as the failure probability of piping is insignificant compared to the failure probability of other systems, structures and components. 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 remains valid. ISI is an integral part of defense in depth, and the RI-ISI process will maintain the basic intent of ISI (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 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 the original SSPSA-1983). This recent update used current methods to address issues related to equipment and operator fragility and the revised hazard spectrum but it did not include an update to the seismic hazard curve. The seismic PRA was subject to external review by industry experts.

The seismic events PRA results are judged to not significantly influence the RI-ISI application conclusions. As provided in EPRI Report 1021467, well engineered systems and structures 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. ISI 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. ISI is an integral part of defense in depth, and the RI-ISI process will maintain the basic intent of ISI (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 safety margins are maintained.

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 for the 1992 IPEEE Report. This portion of the Seabrook PRA has not been subject to formal self assessment or peer review.

The PRA assessment of high winds, external floods and other external hazards are judged to not significantly 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 inspection strategy for piping systems. Other hazards such as high wind and external floods, are not considered in the development of an inservice inspection program for piping. The reasons for this include: the structural ruggedness of the piping system, location, as relevant systems are typically inside well engineered structures, and the consequence assessment for internal events already includes the consideration of spatial impacts. In addition, the substantial industry experience with plants implementing RI-ISI programs has not identified changes based upon insight from the evaluation of other external hazards. The very small potential impact on the potential for piping failure of a RI-ISI process, and the approaches to maintaining defense in depth 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-ISI process. As such, EPRI Report 1021467 guidance on meeting Regulatory Guide 1.200, Revision I and Regulatory Guide 1.174 is sufficient for developing RI-ISI programs. Based on Regulatory Guide 1.174:

• The magnitude of the potential risk impact is not significant, Traditional engineering arguments including defense in depth and safety margin are applied, and Inclusion of other hazard groups would not affect the decision; that is, they would not 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 the Seabrook Station PRA model. The process is consistent with the requirements of the ASME/ANS PRA Standard and ensures that the PRA accurately reflects the current Seabrook Station plant design, operation and performance, and that the PRA remains consistent with current risk technology and modeling. A general description of the configuration control process is as follows:

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3AR-1, REV. 0 (a)

Monitor PRA inputs for new information. This includes monitoring changes to Seabrook Station plant design and operation, monitoring Seabrook Station and industry operating experience, and changes in PRA technology and modeling.

(b)

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

(c)

Assess the significance of new information. The significance of the new information is reviewed with regard to its impact on the PRA model, including cumulative impacts from pending changes. This process identifies the need for a prompt focused PRA revision verses periodic PRA revision, and the need for PRA upgrade (with Peer Review) verses PRA maintenance.

(d)

Perform the PRA revision. The PRA is revised to evaluate the new information and incorporate the model changes identified in the MCDB as appropriate. Control of PRA revisions is provided in PRA Group Instructions. A "periodic" revision to the PRA model is performed at least once every three cycles (-4.5 years) to address open items in the MCDB as well as incorporate any changes in plant design and operations; and to reflect 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" field within the MCDB. The purpose of the independent review is to verify that the model change was performed correctly and adequately reflects the plant or data change. The review may consist of a point-by-point check or an audit of calculations, analysis and documentation. Note that for PRA changes judged to be PRA "upgrades" (new methodology or significant change in scope or capability), a formal peer review would be required in addition to the independent technical review.

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

(e)

Control of computer codes and models. Control of computer codes and models is provided in PRA Group instructions. These instructions provide guidance for maintaining the computer codes that form the basis of the Seabrook Station PRA and

NEXTERA ENERGY SEABROOK, LLC PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 the NextEra Energy corporate Software Quality Assurance Program including: classification of PRA software, identification of associated SQA requirements, and control of computer code configuration.

Seabrook Station PRA staff members have many years of plant engineering, operations and PRA experience. PRA qualification is performed as part of the Engineering Support Personnel Training 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 Capability Assessment, October 2011, Revision 0.

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

Technical Update, July 2010 Regulatory Commitment

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

Regulatory Commitment Due Date / Event NextEra internal flood risk assessment SSPSS-201 I identified that 4" and 6" diameter Fire Protection piping segments located in the Control Building stairwell contributed greater than 1 E-06/yr to the core damage frequency. Therefore, NextEra is committing to a prudent risk management measure to reduce the Fire Protection flooding risk in the Control Building. The modification will limit the postulated maximum fire protection break flow rate in the Control Building, and thus reduce the flood risk from the Fire Protection pipe segments to less than 1E-06/yr. These fire protection piping segments are, therefore, not included in the RISB scope. This modification will be completed prior to implementation of risk-informed inservice inspections.

prior to implementation of risk-informed inservice inspections