Text

Relief Request IP3-ISI-RR-05 for Fourth Ten-Year Inservice Inspection Interval
	ML12039A253
Person / Time
Site:	Indian Point
Issue date:	01/30/2012
From:	Robert Walpole; Entergy Nuclear Northeast
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-12-028
	Download: ML12039A253 (44)
	v • d • e

Enteray Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel 914 254 6700 Robert Walpole Licensing Manager Tel (914) 254-6710 NL-12-028 January 30, 2012 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Relief Request IP3-1SI-RR-05 For Fourth Ten-Year Inservice Inspection Interval Indian Point Unit Number 3 Docket No. 50-286 License No. DPR-64

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. (Entergy) is submitting Relief Request IP3-1SI-RR-05 (Enclosure

1) for the Indian Point Unit No. 3 (IP3) Fourth 10-year Inservice Inspection (ISI) Interval. The enclosed relief request is for the application of Code Case N-716, "Alternative Piping Classification and Examination Requirements", to implement a risk informed/safety based Inservice Inspection (ISI) as an alternative to the ASME Section XI Inservice Inspection requirements. The attached bases concludes this request provides an acceptable level of quality and safety. This relief is requested under the provisions of 10CFR 50.55a(a)(3)(i).

There are no new commitments identified in this submittal. Ifyou have any questions or require additional information, please contact Mr. Robert Walpole, Licensing Manager at 914-254-6710.

Sincerely, RW/sp . Relief Request IP3-1SI-RR-05 Proposed Alternative to Use ASME Code Case N-716 cc: Mr. John P. Boska, Senior Project Manager, NRC NRR DORL Mr. William Dean, Regional Administrator, NRC Region 1 NRC Resident Inspector, IP3 Mr. Francis J. Murray, Jr., President and CEO, NYSERDA Mr. Paul Eddy, New York State Dept. of Public Service

~\TQ

Enclosure 1 TO NL-12-028 RELIEF REQUEST IP3-ISI-RR-05 PROPOSED ALTERNATIVE TO USE ASME CODE CASE N-716 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Indian Point Unit 3 Fourth 10-Year ISI Interval Relief Request No: IP3-ISI-RR-05 Proposed Alternative to Use ASME Code Case N-716 Alternative Piping Classification And Examination Requirements Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

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

2. Applicable Codes Edition and Addenda The applicable Code edition and addenda is ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2001 Edition through the 2003 addenda. In addition, as required by 10 CFR50.55a, piping ultrasonic examinations are performed per ASME Section XI, 2001 Edition, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems.

3. Applicable Code Requirements For the current inservice inspection (ISI) program at Indian Point 3, 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-1 and C-F-2 welds.

4. Reason for the 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 1 and 2 piping.

5. Proposed Alternative and Basis for Use In lieu of the existing Code requirements, Indian Point 3 proposes to use a RISB process as an alternate to the current ISI program for Class 1 and 2 piping. The RISB process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements, Section Xl, 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. ML013470102) which was previously reviewed and approved by the US Nuclear Regulatory Commission (NRC).

In general, a risk-informed program replaces the number and locations of nondestructive examination (NDE) inspections based on ASME Code,Section XI requirements with the number and locations of these inspections based on the risk-informed guidelines. These processes result in a program consistent with the concept that, by focusing inspections on the most safety-

Relief Request IP3-ISI-RR-05 Page 2 of 2 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 Guide 1.174 and 1.178, this proposed RISB 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-220, IWB-2420, IWB-2430 and IWB-2500 (Examination Categories C-F-1 and C-F-2) is requested in accordance with 10CFR50.55a(a)(3)(i). An Indian Point Unit 3 specific relief request is attached that mirrors previous RISB submittals to the NRC.

6. Duration of Proposed Alternative Through July 20, 2019

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 and 2.

8. References Vogtle Electric Generating Plant Safety Evaluation - see ADAMS Accession No. ML100610470.

DC Cook Safety Evaluation - see ADAMS Accession No. ML072620553. Grand Gulf Nuclear Station Safety Evaluation - see ADAMS Accession No. ML072430005. Waterford-3 Safety Evaluation - see ADAMS Accession No. ML080980120.

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Technical Acronyms/Definitions Used in the Template AC Alternating Current AF Auxiliary Feedwater AS Accident Sequence Analysis ASEP Accident Sequence Evaluation Program ASME American Society of Mechanical Engineers BER Break Exclusion Region CAFTA Computer-Aided Fault Tree Analysis CC PRA abbreviation for Capacity Category CC Crevice Corrosion CCDP Conditional Core Damage Probability CCF Common Cause Failure CDF Core Damage Frequency CIV Containment Isolation Valve Class 2 LSS Class 2 Pipe Break in LSS Piping CLERP Conditional Large Early Release Probability CV Chemical Volume and Control System DA Data analysis DC Direct Current DM Degradation Mechanism E-C Erosion-Corrosion ECSCC External Chloride Stress Corrosion Cracking EOOS Equipment Out of Service FAC Flow-Accelerated Corrosion F&O Facts and Observations FLB Feedwater Line Break FT Fault tree FW Feedwater HELB High Energy Line Break (synonymous with BER)

HEP Human Error Probability HFE Human Failure Event HR Human Reliability HRA Human Reliability Analysis HSS High Safety-Significant IE Initiating Events Analysis IF Internal Flooding IFIV Inside First Isolation Valve IGSSC Intergranular Stress Corrosion Cracking ILOCA Isolable Loss of Coolant Accident IPE Individual Plant Evaluation LE LERF Analysis LERF Large Early Release Frequency LOCA Loss of Coolant Accident E1-1

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RIS_B PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Technical Acronyms/Definitions Used in the Template (Continued)

LOSP Loss of Off-Site Power LSS Low Safety-Significant MAAP Modular Accident Analysis Program MIC Microbiologically-lnfluenced 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, RH 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 SLB Steam Line Break SGTR Steam Generator Tube Rupture SSC Systems, Structures, and Components SR Supporting Requirements SW Service Water SXl Section Xl SY Systems Analysis TASCS Thermal Stratification, Cycling, and Striping TGSCC Transgranular Stress Corrosion Cracking TR Technical Report TT Thermal Transients Vol Volumetric El -2

ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM TEMPLATE PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table of Contents

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

2. Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section Xl 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 - Indian Point Unit 3 PRA Quality Review El -3

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1. INTRODUCTION Indian Point Unit 3 (IP3) is currently in the fourth inservice inspection (ISI) interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section Xl Code for Inspection Program B. Indian Point Unit 3 plans to implement a risk-informed/safety-based inservice inspection (RISB) program in the first Period of the fourth ISI interval. The fourth ISI interval began on July 21, 2009.

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

The RISB process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classificationand 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 ProbabilisticRisk Assessment in Risk-Informed Decisions On Plant-SpecificChanges to the Licensing Basis, and Regulatory Guide 1.178, An Approach for Plant-SpecificRisk-Informed Decision making 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 flooding 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 flooding PRA is capable of successfully identifying any significant flooding contributors that are not identified in the generic population.

The Indian Point Unit 3 PRA is based on a detailed model of the plant that was originally developed for the Individual Plant Examination (IPE) and Individual Plant Examination for External Events (IPEEE) projects. The IP3 internal events PRA model has been upgraded since the original IPE to meet the guidance of RG 1.200 Rev 2 "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," as well as the American Society of Mechanical Engineers and American National Standard (ASME/ANS) PRA Standard RA-Sa-2009.

A formal, PWROG-sponsored industry peer review of the upgraded internal events model was completed in December 2010. The peer review utilized the process described in Nuclear Energy Institute document NEI 05-04, "Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard," January 2005, and the ASME/ANS PRA Standard. This review confirmed that the PRA model met the requirements of RG 1.200, Revision 2, and ASME/ANS RA-Sa-2009. There were 11 findings identified by the peer review team.

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Attachment A contains a summary of these findings, including the status of the resolution for each finding and the potential impact of each finding on this application.

The IP3 PRA technical capability evaluations and the maintenance and update processes described above and Attachment A provide a robust basis for concluding that the IP3 PRA model is suitable for use in the risk-informed process used for this application.

External Events are addressed in Parts 4 through 9 of the ASME/ANS standard. The EPRI Report 1021467 proposes a qualitative treatment of the risk from fire events and from events that impose extreme loads on piping systems. The NRC Safety Evaluation concurred in the TR conclusion that challenges from fire events are expected to be less frequent and not significantly different than challenges caused by the random occurrence of internal initiating events. The NRC SE also concluded that additional analysis of extreme loading events are not needed and will not change the conclusion derived from the RI-ISI program.

2. PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS 2.1 ASME Section XI ASME Section Xl Examination Categories B-F, B-J, C-F-i, 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-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section XI Code will be unaffected.

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 RISB application scope (e.g.,

Class 1 and 2 piping).

A plant augmented inspection program has been implemented in response to NRC Bulletin 88-08, ThermalStressesin 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 RIS_B Program.

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

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

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 Indian Point Unit 3, the unmitigated Alloy 82/182 Category B-F dissimilar metal welds (greater than NPS 1) subject to PWSCC are the three RPV hot leg nozzle to safe-end welds and the three cold leg nozzle to safe-end welds. The Steam Generator dissimilar metal welds are not subject to PWSCC because the welds are Alloy 52/152, and all of the pressurizer dissimilar metal welds (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.

Even though Code Case N-716 only considers the RPV hot leg nozzle Alloy 82/182 weld locations to be susceptible to PWSCC, Indian Point Unit 3 has selected 4 welds to be ultrasonically examined for PWSCC within the scope of Code Case N-716. Code Case N-716 requires the examination of these welds every ten years. However, the examination frequency for these eight welds is currently based on the frequencies established by the requirements of Materials Reliability Program (MRP)-139, Revision 1. MRP-139 currently requires that the unmitigated hot legs be examined on a five year frequency and the unmitigated cold legs be examined on a six year frequency. These frequencies are subject to change based on factors such as industry experience and issuance of NRC rule making. The RISB Program will not be used to eliminate any MRP-139 or regulatory requirements. Indian Point Unit 3 plans to manage Alloy 82/182 welds per the requirements of Code Case N-770-1 once the program has been formally implemented in 2013.

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 RISB 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)

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Risk Impact Assessment (see Section 3.4)

" Implementation Program (see Section 3.5)

Feedback Loop (see Section 3.6)

Each of these six steps is discussed below:

3.1 Safety Significance Determination The systems assessed in the RISB 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 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flow path either:

(a) As part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or (b) Other systems or portions of systems from the RPV to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; (3) That portion of the Class 2 feedwater system [> 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. There is no BER augmented program at Indian Point Unit 3.

(5) Any piping segment whose contribution to Core Damage Frequency (CDF) is greater than 1 E-06 [and per NRC feedback on the Grand Gulf and D. C. Cook 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. Service water piping in the 480 Volt Switchgear Room was identified as HSS due to CDF exceeding the above criteria.

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 Indian Point Unit 3 during last interval, a deviation to the EPRI RISB methodology has been implemented in the failure potential assessment. Table 3-16 of EPRI TR-1 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 > 50'F, 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 addresses the potential for fluid cycling.

The impact of this additional consideration on the existing TASCS susceptibility criteria is presented below.

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RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

> 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 ifthe 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. However, since this is generally a "steady-state" phenomenon with no potential for cyclic temperature changes, the effect of TASCS is not significant and can be neglected.

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

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RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

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 RISB program submittals for D. C. Cook, Grand Gulf Nuclear Station, Waterford-3, and the Vogtle Electric Generating Plant as well as Indian Point Unit 3 during the past interval. The methodology used in the Indian Point Unit 3 RISB application for assessing TASCS potential conforms to these updated criteria. Additionally, materials reliability program (MRP) MRP-1 46 guidance on the subject of TASCS was also incorporated into the Indian Point Unit 3 RIS_B application.

3.3 Element and NDE Selection Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RIS_B applications provided criteria for identifying the number and location of required examinations. Ten percent of the HSS welds shall be selected for examination as follows:

(1) Examinations shall be prorated equally among systems to the extent practical, and each system shall individually meet the following requirements (for Indian Point Unit 3, because there are limited IFIV welds present in the RH and SI systems due to the fact that most branch lines are classified as RC out to the first isolation valve, the overall IFIV 2/3 requirement must be satisfied by selecting RC system welds in lieu of normal system-specific selections.):

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

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

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

(3) For the RCPB, at least two-thirds of the examinations shall be located between the inside first isolation valve (IFIV) (i.e., isolation valve closest to the RPV) and the RPV (for Indian Point Unit 3, because there are limited IFIV welds present in the RH and SI systems due to the fact that most branch lines are classified as RC out to the first isolation valve, the overall IFIV 2/3 requirement must be satisfied by selecting RC system welds in lieu of normal system-specific selections.).

(4) A minimum of 10% of the welds in that portion of the RCPB that lies outside containment (not applicable for Indian Point Unit 3) shall be selected.

(5) A minimum of 10% of the welds within the break exclusion region (BER) shall be selected (not applicable to Indian Point Unit 3).

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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-1 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': Class 2 Weldsi All Piping Welds Unit Total Selected Total fSelected Total Selected 3 631 60 1111 30 1742 90 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, 292 are HSS; the remaining are LSS.

(3) Regardless of safety significance, Class 1, 2, and 3 ASME Section Xl 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.

(4) Class 3 Service water piping in the 480 Volt Switchgear Room was identified as HSS and is included in the RISB Program.

3.3.1 Current Examinations Indian Point Unit 3 is currently using the NRC previously approved application using EPRI-TR 112657B-A.

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 Xl, 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. Ifthe flaw is acceptable for continued service, successive examinations will be scheduled per Section 6 of Code Case N-716. Ifthe flaw is found unacceptable for continued operation, it will be repaired in accordance with IWA-4000, applicable ASME Section XI Code Cases, or NRC approved alternatives. The IWB-3600 analytical evaluation will be submitted to the NRC. Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section XI. Evaluation of indications attributed to PWSCC and successive examinations of PWSCC indications will be performed in accordance with MRP-1 39 or a subsequent NRC rule making.

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RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE INACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 3.3.3 Scope Expansion Ifthe 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 MRP-139 or subsequent NRC rule makings.

3.3.4 Program Relief Requests Consistent with previously approved RISB submittals, Indian Point Unit 3 will calculate coverage and use additional examinations or techniques in the same manner it has for traditional Section Xl examinations. Experience has shown this process to be weld-specific (e.g., joint configuration). As such, the effect on risk, if any, will not be known until the examinations are performed. Relief requests for those cases where greater than 90% coverage is not obtained will be submitted per the requirements of 10 CFR 50.55a(g)(5)(iv).

No Indian Point Unit 3 relief requests are being withdrawn due to the RIS_B 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.

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) 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 1E-08 per year per system, respectively.

For LSS welds, Conditional Core Damage Probability (CCDP)/Conditional Large Early Release Probability (CLERP) values of 1E-4/1 E-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 1E-4 (CCDP)/1 E-5 (CLERP) and between Medium and Low consequence categories are 1E-6 (CCDP)/1 E-7 (CLERP) from the EPRI RI-ISI Risk Matrix. Using these threshold values streamlines the change-in-risk evaluation as well as stabilizes the update process. For example, if a CCDP changes from 1E-5 to 3E-5 due to an update, it will remain below the 1E-4 threshold value; the change-in-risk evaluation would not require updating.

The updated internal flooding PRA was also reviewed to ensure that there is no LSS Class 2 piping with a CCDP/CLERP greater than 1E-4/1 E-5. This review identified some piping in the RHR and SI systems located outside of containment with a CCDP greater than 1E-4. As a result, all LSS welds in these systems are conservatively assigned CCDP/CLERP equal to 1.2E-2 /

1.2E-3.

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 RISB applications shows this to be conservative.

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RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RIS_B PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Indian Point Unit 3 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-1 12657 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).

CCDP and CLERP Values Based on Break Location Break Location Estimated Consequence Rank Upper Bound Designation CCDP CLERP CCDP CLERP LOCA 8E-03 8E-04 HIGH 8E-03 8E-04 RCPB pipe breaks that result in a loss of coolant accident - The highest CCDP for Large LOCA, IE-A, was used (0.1 margin used for CLERP). Unisolable RCPB piping of all sizes.

PLOCA(1 )(" I 3E-05 3E-06 MEDIUM IE-04 1E-05 Isolable or Potential LOCA (1 open valve or I closed valve) inside containment - RCPB pipe breaks that result in an isolable or potential LOCA - Calculated based on Large LOCA CCDP of 8E-3 and valve fail to close probability of -3E-3 (0.1 margin used for CLERP). Between 1st and 2nd isolation valve inside drywell.

PPLOCA(" I <1E-5 <1E-06 MEDIUM IE-04 1E-05 Potential LOCA (2 closed valves) inside containment - Based on failure of two normally closed valves in series from the ISLOCA analysis. Applies to RHR shutdown cooling suction and discharge paths. Although the CCDP is less than 1E-6, 1E-5 is used as a bounding value in consideration of RHR operation during shutdown.

FB <1E-05 <1E-06 MEDIUM 1E-04 1E-05 Feedwater breaks based on bounding value for IE-T4, T5U, T5D and IE-FLD-AF- 1 (0.1 margin used for CLERP)

Class 2 LSS IE-04 1E-05 MEDIUM IE-04 1E-05 Class 2 pipe breaks that occur in the remaining system piping designated as low safety significant except for AF and SI - Estimated based on upper bound for Medium Consequence.

Class 2 LSS SIS 1.2E-02 1.2E-03 MEDIUM 1.2E-02 1.2E-03 Class 2 pipe break with internal flooding CCDP > IE-4. The 1.2E-2 value is conservatively applied to all RHR and SI piping (only certain pipe sections apply on El 32 of PAB from IE-FLD-PB-3 1).

The PRA does not explicitly model potential and isolable LOCA events, because such events are subsumed by the LOCA initiators in the PRA. That is, the frequency of a LOCA in this limited piping downstream of the first RCPB isolation valve times the probability that the valve fails is a small contributor to the total LOCA frequency. The N-716 methodology must evaluate these segments individually; thus, it is necessary to estimate their contribution. This is estimated by taking the LOCA CCDP and multiplying it by the valve failure probability.

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE INACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

2. PLOCA is identified and used in the quantification of both ILOCA and PLOCA The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability. The basic likelihood of PBF for a piping location with no degradation mechanism present is given as Xo and is expected to have a value less than 1E-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-1 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 RIS_B approach.

Table 3.4 presents a summary of the RIS_B Program versus the ASME Section XI program requirements on a "per system" basis. 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 RIS_B 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 tables, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RIS_B Program, and that the acceptance criteria of Regulatory Guide 1.174 and Code Case N-716 are satisfied.

With POD Credit Without POD Credit System Delta CDF Delta LERF Delta CDF Delta LERF CH - Chemical Volume & Control -8.65E-09 -8.65E-10 -4.80E-09 -4.80E-10 FW - Feedwater O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 RC - Reactor Coolant -1.14E-08 - 1.14E-09 -3.68E-09 -3.68E-10 RHR - Residual Heat Removal 1.43E-08 1.43E-09 1.43E-08 1.43E-09 SI - Safety Injection 1.34E-08 1.34E-09 1.79E-08 1.79E-09 CS - Containment Spray O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 MS - Main Steam 1.80E-10 1.80E-11 1.80E- 10 1.80E- 11 Total 7.89E-09 7.89E-10 2.39E-08 2.39E-09 As shown in Table 3.4, new RISB locations were selected such that the RISB selections exceed the Section Xl selections for certain categories (Delta column has a positive number).

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

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 RISB selections were set equal to the Section XI selections (Delta changed from positive number to zero). The acceptance criteria are met when the number of RIS_B selections is not allowed to exceed Section Xl.

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 PressureRetaining 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 1E-06 (or 1E-07 for LERF) be included in the scope of the application. Indian Point Unit 3 identified Class 3 service water piping in the 480 Volt Switchgear Room as HSS.

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 fourth ISI interval. No changes to the Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.

The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements.

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RIS_B PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Existing ASME Section Xl 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 RISB 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.

Ifan 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).

For preservice examinations, Indian Point Unit 3 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.

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

4. PROPOSED ISI PLAN CHANGE Indian Point Unit 3 is currently in the first period of the fourth ISI interval and plans to implement this RISB submittal for the entire fourth interval. The traditional ASME Section XI weld selections and inspections are being implemented until approval. In anticipation of the approval of this RISB submittal, any traditional ASME Section XI selected welds that require examination during the 1 st Period prior to approval will also meet the examination requirements of Table 1 of Code Case N-716. After approval of the RISB submittal, those welds in the RISB scope that were examined during the 1 st period that also met Table 1 requirements may be credited toward the RISB requirements for the Period.

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 previousSection XI program requirements for in-scope piping is provided in Table 4. For Class 1 piping welds, this includes inspections conducted for the 2 nd interval (prior to the N578 application in the 3 rd interval) and for Class 2 piping welds, this included inspections conducted for the 3 rd interval. In addition, service water piping in the 480 Volt Switchgear Room was identified as high safety significant and is included in the RISB Program. Ten percent of the welds will be inspected during the interval. No degradation mechanism was identified for this piping, but a wall thickness type of volumetric exam will be conducted since this is considered most relevant to service water systems.

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RIS_B PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

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 Classificationand Examination Requirements, Section X1 Division 1.

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

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

Regulatory Guide 1.200, Rev 2 An Approach For Determining The TechnicalAdequacy Of ProbabilisticRisk Assessment Results ForRisk-Informed Activities.

USNRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for Alternative GG-ISI-002-1mplement 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: ProbabilisticRisk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs.

Supporting Onsite Documentation Structural Integrity Report 0800767.302, Rev 0 "N-716 Evaluation for Indian Point Unit 3" E1-19

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.1 Code Case N-716 Safety Significance Determination N-716 Safety Significance Determination Safety Significance System Weld PWR: CDF > 1E-Description Count RCPB SDC FW BER 6 High Low CH 69 FW 64 254 ____ ___

RC 4 ___ _ _ 1 20 RHR 228 171 234 V '(

SI 50 V 1" 391 1 CS 74 __

MS 183 V, 557 1 1_

Summary 74__"_' I' Results for 228 LI / "

all Systems 64 / ,/

819 _ _

TOTAL 1742 (1) System Scope:

CH = Chemical Volume and Control System FW = Main Feedwater RC = Reactor Coolant RHR = Residual Heat Removal SI = Safety Injection CS = Containment Spray MS = Main Steam (2) Service water piping in the 480 volt switchgear room is included in the HSS scope E1-20

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.2 Failure Potential Assessment Summary Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive System(') TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC CH " _

FW RC " " _"

RHR SI ,_ "

CS MS 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 CS and MS in its entirety, as well as portions of the RHR and SI systems.

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.3: Code Case N716 Selections System Weld Count N716 Selection Considerations Selections HSS LSS DMs RCPB RCPB (LFIV) RCPB (OC) BER CH 7 T / 3 7 TT V 1 3 TASCS V V 3 42 None V V 0 10 None V/ 0 FW 64 None 7 RC 13 Tr V V 1 9 TT,TASCS V V 7 3 TASCS V V 2 4 PWSCC V V 4 192 None V/ 12 37 None V_ 0 RHR 7 None V V 2 13 None V 0 228 None 23 171 Assumed None 0 SI 4 TT,IGSCC V_ 1 14 IGSCC V_ 4 20 TT V V 5 70 None V V 15 176 None V 4 391 Assumed None 0 CS 74 Assumed None 0 MS 183 Assumed None 0 40 TT V V 9 7 TT V 1 9 TT,TASCS V V 7 6 TASCS V V 5 Summary 4 PWSCC V V 4 Results All 4 TT,IGSCC I1 Systems 14 IGSCC V 4 311 None V V 29 236 None V 4 292 None 30 819 None 0 Totals 923 j819 94 Note: Systems are described in Table 3.1 El-22

RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 3.4: Risk Impact Analysis Results System Safety Break Location Failure Potential Inspections CDF Impact LERF Impact System Significance DMs Rank SXI RISB Delta w/POD w/o POD w/POD w/o POD CH High LOCA TT Medium 0 3 3 -4.32E-09 -2.40E-09 -4.32E-10 -2.40E-10 CH High LOCA TASCS Medium 0 3 3 -4.32E-09 -2.40E-09 -4:32E-10 -2.40E-10 CH High PLOCA TT Medium 0 1 1 -1.80E-11 -1.00E-11 -1.80E-12 -1.00E-12 CH High LOCA None Low 0 0 0 0.00E+00 0.00E+00 0.00E+00 0.00E+00 CH High PLOCA None Low 0 0 0 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 CH Total -8.66E-09 -4.81E-09 -8.66E-10 -4.81E-10 FW High FB None Low 7 7 0 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 FW Total 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 RC High LOCA TT Medium 2 1 -1 -4.80E-10 8.00E-10 -4.80E-11 8.OOE-11 RC High LOCA TT,TASCS Medium 2 7 5 -9.12E-09 -4.OOE-09 -9.12E-10 -4.00E-10 RC High LOCA TASCS Medium 0 2 2 -2.88E-09 -1.60E-09 -2.88E-10 -1.60E-10 RC High LOCA PWSCC Medium 4 4 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 RC High LOCA None Low 40 12 -28 1.12E-09 1.12E-09 1.12E-10 1.12E-10 RC High PLOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 RC Total -1.14E-08 -3.68E-09 -1.14E-09 -3.68E-10 RH High LOCA None Low 0 2 2 -8.OOE-11 -8.OOE-11 -8.OOE-12 -8.OOE-12 RH High PLOCA None Low 3 0 -3 1.50E-12 1.50E-12 1.50E-13 1.50E-13 RH High PPLOCA None Low 28 23 -5 2.50E-12 2.50E-12 2.50E-13 2.50E-13 RH Low Class 2 LSS SIS Assume Medium 12 0 -12 1.44E-08 1.44E-08 1.44E-09 1.44E-09 RH Total 1.43E-08 1.43E-08 1.43E-09 1.43E-09 SI High LOCA TI" Medium 4 5 1 -5.28E-09 -8.OOE-10 -5.28E-10 -8.OOE-11 SI High PLOCA TT, IGSCC Medium 0 1 1 -1.00E-11 -1.00E 1.00E-12 -1.00E-12 SI High PLOCA IGSCC Medium 0 4 4 -4.OOE-11 -4.OOE-11 -4.OOE-12 -4.OOE-12 SI High LOCA None Low 2 15 13 -5.20E-10 -5.20E-10 -5.20E-11 -5.20E-11 SI High PLOCA None Low 13 0 -13 6.50E-12 6.50E-12 6.50E-13 6.50E-13 SI Low Class 2 LSS SIS Assume Medium 16 0 -16 1.92E-08 1.92E-08 1.92E-09 1.92E-09 SI Total 1.34E-08 1.78E-08 1.34E-09 1.78E-09 CS Total Low Class 2 LSS Assume Medium 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 MS Total Low Class 2 LSS Assume Medium 18 0 -18 1.80E-10 1.80E-10 1.80E-11 1.80E-11 Grand Total 1 151 90 -61 7.84E-09 2.39E-08 7.84E-10 2.39E-09 Notes

1. Systems are described in Table 3.1 E1-23

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2. Only those ASME Section Xl 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.

3. 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.

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

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

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RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 INDIAN POINT UNIT 3 RISB PROGRAM PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table 4: Inspection Location Selections Comparison System Safety Significance Break Location Failure Potential Code Weld Section XI Code Case N716 Syste High Low DMs Rank Category Count Vol Surface RISB Other CH V" LOCA TT Medium B-J 7 0 1 3 NA CH Vs LOCA TASCS Medium B-J 3 0 0 3 NA CH V" PLOCA 1T1 Medium B-J 7 0 0 1 NA CH V" LOCA None Low B-J 42 0 12 0 NA CH V" PLOCA None Low B-J 10 0 6 0 NA FW Vs FB None Low C-F-2 64 7 0 7 NA RC Vs LOCA TT Medium B-F, B-J 13 2 1 1 NA RC V" LOCA TT,TASCS Medium B-J 9 2 0 7 NA RC Vs LOCA TASCS Medium B-J 3 0 1 2 NA RC Vs LOCA PWSCC Medium B-F 4 4 0 4 NA RC Vs LOCA None Low B-F, B-J 192 40 35 12 NA RC Vs PLOCA None Low B-J 37 0 5 0 NA RH Vs LOCA None Low B-J 7 0 0 2 NA RH Vs PLOCA None Low B-J 13 3 0 0 NA RH Vs PPLOCA None 'Low C-F-1 228 28 0 23 NA RH Vs Class 2 LSS SIS Assume Medium C-F-1 171 12 1 0 NA SI Vs LOCA TT Medium B-J 20 4 0 5 NA SI Vs PLOCA TT, IGSCC Medium B-J 4 0 0 1 NA SI Vs PLOCA IGSCC Medium B-J 14 0 2 4 NA SI Vs LOCA None Low B-J 70 2 6 15 NA SI Vs PLOCA None Low B-J 176 13 59 0 4 SI Vs Class 2 LSS IF Assume Medium C-F-i 391 16 19 0 NA CS Vs Class 2 LSS IF Assume Medium C-F-1 74 0 0 0 NA MS Vs Class 2 LSS Assume Medium C-F-2 183 18 3 0 NA Total 1742 151 151 90 4 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 Indian Point Unit 3 RISB 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).

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").

E1-25

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RIS_B PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE INACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

The Indian Point 3 (IP3) Probabilistic Risk Assessment (PRA) model used for this application [Reference 1] is the most recent evaluation of the IP3 risk profile for internal event challenges. The IP3 PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause failure events.

The PRA model quantification process used for the IP3 PRA is based on the event tree and fault tree methodology, which is a well-known methodology in the industry.

Entergy employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating Entergy nuclear power plants. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the IP3 PRA model.

PRA Maintenance and Update The Entergy risk management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plant. This process is defined in the Entergy fleet procedure EN-DC-151, "PSA Maintenance and Update" [Reference 2]. This procedure delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating Entergy nuclear power plants. In addition, the procedure also defines the process for implementing regularly scheduled and interim PRA model updates, and for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.). To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following activities are routinely performed:

" Design changes and procedure changes are reviewed for their impact on the PRA model. Potential PRA model changes resulting from these reviews are entered into the Model Change Request (MCR) database, and a determination is made regarding the significance of the change with respect to current PRA model.

New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated approximately every four years, and

" Industry standards, experience, and technologies are periodically reviewed to ensure that any changes are appropriately incorporated into the models.

In addition, following each periodic PRA model update, Entergy performs a self assessment to assure that the PRA quality and expectations for all current applications are met. The Entergy PRA maintenance and update procedure requires updating of all risk informed applications that may have been impacted by the update.

Regulatory Guide 1.200 PWROG PeerReview of the IP3 Internal Events PRA Model The IP3 PRA internal events model went through a Regulatory Guide 1.200 PWR Owners Group peer review in December 2010. The NEI 05-04 process [Reference 3], the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA Standard A-1

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

[Reference 4], and Regulatory Guide 1.200, Rev. 2 [Reference 5]) were used for the peer review.

The 2010 1P3 PRA Peer Review was a full-scope review of all the Technical Elements of the internal events, at-power PRA:

" Initiating Events Analysis (IE)

" Accident Sequence Analysis (AS)

Success Criteria (SC)

" Systems Analysis (SY)

Human Reliability Analysis (HR)

Data Analysis (DA)

Internal Flooding (IF)

Quantification (QU)

LERF Analysis (LE)

Maintenance and Update Process (MU)

During the IP3 PRA model Peer Review, the technical elements identified above were assessed with respect to Capability Category II criteria to better focus the Supporting Requirement assessments. The ASME/ANS PRA Standard has 326 individual Supporting Requirements. Eleven (11) of the ASME/ANS PRA Standard Supporting Requirements are not applicable to IP3 (e.g., BWR related, multi-site related). Of the 315 ASME/ANS PRA Standard Supporting Requirements applicable to the IP3 PRA model, approximately 97%

were satisfied at Capability Category II criteria or greater. The Facts and Observations (F&Os) for the IP3 PRA peer review are provided in the report, entitled, "RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for the Indian Point 3 Probabilistic Risk Assessment' [Reference 6]. Of the 68 Facts and Observations (F&Os) generated by the Peer Review Team, 11 were considered Findings, 52 were Suggestions, and five were Best Practices.

As a result of the Regulatory Guide 1.200 PWROG peer review, all the abovementioned F&Os (other than best practices) have been identified as potential improvements to the IP3 PRA model and are tracked in the Entergy Model Change Request (MCR) database. Table A-1 contains the findings resulting from the peer review, including the status of the resolution for each finding and the potential impact of each finding on this application. In summary, a majority of the findings were related to documentation and have no material impact. Resolution of the peer review findings is expected to have a minor impact on the model and its quantitative results and will have a negligible, if any, impact on the conclusions of this application.

In resolving the peer review findings, several additional internal flooding sources were identified as not being addressed in the original internal flooding analysis report. Most of those sources involved fire protection piping, but they also included auxiliary component cooling water (ACCW) piping in the fan house and short sections of component cooling water (CCW) piping in a pipe chase in the foyer outside the charging pump rooms. These additional sources are described in more detail in Table A-2, including their expected impact on this application.

It should be noted that, while the model documentation has been revised to resolve most of the documentation related findings, since the revised documents will be formally issued with the final update package, those findings are considered resolved but will not be considered A-2

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RIS_B PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE INACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) closed until the final revised model and report, addressing the peer review findings, are formally released, which is expected later this year.

External Events External Events are addressed in Parts 4 through 9 of the ASME/ANS standard. The EPRI Topical Report (TR) [Reference 7] proposes a qualitative treatment of the risk from fire events and from events that impose extreme loads on piping systems. The NRC Safety Evaluation concurred in the TR conclusion that challenges from fire events are expected to be less frequent and not significantly different than challenges caused by the random occurrence of internal initiating events. The NRC SE also concluded that additional analysis of extreme loading events are not needed and will not change the conclusion derived from the RI-ISI program.

Summary The IP3 PRA technical capability evaluations and the maintenance and update processes described above provide a robust basis for concluding that the IP3 PRA model is suitable for use in the risk-informed process used for this application.

References

[1] Engineering Report, IP3-RPT-10-00023, Rev.0, "Indian Point Unit 3 Probabilistic Safety Assessment (PSA)", November 2010.

[2] Entergy Fleet Procedure EN-DC-151, Revision 2, "PSA Maintenance and Update",

January 2011.

[3] NEI 05-04, Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard, Nuclear Energy Institute, Rev. 2, November 2008.

[4] American Society of Mechanical Engineers/American Nuclear Society, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, (ASME RA-Sa-2009), February 2009.

[5] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, Revision 2, March 2009.

[6] PWR Owners Group LTR-RAM-1-1 1-055, "RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for the Indian Point 3 Probabilistic Risk Assessment," October 2011.

[7] EPRI Technical Report 1021467, "Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs", July 2011.

A-3

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RIS_B PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-1 Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc Review Team Finding Finding Description SR Basis for Significance Suggested Status / N716 Impact Resolution 1-11 Appendix C1 of IP-RPT-10-00023, Rev. 0 IFSN-B1 Analysis details available to the Provide required Resolved - No impact provides a high to medium level summary peer review team such as documentation of the flood scenarios, and provides flooding calculations, were not This is a documentation issue that would impact future model greater depth in some areas. Analysis sufficient to support upgrades updates and upgrades. The backup spreadsheets used for details available to the peer review team and would have to be obtained flooding rates and frequency calculations have been obtained as such as flooding calculations, were not or reproduced for future model well as the software used for flood level calculations, sufficient to support upgrades and would changes. The documentation instructions for use of this software and material that supports its have to be obtained or reproduced for also lacks in reference to application. This additional documentation will be included in future model changes. The quantification input the final model documentation package. Initiator specific flag documentation also lacks in reference to documentation (initiator specific files exist but were not included in either the internal flooding or quantification input documentation flag files) quantification notebooks but are contained in the electronic files (initiator specific flag files) to be included in the model update documentation package.

These flag files will be added to the internal flooding notebook

,(This F&O originated from SR IFSN-B1) as well.

A-4

RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RIS_B PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-1 Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc Review Team Finding Finding Description SsR Basis for Significance Suggested Status / N716 Impact Resolution 1-12 The walkdown notes in Appendix A of IP- IFSN-A5 There is no specific physical For SSCs Resolved - No Impact RPT-10-00023, Rev. 0, Appendix C.1 location information found in susceptible to note the general location of each SSC the documentation for SSCs spray failure (also Additional discussion has been added to the walkdown with respect to its room and elevation as other than flood area and see F&O 2-3), Appendix to support the spray impacts included in the model.

well as its submergence height. Some elevation. Therefore, it cannot ensure sufficient This includes reference to environmental qualification additional general locational information is be determined which SSCs in relational location documents where these were used as a basis for stating that sometimes identified in Section 4.2 of IP- any area are susceptible to information equipment would not be vulnerable to spray damage. A RPT-10-00023, Rev. 0, Appendix C.1. spray from any specific spray between the target conservative separation criterion of 30 feet was used in For example, it may state that a flood source. In the scenario SSC and spray examining the potential for spray impacts in the analysis. The source may impact one but not both trains development it identifies which sources are composite piping and general arrangement drawings were of equipment; specifics are not given as to equipment is impacted by provided so that a scrutinized to ascertain whether equipment could be sprayed why both cannot be impacted (e.g., spray, but it cannot be determination can should a line or other piece of equipment rupture. The text of shielding, curbs, etc.), but the information determined how that be made as to the report has been changed to note this. Providing additional implies the impact of spatial information, information was obtained or if it whether the SSCs specific location information within the model documentation will is correct. can be damaged be considered to support future updates but is considered a There is no specific physical location by each potential documentation enhancement issue with no expected impact on information related to spray type failures spray source. the analysis.

found in the documentation. SSCs are only identified locationally by their flood area and elevation. It cannot be determined which SSCs in any area are susceptible to spray from any specific spray source.

,(This F&O originated from SR IFSN-A5)

A-5

RELIEF REQUEST iP3-ISi-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF, PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

I. Table A-i Summary of Industry Peer Review Findings for the 11P3 Internal Events PRA Model Update Assoc Review Team Finding Description Basis for Significance Suggested Status / N716 Impact SR Resolution The effects of the flood on PSFs were not IFQU-A6 Limited flooding-related human Discuss flood Open - Minor Impact (Increase) No short term isolation specifically addressed in the HRA actions are included in the HRA effects on PSFs actions were credited in the flooding analysis. The only analysis. discussion in Appendix H, but and make significant field actions credited in the internal events model that there is no mention of any adjustments to the could be impacted by the plant conditions associated with affects of the flood on PSFs. HRA analysis if flooding are alignment of alternate cooling to the charging (This F&O originated from SR IFQU-A6) needed. pumps on loss of CCW and operator actions associated with locally operating the turbine-driven AFW (TDAFW) pump.

Major flood scenarios in these areas would fail the components involved, thus rendering any impact on operator actions moot.

Lesser flood or spray scenarios could affect operator actions.

With respect to the need to locally align the TDAFW Pump, this action is only required coincident with a station blackout, or a substantial number of other failures. Since the combined frequency of the flooding events that could impact this action is approximately 1 E-5/yr, it is reasonable to conclude that such scenarios would be well below the criteria for low safety significance and would have no significant impact on the application.

With respect to the operator action to align city water to the charging pumps, since flooding initiator IE-FLD-PB-8 has the same impact as assuming failure of this operator action (i.e.

both CCW and backup city water to the charging pumps are lost) the effect of any HRA impacts on the flood scenarios in this area were bounded by assuming the operator actions were precluded by the flood event and comparing the impact to this existing flooding initiator. The frequency of a failure of the CCW piping in the charging pump foyer that would require operator alignment of city water and could impact that action is approximately 2.3E-6 per year. Existing flood initiator IE-FLD-PB8 has a flood frequency of 2.76E-6 and CDF and LERF contributions of 5.85 E-7/yr and 1 E-9/yr, respectively. Since both of these impacts are below the criteria for low safety significance, the impact of assuming that the operator action is precluded for breaks in the charging pump foyer would be similar.

A-6

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-1 Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc Review Team Finding Finding Description SR Basis for Significance Suggested Status / N716 Impact Resolution 6-6 Supporting requirement IFSO-A4 is IFSO-A4 This supporting requirement is Identify the Resolved - No impact intentionally not met as stated in IP-RPT- intentionally not met as stated flooding 10-00023, Rev. 0, Appendix Cl, Section in IP-RPT-10-00023, Rev. 0, mechanisms that The intent of the statement in the report was to acknowledge 3.3: The one supporting requirement of Appendix C1, Section 3.3: The would result in a that the EPRI data used for the analysis included all rupture the ASME standard that we have made one supporting requirement of release for each mechanisms that contribute to piping system failures and to note no attempt to meet is IF-B2: "for each the ASME standard that we potential source of there are no readily available data that would allow us to potential source of flooding, identify the have made no attempt to meet flooding t6 meet distinguish between different release mechanisms. The mechanisms that would result in a is IF-B2: "for each potential the SR. identification of specific causes of failure is therefore a flooding release". In this analysis, no source of flooding, identify the documentation issue. The only contributor not included in the distinction was made between the various mechanisms that would result EPRI data is human induced flooding events. Since no causes of floods because the rupture in a flooding release". In this applicable generic data exists related to human induced events, frequencies used included all floods." analysis, no distinction was plant specific condition reports were reviewed for applicable made between the various events (none were identified) and discussions were held with (This F&O originated from SR IFSO-A4) causes of floods because the plant operations personnel. Based on those discussions, rupture frequencies used activities that could challenge system integrity such as large included all floods." scale movements of water and plant modifications are typically performed during outages and would not constitute significant contributors to flooding risk. Nonetheless, the model documentation has been modified to specifically discuss both failure mechanisms and the conclusions of these human induced failure evaluations.

6-7 As stated in IP-RPT-10-00023, Rev. 0, IFSO-A5 As stated in IP-RPT-10-00023, Identify the Resolved - No Impact Appendix Cl, Table 3.3.1.1 for IFSO-A5, Rev. 0, Appendix C1, Table characteristic of maximum flow rate resulting from a 3.3.1.1 for IFSO-A5, maximum release for each No impact. We consider this a documentation issue. While the guillotine rupture is determined and used, flow rate resulting from a source and its table mentioned in the finding did state that a maximum flow instead of identifying the characteristic of guillotine rupture is determined identified failure rate resulting from a guillotine rupture was determined, it also release for different failure mechanism. and used, instead of identifying mechanism. noted that the frequency of this and lesser releases were the characteristic of release for calculated. A range of release sizes consistent with the (This F&O originated from SR IFSO-A5) different failure mechanism. available EPRI pipe rupture frequency data were, in fact, This is in contrary to the SR. considered and a flow rate and frequency of occurrence derived for each. By this means, the size and frequency of possible releases were matched as required for the quantitative determination of the consequences of internal flooding. The text in the report has been modified to clarify this matter. Additional information regarding the pressures and temperatures of the ruptured systems has also been added to the documentation.

A-7

RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-1 Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc .Review Team Finding Finding Description SR Basis for Significance Suggested Status I N716 Impact Resolution 6-8 IP-RPT-1 0-00023, Rev. 0, Appendix Cl, IFSO-A1 IP-RPT-10-00023, Rev. 0, Identify the Resolved - No Impact Section 4.1.3 states that the potential Appendix C1, Section 4.1.3 potential sources flood sources were identified by states that the potential flood of flooding for All accessible fire areas were included in the plant walkdowns.

walkdowns and the examination of sources were identified by each flood area Appendix A has been revised to include the areas that were drawings, and listed in Appendix A, Plant walkdowns and the per the standard. omitted from the documentation, including those areas Walkdown. However, Appendix A does examination of drawings, and mentioned in the finding.

not provide adequate information on flood listed in Appendix A, Plant Perform and source as (1) some flood areas are not Walkdown. However, Appendix document The statement in the introduction to the walkdown notes was included in the walkdown such as A does not provide adequate walkdowns for intended only to acknowledge that there might be small bore, 3PAB41-1A,43-60A, 46-73A,55-63A, information on flood source as missed flood field run piping (less than 1 inch diameter) that were not shown 3FH72-B, 3FH80-A, etc.; (2) Appendix A (1) some flood areas are not areas. If these on system drawings and would not have been confirmed by the has stressed that the walkdown notes do included in the walkdown such areas cannot be walkdown. Such small bore pipes were not considered to be NOT provide a definitive listing of all as 3PAB41 -1A,43-60A, 46- walked down for significant flood sources.

equipment and lines or other flood 73A,55-63A, 3FH72-B, 3FH80- operational or sources. Also other fluid sources have not A, etc.; (2) Appendix A has health reasons, been considered in the analysis. stressed that the walkdown other methods of notes do NOT provide a obtaining this data (This F&O originated from SR IFSO-Al) definitive listing of all (e.g., plant equipment and lines or other drawings, operator flood sources. Also other fluid interviews, etc.)

sources have not been should be considered in the analysis. employed and documented.

Prepare an integrated list of the internal flood source.

A-8

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-1 Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc Review Team Finding Finding Description SR Basis for Significance Suggested Status / N716 Impact Resolution 6-11 IP-RPT-10-00023, Rev. 0, Appendix C, IFSO-B1 There is no list of the internal Prepare an Resolved - No impact Section 4.1.3, which is the section in the flood sources in the analysis integrated list of main report for flood sources, just refers that may facilitate PRA the internal flood This is documentation issue. A list of internal flooding sources Appendix A, Plant Walkdown for the applications, upgrades, and source. has been developed and will be included in a new Table 4.2.1.1 information. There is no list of the internal peer review, in the final update report. This table identifies all the flooding flood sources in the analysis that may sources in each area and identifies adjacent or lower areas facilitate PRA applications, upgrades, and It could facilitate applications, through which floodwater might propagate.

peer review. update and review if sources were identified in the main (This F&O originated from SR IFSO-B1) report.

6-12 IP-RPT-10-00023, Rev. 0, Appendix C IFSO-B2 IP-RPT-10-00023, Rev. 0, Provide adequate Resolved - No impact identifies applicable flood sources in its Appendix C identifies documentation on Appendix A, Plant Walkdown, which is not applicable flood sources in its the process used Although Section 3.1.2 previously described the process for adequate for process documentation Appendix A, Plant Walkdown, to identify identifying flooding sources, additional description has been purpose. For example, the walkdown which is not adequate for applicable flood added to that section and an additional table (Table 4.2.1.1) has notes stressed that they do NOT provide process documentation sources been added, which provides additional detail describing the a definitive listing of all equipment and purpose. For example, the sources in each flood zone. In any case, this is an issue of lines or other flood sources; there is no list walkdown notes stressed that enhanced documentation and does not impact this application.

of sources to be examined, they do NOT provide a definitive listing of all The statement in the introduction to the walkdown notes was (This F&O originated from SR IFSO-B2) equipment and lines or other intended only to acknowledge that there might be small bore, flood sources; there is no list of field run piping (less than 1 inch diameter) that were not shown sources to be examined, on system drawings and would not have been confirmed by the walkdown. Such small bore pipes were not considered to be significant flood sources.

A-9

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-1 Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc Review Team Finding Finding Description SR Basis for Significance Suggested Status N716 Impact Resolution 6-1 The justification/statement that the CST SC-B1 The justification/ statement that Perform rigorous Resolved - No impact inventory is sufficient for AFW for 24 hrs the CST inventory is sufficient evaluation/justifica should be enhanced, for AFW for 24 hrs should be tion of the CST Plant design documentation supports the 24 mission time for the enhanced. IP-RPT-10-00023, inventory to CST. In addition, as noted, CST inventory is typically (This F&O originated from SR SC-B1) Rev. 0, Appendix B, Section support 24-hour maintained above the minimum inventory level, providing B13.3.1.3.2 states early that AFW operation. additional margin. Final model documentation will be modified CST inventory is sufficient for to remove the apparent discrepancies.

24 hrs while later reveals that the MAAP analysis shows insufficient CST inventory with statement that alignment to the city water supply may be required. An informal calculation with the minimum flow requirement in EOP concludes that "itwould seem that there is enough inventory in the CST to allow the AFW system to operate for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />". Then in IP-RPT 0023, Section Insights states that 'As the normal CST inventory is sufficient to supply the AFW pumps for the 24-hour mission time inthe PSA', no credit is taken for the alternate suction path from city water supply.

A-10

RELIEF REQUEST R P3-ISi-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB3 PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-i Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc Review Team Finding Finding Description SR Basis for Significance Suggested Status / N716 Impact Resolution 4-14 Failure modes and success criteria DA-A2 Based on the documents As described in Open - No significant impact defined in Systems Analysis are reviewed and the issues Sections 5.10 and consistent with the Data Analysis. This SR identified, component 6.3.11 of Appendix This is a documentation issue. The model documentation will also asks for establishing consistent SSC boundaries are not consistent DO, assure be revised to provide sufficient detail to show that all system and boundaries between the system level among failure rate, CCF and component component boundaries in the current model meet or are analysis and the data analysis. unavailability data. Plant- boundaries conservative when compared to the way the generic databases specific features need to be defined in failure define the boundaries. Resolution will only impact future model Reviewed Appendix E6 and E27 of the considered for boundary rate and CCF data updates and upgrades if the apparent discrepancies in the systems notebooks and Appendix D for definitions. match the PSA generic boundary definitions are resolved or change. Any future the Data Analysis. Below is a list of issues model. Assure the impact is not likely to be significant.

identified: It is possible to ensure that the boundaries used inconsistent boundary in the test and Note that Battery Charger input and output breakers are

1. System notebooks do not define the included in the generic database boundary definition for definitions result in maintenance data component boundaries. The component common cause failures whereas the input breakers are not conservative results, but is consistent with boundaries are defined by the generic clearly identified to be included in the generic independent realistic rather than the PSA model.

failure rate data source with limited failure rate. The PSA model does not include common cause conservative results is ideal. Make adjustments discussions on plant-specific SSC failure of the input or output breakers but does conservatively CCF events tend to dominate or provide features and modeling considerations. include independent failure of the input breakers due to specific system level cutsets and justification for any conservative CCF basic event mismatch modeling considerations.

2. The guidance document Appendix DO values may mask other identified.

Section 5.10 states 'Assure the important components in a component boundaries established in system. Review plant-the generic data match those defined in specific CCF the PSA model. Make adjustments or experience for justify differences'. Also, Attachment 4, consistency to Section 3.0 of the same document states that CCF boundaries are dictated meet SY DA-D6 by the fault tree modeling. However, the requirements.

component boundaries defined for failure rate and CCF data do not match.

The justification for using the data that way is that it is the conservative to do so. It is true that this approach is conservative for Emergency Diesel Generators, but it may not be conservative for other cases like batteries and battery chargers where CCF of output breakers are not modeled.

A-11

RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-1 Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc Review Team Finding Finding Description SR Basis for Significance Suggested Status I N716 Impact Resolution 4-14 (continued)

3. Sections 1.2 and 1.4 of Appendix D1 state that the data analysis package is consistent with the system analysis.

However, as discussed in Item number 1 above, systems analysis only defines the system boundary and not the component boundaries within the system.

4. Boundaries of the test and maintenance unavailability events are not specifically discussed, but seem to be same as the boundaries for the failure rates. Data from the Maintenance Rule program is used for this case, but it is not clear if the system and component boundaries considered in this program is consistent with the PSA model boundaries. Section 6.3.11 of Appendix DO discusses this issue, but there is no evidence that the analysis done in Appendix D1 considered boundaries applies to routine test and maintenance practices at IP3.

1(This F&O originated from SR DA-A2)

A-12

RELIEF REQUEST IP3-1SI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A-1 Summary of Industry Peer Review Findings for the IP3 Internal Events PRA Model Update Assoc Review Team Finding Finding Description SR Basis for Significance Suggested Status / N716 Impact Resolution 1-15 The initiating event frequencies are not IE-C5 The initiating event frequencies Include the plant Open - No significant impact weighted by the fraction of time the plant are not weighted by the fraction availability factor is at power. of time the plant is at power. in the calculation While we agree that the wording in the SR itself indicates that of initiating event weighting should be done, the ASME standard acknowledges Section 10.9 of Appendix AO provides frequencies. that the SR wording is somewhat unclear by providing a guidance to account for plant availability lengthy and detailed note of explanation (i.e. Note 1 of the in initiating event calculations. Section 4.0 SR). Entergy believes that the annual average model, which of Appendix Al states that the availability Note (1) acknowledges should not include the weighting factor for the data update period was factors, is the appropriate baseline model in the absence of an calculated. However, the calculated value all modes model. We do agree, as the standard states, that an is not incorporated into the initiating event all modes model should account for the time in each operating or final CDF results. state. Since we do not have an all modes model at this time and we believe that tying risk values to plant availability without (This F&O originated from SR IE-C5) an all modes model can potentially provide inappropriate risk insights to non-PSA personnel, in that it does not apply any risk to other operating states, we believe that at the least, our current model meets the SR, when taken in concert with the associated Note 1.

In any case, the current approach provides, at most, a slightly conservative result in comparison to use of the stipulated weighting approach and would have no significant impact on this application.

A-13

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A Additional Flood Sources and Impact Location Source N716 Impact Primary Auxiliary Building - 55 ft Component Cooling Water (CCW) As noted in the N716 impact discussion for Finding 3-7 in Table A-1 above, failures of Elevation lines 148 and 149 in a pipe chase these lines can potentially impact the ability to align alternate cooling to the charging adjacent to the charging pump pumps in addition to the loss of CCW, similar to the impact of existing flooding initiator IE-room foyer FLD-PB-8. The frequency of such a failure of the CCW piping is approximately 2.3E-6 per year. Existing flood initiator IE-FLD-PB-8 has a flood frequency of 2.76E-6 and CDF and LERF contributions of 5.85 E-7/yr and 1E-9/yr, respectively. This is bounding for this additional impact since some potential would still exist for the operator to successfully align alternate cooling to the charging pumps. Therefore, since the impact of assuming that the operator action is completely precluded for breaks in the charging pump foyer would still be below the 1E-6/yr and 1E-7/yr criteria for low safety significance, this additional source is not expected to impact this application.

Primary Auxiliary Building - 55 ft Primary Water line 393 in a pipe The loss of primary water has no significant impact on the risk model and, in any case, Elevation chase adjacent to the charging has a failure frequency of 4.1 E-7/yr, which is already below the CDF threshold for low pump room foyer safety significance.

Primary Auxiliary Building - 46 ft Fire Protection line traversing the No damage or plant transient is predicted from this scenario due to the large duct and Elevation upper electrical penetration area drains in the floor of the electrical tunnel.

Primary Auxiliary Building - 34 ft City water system in the lower The rupture of the city water system in the lower electrical tunnel will not result in any Elevation electrical tunnel, flood zone spray damage since the city water line is surrounded by a guard tube.

3PAB34-7A Primary Auxiliary Building - 43 ft Fire protection system in both the The fire protection system in both the upper and lower electrical tunnels are dry-pipe pre-Elevation and 34 ft Elevation upper and lower electrical tunnels action systems.

Primary Auxiliary Building - 34 ft Fire protection line in the boron No damage or plant transient is predicted from this scenario due to the large duct and Elevation injection tank room. drains in the floor of the electrical tunnel.

Fan House - 67 ft Elevation Component Cooling Water (CCW) The impact of the failure of these lines may result in a loss of CCW event but has no lines on the fan house mezzanine other consequential impacts. Since the frequency of such a failure is less than 2E-6/yr associated with auxiliary and the conditional core damage probability (CCDP) following a Loss of CCW initiating component cooling water pumps event is 1.6E-3, the contribution of such a failure would be several orders of magnitude 31, 32, 33 and 34 below the 1E-6/yr threshold for low safety significance.

Fuel Storage Building - 55 ft Liquid waste lines enter and This flood scenario will neither require a plant shutdown nor damage safety related Elevation traverse the fuel storage building. equipment.

A-14

RELIEF REQUEST IP3-ISI-RR-05 ENCLOSURE 1 ATTACHMENT A INDIAN POINT UNIT 3 RISB PROGRAM CONSIDERATION OF THE ADEQUACY OF PROBABILISTIC RISK ASSESSMENT MODEL PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

Table A Additional Flood Sources and Impact Location Source N716 Impact Pipe Bridge - 41 ft Elevation Fire Protection lines traverse the The only flooding scenario of consequence would be a catastrophic failure of this fire pipe bridge between the Turbine protection line. Although there are multiple egress paths, the bounding impact of such a Building and the Feedwater failure would involve all the flood water entering the turbine building through the openings Feedvater Buid18 ft ElBuilding, the main boiler feedwater in the west wall of the pipe bridge. Such a failure would have to continue at maximum area on the 18 ft Elevation, and the flow for well over an hour before building up on the turbine floor sufficiently to challenge main steam and feedwater valve the normal offsite power busses. Since the flood frequency for this event is area on the 43 ft Elevation. approximately 1.4E-5/yr and the onsite EDGs would remain unaffected by this event, it can be concluded that the CDF contribution would be well below the 1E-6 threshold for low safety significance and would not significantly impact this application.

Feedwater Building - 18 ft Fire protection lines in the AFW An evaluation of the fire protection system in the IP3 AFW Pump room, done as part of Elevation Pump room the fire suppression analysis performed for IP3 in response to GI-57, concluded that the AFW pump motors would not be impacted by spray from an inadvertent actuation or rupture of the fire protection piping in that room. Although the AFW pump motors were not specifically qualified for the chemical spray associated with a DBA, the qualification testing did impose HELB conditions, including a period of immersion. It is therefore expected that the AFW pumps will operate successfully should they be subjected to spray following a fire protection system failure.

A catastrophic failure of the fire protection system will not result in submergence of the AFW pumps since the AFP room has a door flap designed to relieve such an inflow.

Such a failure of fire protection in this area would not lead to a plant transient and operators would not be expected to manually trip the plant or perform a controlled shutdown prior to assessing any impacts.

It is therefore concluded that this additional scenario would not impact this application.

A-15

v t e Letter Sequence Request
TAC:ME7940, (Open)
Initiation Request Acceptance Administration Questions 19 July 2012 Answers 16 August 2012 Results Approval... Other:
MONTHYEAR2012-02-27 [Table View]

NL-12-028, Relief Request IP3-ISI-RR-05 for Fourth Ten-Year Inservice Inspection Interval

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NL-12-028, Relief Request IP3-ISI-RR-05 for Fourth Ten-Year Inservice Inspection Interval

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