ML16057A534

NYS Exhibit 1 Re Entergy December 9, 2014 Letter on Proposed License Amendment Re Extending Containment Type a Leak Rate Testing Frequency to 15 Years

ML16057A534
Person / Time
Site:	Indian Point
Issue date:	12/09/2014
From:	State of NY, Office of the Attorney General
To:	NRC/OCM
SECY RAS
Shared Package
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References
RAS 50964, ASLBP 15-942-06-LA-BD01, 50-247-LA
Download: ML16057A534 (140)

Text

-EntergaEnteray Nuclear NortheastIndian Point Energy Center450 Broadway, GSBP.O. Box 249Buchanan, NY 10511-0249Tel 914 254 6700Lawrence CoyleSite Vice PresidentNL-14-128December 9, 2014U.S. Nuclear Regulatory CommissionATTN: Document Control Desk11545 Rockville Pike, TWFN-2 F1Rockville, MD 20852-2738

SUBJECT:

Proposed License Amendment Regarding Extending the Containment Type A LeakRate Testing Frequency to 15 yearsIndian Point Unit Number 2Docket No. 50-247License No. DPR-26

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby requests a LicenseAmendment to Operating License DPR-26, Docket No. 50-247 for Indian Point Nuclear GeneratingUnit No. 2 (IP2). The proposed TS change contained herein would revise Appendix A, TechnicalSpecifications (TS), to allow extension of the ten-year frequency of the Type A or Integrated LeakRate Test (ILRT) that is required by Technical Specification (TS) 5.5.14 to 15 years on apermanent basis.Entergy has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1) using thecriteria of 10 CFR 50.92(c) and Entergy has determined that this proposed change involves nosignificant hazards, as described in Attachment 1. The marked up page showing the proposedchange is provided in Attachment 2. An assessment of the risk impact of extending the ILRTinterval is provided in Attachment 3. A copy of this application and the associated attachments arebeing submitted to the designated New York State official in accordance with 10 CFR 50.91.Entergy requests approval of the proposed amendment in one calendar year and an allowance of30 days for implementation. There are no new commitments being made in this submittal. If youhave any questions or require additional information, please contact Mr. Robert Walpole, Manager,Regulatory Assurance at (914) 254-6710.AD/7 NL-14-128Docket 50-247Page 2 of 2I declare under penalty of perjury that the foregoing is true and correct. Executed on December,2014.Sincerely,LC/spAttachments: 1. Analysis of Proposed Technical Specification Changes Regarding 15Year Containment ILRT2. Marked Up Technical Specifications Page for Proposed ChangesRegarding 15 Year Containment ILRT3. Risk Impact of Extending the ILRT interval Associated with the ProposedTechnical Specification Changescc: Mr. Douglas Pickett, Senior Project Manager, NRC NRR DORLMr. Daniel H. Dorman, Regional Administrator, NRC Region 1NRC Resident InspectorMr. John B. Rhodes, President and CEO, NYSERDAMs. Bridget Frymire, New York State Dept. of Public Service ATTACHMENT 1 TO NL-14-128ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGESREGARDING 15 YEAR CONTAINMENT ILRTENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NO. 2DOCKET NO. 50-247 NL-14-128Docket No. 50-247Attachment 1Page 1 of 191.0 DESCRIPTIONEntergy Nuclear Operations, Inc. (Entergy) is requesting an amendment to Operating LicenseDPR-26, Docket No. 50-247 for Indian Point Nuclear Generating Unit No. 2 (IP2). The proposedTechnical Specification (TS) change contained herein would revise Appendix A, TS, to allowextension of the ten-year frequency of the Type A or Integrated Leak Rate Test (ILRT) that isrequired by TS 5.5.15 to 15 years on a permanent basis.The specific proposed changes are listed in the following section.2.0 PROPOSED CHANGESThe containment leakage rate testing program in Technical Specification 5.5.15 currently says"A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordance withthe guidelines contained in Regulatory Guide 1.163, "Performance-BasedContainment Leak-Test Program," dated September, 1995."The proposed TS 5.5.15 is as follows:"A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordancewith NEI 94-01, Revision 2A, "Industry Guideline for ImplementingPerformance-Based Option of 10 CFR Part 50, Appendix J," October 2008."3.0 BACKGROUNDThe testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from thecontainment, including systems and components that penetrate the containment, do not exceedthe allowable leakage values specified in the TS. Furthermore, the requirements ensure thatperiodic surveillance of the containment, containment penetrations and isolation valves isperformed so that proper maintenance and repairs are made during the service life of thecontainment, the systems and penetrations. The limitation on containment leakage providesassurance that the containment would perform its design function following an accident up to andincluding the plant design basis accident. Appendix J identifies three types of required tests: (1)Type A tests, intended to measure the containment overall integrated leakage rate; (2) Type Btests, intended to detect local leaks and to measure leakage across pressure-containing orleakage limiting boundaries (other than valves) for containment penetrations; and (3) Type C tests,intended to measure containment isolation valve leakage. Type B and C tests identify the vastmajority of potential containment leakage paths. Type A tests identify the overall integratedcontainment leakage rate and serve to ensure continued leakage integrity of the containmentstructure by evaluating those structural parts of the containment not covered by Type B and Ctesting.

NL-14-128Docket No. 50-247Attachment 1Page 2 of 19In 1995, 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," was amended to provide a performance-based Option B for thecontainment leakage testing requirements. Option B requires that test intervals for Type A, TypeB, and Type C testing be determined by using a performance-based approach. Performance-based test intervals are based on consideration of the operating history of the component andresulting risk from its failure. The use of the term "performance-based' in 10 CFR 50, Appendix Jrefers to both the performance history necessary to extend test intervals as well as to the criterianecessary to meet the requirements of Option B.Regulatory Guide (RG) 1.163 was also issued in 1995. The RG endorsed NEI 94-01, Revision 0,"Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," withcertain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successfulType A tests) to reduce the test frequency from the containment Type A (ILRT) test from threetests in ten years to one test in ten years. This relaxation was based on an NRC risk assessmentcontained in NUREG-1493, "Performance-Based Containment Leak-Test Program," and ElectricPower Research Institute (EPRI) TR-1 04285, "Risk Impact Assessment of Revised ContainmentLeak Rate Testing Intervals." These documents illustrated that the risk increase associated withextending the ILRT surveillance interval was very small.By letter dated August 7, 1996, Indian Point Unit 2 submitted a TS change request, supplementedby letter dated March 12, 1997, to implement 10 CFR 50, Appendix J, Option B. The NRCapproved this request as Amendment 190 issued in NRC letter of April 10, 1997. The NRC notedthe proposed TS changes were in compliance with the requirements of Option B, and areconsistent with the guidance in RG 1.163. With the approval of the amendment, IP2 transitioned toa performance-based ten year frequency for the Type A tests.Entergy submitted an Amendment request to extend the ILRT interval one time from ten years to15 years in a letter dated July 13, 2001 that was supplemented by letters dated November 30,2001 March 13, April 3, May 30, and June 13, 2002. This one-time extension was approved bythe NRC, as license Amendment 232 on August 5, 2002.By letter dated August 31, 2007, NEI submitted NEI 94-01, Revision 2, and EPRI report No.1009325, Revision 2, "Risk Impact Assessment of Extended Integrated Leak Rate TestingIntervals," to the NRC Staff for review. NEI 94-01, Revision 2, describes an approach forimplementing the optional performance-based requirements of Option B, which includes provisionsfor extending Type A intervals to up to 15 years and incorporates the regulatory positions stated inRG 1.163. It delineates a performance-based approach for determining Type A, Type B, and TypeC containment leakage rate surveillance testing frequencies. This method uses industryperformance data, plant-specific performance data, and risk insights in determining the appropriatetesting frequency. NEI 94-01, Revision 2, also discusses the performance factors that licenseesmust consider in determining test intervals.The NEI guideline does not address how to perform the tests because these details are included inreferenced industry documents (e.g., American National Standards institute/American NuclearSociety (ANSI/ANS) 56.8-2002).The NRC final Safety Evaluation (SE) issued by letter dated June 25, 2008, documents theevaluation and acceptance of NEI 94-01, Revision 2, subject to the specific limitations andconditions listed in Section 4.1 of the SE. The accepted version of NEI 94-01 Rev 2A was NL-14-128Docket No. 50-247Attachment 1Page 3 of 19issued as Revision 2A dated October 2008.EPRI Report No. 1009325, Revision 2, provides a risk impact assessment for optimized ILRTintervals of up to 15 years, using current industry performance data and risk-informed guidance,primarily Revision 1 of RG 1.174, "An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." The NRC's final SE issuedby letter dated June 25, 2008, documents the evaluation and acceptance of EPRI Report No.1009325, Revision 2, subject to the specific limitations and conditions listed in Section 4.2 of theSE. An accepted version of EPRI Report No. 1009325 has subsequently been issued asRevision 2A (also identified as Technical Report TR-1 018243) dated October 2008.The proposed amendment would revise TS 5.5.14, "Containment Leakage Rate Testing Program,"by replacing the reference to Regulatory Guide (RG) 1.163, "Performance-Based ContainmentLeak Test Program," with a reference to Nuclear Energy institute (NEI) topical report NEI 94-01,"Industry Guideline for implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"Revision 2A, dated October 2008, as the implementation document used by Entergy to developthe Indian Point 2 performance-based leakage testing program in accordance with Option B of 10CFR 50, Appendix J (Option B).Revision 2A of NEI 94-01 describes an approach for implementing the optional performance-basedrequirements of Option B, including provisions for extending primary containment integrated leakrate test (ILRT) intervals to 15 years, and incorporates the regulatory positions stated in RG 1.163.In the SE issued by NRC letter dated June 25, 2008, the NRC concluded that NEI 94-01, Revision2, describes an acceptable approach for implementing the optional performance-basedrequirements of Option B, and found that NEI 94-01, Revision 2, is acceptable for referencing bylicensees proposing to amend their TS in regard to containment leakage rate testing, subject to thelimitations and conditions noted in Section 4.0 of the SE. IPEC is not applying for the extendedType C performance based testing beyond 60 months but will be adopting the testing criteriaANSI/ANS 56.8 -2002 rather than the criteria of ANSI/ANS 56.8 -1994.The proposed extension of the interval for the primary containment ILRT, which is currentlyrequired to be performed at ten year intervals, to 15 years from the last ILRT would revise the nextscheduled ILRT to March 2021 as opposed to the ILRT currently scheduled for March 2016. Thisis approximately 15 years since the last ILRT which was completed in April 2006.The currently proposed change would allow successive ILRTs to be performed at 15-year intervals(assuming acceptable performance history). The performance of fewer ILRTs would result insignificant savings in radiation exposure to personnel, cost, and critical path time during futurerefueling outages.4.0 Technical EvaluationAs required by 10 CFR 50.54(o), the IP2 containment is subject to the requirements set forth in 10CFR 50, Appendix J. Option B of Appendix J which requires that test intervals for Type A, Type B,and Type C testing be determined by using a performance-based approach. Currently, the 10CFR 50 Appendix J Testing Program Plan is based on RG 1.163, which endorses NEI 94-01,Revision 0. This LAR proposes to revise the 10 CFR 50, Appendix J Testing Program Plan byimplementing the guidance in NEI 94-01, Revision 2A but will not extend the Type B and Cleakage beyond 60 months. Testing will be performed in accordance with ANSI/ANS 56.8 -2002.

NL-14-128Docket No. 50-247Attachment 1Page 4 of 194.1 Limitations and ConditionsIn the June 25, 2008 NRC SE, the NRC concluded that NEI 94-01, Revision 2, describes anacceptable approach for implementing the optional performance-based requirements of Option B,and found that NEI 94-01, Revision 2, is acceptable for referencing by licensees proposing toamend their TS in regard to containment leakage rate testing, subject to the limitations andconditions noted in Section 4.0 of the SE.The following Table 4.1 -1 lists the SE Section 4.1 Limitations and Conditions as well ascompliance with each of the six limitations and conditions.Table 4.1-1Limitations and Conditions (Section IP2 Compliance4.1 of Safety Evaluation Dated June,25,2008)For calculating the Type A leakage rate, Implementation of NEI 94-01 Rev 2A willthe licensee should use the definition in require use of the definition of "performancethe NEI TR 94-01, Revision 2, in lieu of leakage rate" defined in Section 5.0 forthat in ANSI/ANS-56.8-2002. (Refer to SE calculating the Type A leakage rate whenSection 3.1.1.1). performing Type A tests.The licensee submits a schedule of NEI-94-01 Rev 2A, Section 9.2.3.2 requires acontainment inspections to be performed general visual examination prior to each Typeprior to and between Type A tests. (Refer A test and at least 3 other outages before theto SE Section 3.1.1.3). ILRT. This should be scheduled inconjunction with or coordinated withexaminations required by ASME Code,Section Xl, Subsections IWE and IWL. Aschedule of containment inspections isprovided in Section 4.4The licensee addresses the areas of the A general visual examination of accessiblecontainment structure potentially interior and exterior surfaces is conducted persubjected to degradation. (Refer to SE the Containment Inservice Inspection PlanSection 3.1.3). which implements the requirements of ASME,Section Xl, Subsections IWE and IWL. IP2will explore / consider inaccessibledegradation-susceptible areas that can beinspected using viable, commercially availableNDE methods.The licensee addresses any tests and The design change process will address anyinspections performed following major testing and inspection requirements followingmodifications to the containment future major modifications to the containmentstructure, as applicable. (Refer to SE structure. This process provides a disciplinedSection 3.1.4). approach for determining the program andsystem interfaces associated with designchange. This process evaluates requirementspertaining to the ASME Containment In-Service Inspection Program, ASME AppendixJ (Primary Containment Leak Rate Testing)

NL-14-128Docket No. 50-247Attachment 1Page 5 of 19Table 4.1-1Limitations and Conditions (Section IP2 Compliance4.1 of Safety Evaluation Dated June,25,2008)Program, and ASME Section Xl.The normal Type A test interval should be IP2 is adopting, consistent with Section 9.2.2less than 15 years. If a licensee has to of NEI 94-01 Rev 2A, a Type A test intervalutilize the provision of Section 9.1 of NEI defined as the time period from the completionTR 94-01, Revision 2, related to extending of a Type A test to the start of the next test.the ILRT interval beyond 15 years, the This definition will be used for scheduling andlicensee must demonstrate to the NRC planning of the next Type A test to the monthstaff that it is an unforeseen emergent and year (see RIS 2008-27).condition. (Refer to SE Section 3.1.1.2).For plants licensed under 10 CFR Part 52, Not applicable to IP2.applications requesting a permanentextension of the ILRT surveillance intervalto 15 years should be deferred until afterthe construction and testing ofcontainments for that design have beencompleted and applicants have confirmedthe applicability of NEI TR 94-01, Revision2, and EPRI Report No. 1009325,Revision 2, including the use of pastcontainment ILRT data.4.2 Existing ExceptionsThe provisions of RG 1.163 have been incorporated into NEI 94-01 Revision 2A so if there hadbeen an exception to RG 1.63 it would remain unchanged.4.3 Previous Test results4.3.1 ILRT Test ResultsPast IP2 ILRT results have confirmed that the containment is acceptable with respect to the designcriterion of 0.1% leakage of containment air weight at the design basis loss of coolant accidentpressure (La). Since the last two Type A "as found" tests for IP2 had "as found" test results of lessthan 1.01La, a test frequency of 15 years in accordance with NEI 94-01 Revision 2A would beacceptable. The last two tests were:1. The last ILRT in April 2006 had a measured containment leak rate (Ltm) at the testpressure of 60.5 psia was 0.0636 % containment air weight / day with a 95% confidencelevel.2. The prior ILRT in June 1991 had a measured containment leak rate (Ltm) at the testpressure of 61.7 psia was 0.0478 % containment air weight / day with a 95% confidencelevel.For background, the prior three Type A tests had the following results:

NL-14-128Docket No. 50-247Attachment 1Page 6 of 19Date As found Leakage (% Test Pressure (psia)Containment weight perday)December, 1987 0.0342 62.9September, 1984 0.0320 65.6August, 1979 0.0260 62.74.3.2 Type B and C testingThe IP2 Appendix J, Type B and Type C testing program requires testing of the componentsrequired by 10 CFR 50, Appendix J, Option B. Technical Specification Amendment 174, datedJune 17, 1997, approved the adoption of 10 CFR 50, Appendix J, Option B performance basedtesting requirements for containment leakage testing. The minimum pathway combined Type Band Type C leakage from the March 2006 outage, when the last Type A test was performed, isprovided below. The subsequent combined as found Type B and Type C test values during eachsuccessive outage since the last Type A test are also provided below. The data is provided inpercentage of leakage allowed (0.6La).Table 4.3-1Date As-Found La (ccm) Percent ((As- Percent ((As-Leakage Found/La) xl00) Found/.6La))xl 00)(sccm)April 46,105.04 215490 0.214 0.3572006April 54,659.95 215490 0.254 0.4232008April 28,880.44 215490 0.134 0.2232010April 47,304.18 215490 0.220 0.3662012March 79,176.85 215490 0.367 0.6122014 _I IBased on the results the largest as found leakage and the as left conditions are within theacceptance criterion associated with the 15 year ILRT.Table 4.3-2 provides a listing of the containment penetrations subject to Type B and C testing, thetest frequency, the last test date and the next test date, and the as left leakage. Notes are providedfor test failures.

NL-14-128Docket No. 50-247Attachment 1Page 7 of 19Table 4.3-2Penetration Description Type Test Frequency Last Test date Next test date "as -Left"(months) Leakage(cc/min)Penetration UU B 30 3/2/14 3/16 0.00Penetration W B 30 3/2/14 3/16 0.00Fuel Transfer Tube B 30 3/16/14 3/16 38.25Equipment Hatch Seal B 30 3/15/14 3/16 93.0080ALOK Personnel Airlock -80 foot B 30 6/12/14 12/16 6818.6095ALOK Personnel Airlock -95 foot B 30 6/6/14 12/16 14481.00WCCPP Zone 2 -Racks 10, 11 B 36 4/12/13 4/12/15 12744.00WCCPP Zone 2 -Racks 12,13 B 36 4/12/13 4/12/15 2265.60Y Pressurizer relief tank N2 supply tank C 30 2/28/14 3/16 1387.50RCS -Valve RC-518Y Pressurizer relief tank N2 supply tank C 60 3/10/14 3/18 3.50RCS -Valve RC-3418, 3419 and 4136GG Containment spray headers -Valve C 60 3/12/14 3/18 1570.25867A,878AP Containment spray headers -Valve SI- C 60 3/6/14 3/18 0867BRR Accumulator N2 supply -Valve 863- C 60 3/16/12 3/16 199.00RR Accumulator N2 supply -Valve 4312 C 60 3/16/12 3/16 6.00V Primary system vent and N2 supply -C 60 3/14/14, 3/18 31.00Valve WD-3416, 3417, 5459V Primary system vent and N2 supply- C 30 3/14/14 3/16 21000Valve WD-1616RR Containment Air Sample In (Rad) -C 60 3/5/13 3/18 32.50Valves PCV-1234, PCV-1235RR Containment Air Sample Pot (Rad) -C 60 3/5/13 3/18 2.80Valves PCV-1236, PCV-1237R Air Ejector Discharge to Containment -C 30 3/3/14 3/16 271.50Valve CA-1229R Air Ejector Discharge to Containment -C 30 3/3/14 3/16 135.75 NL-14-128Docket No. 50-247Attachment 1Page 8 of 19Table 4.3-2Penetration Description Type Test Frequency Last Test date Next test date "as -Left"(months) Leakage(cc/min)Valve CA-1230EE Vent Purge Supply Duct -Valve VS- C 30 3/12/14 3/16 6380.001170 and VS-1171FF Vent Purge Exhaust Duct -Valve VS- C 30 3/12/14 3/16 9482.501172 and VS-1173PP Cont Pressure Relief Vent -Valves VS- C 30 3/12/14 3/16 300.001190, VS-1191PP Cont Pressure Relief Vent -Valve VS- C 30 3/12/14 3/16 294.001192TT Post Accident Sample system supply C 60 3/12/14 3/18 0.00lines -Valve SP-5018 and SP-5019LL Post Accident Sample system supply C 60 3/12/14 3/18 3.00lines -Valve SP-5020 and SP-5021R Post Accident Sample system return C 60 2/28/14 3/18 0.00lines -Valve SP-5022 and SP-50230 Post Accident Sample system return C 60 2/28/14 3/18 0.00lines -Valve SP-5024 and SP-5025Y Instrument air (post accident vent C 60 3/3/14 3/18 8.50supply) -Valve IA-39Y Instrument air (post accident vent C 30 3/29/11 3/16 24.25supply) -Valve IA-1228LL Post Accident Vent Exhaust Valves E-2 C 60 2/26/14 3/18 0.00and E-1, E-3, E-5Personnel air lock -Outer Door Valve C 60 2/28/13 3/18 57.0085APersonnel air lock -Outer Door Valve C 60 2/28/13 3/18 250.1095APersonnel air lock -Inner Door Valve C 60 2/28/13 3/18 59.5085BPersonnel air lock -Inner Door Valve C 60 2/28/13 3/18 0.3595B I IIII_ I NL-14-128Docket No. 50-247Attachment 1Page 9 of 19Table 4.3-2Penetration Description Type Test Frequency Last Test date Next test date "as -Left"(months) Leakage(cc/min)Personnel air lock- Inner Door Valve C 60 2/28/13 3/18 37.5085CPersonnel air lock -Inner Door Valve C 60 2/28/13 3/18 0.0095CPersonnel air lock- Inner Door Valve C 60 2/28/13 3/18 47.2585DPersonnel air lock -Inner Door Valve C 60 2/28/13 3/18 1.8095DPneumatic Indicator Lines (SG level-2, C 60 3/14/14 3/18 5.00pressurizer level-i, pressurizerpressure-I) -Valve IIP-500 and IIP-501Pneumatic Indicator Lines (SG level-2, C 30 3/14/14 3/18 590.00pressurizer level-i, pressurizerpressure-I) -Valve IIP-502 and IIP-503Pneumatic Indicator Lines (SG level-2, C 60 3/14/14 3/18 7.00pressurizer level-i, pressurizerpressure-I) -Valve IIP-504 and IIP-505Pneumatic Indicator Lines (SG level-2, C 60 3/14/14 3/18 16.00pressurizer level-i, pressurizerpressure-1) -Valve IIP-506 and IIP-507 NL-14-128Docket No. 50-247Attachment 1Page 10 of 194.4 Code InspectionsPrior to each Type A test a general visual examination is required of accessible interior andexterior surfaces of the containment for structural issues that may affect the performance of theType A test. This inspection will be performed as part of the Containment Inservice Inspection (ISI)Plan to implement the requirements of ASME, Section Xl, Subsection IWE and IWL (the applicablecode edition and addenda for the fourth 10 year interval is ASME Section Xl, 2001 Editionincluding the 2002 and 2003 Addenda in paragraph (b)(2)).The examination performed in accordance with the ISI program to meet Subsections IWE and IWLsatisfies the general visual examination requirements specified in Option B. The identification andevaluation of inaccessible areas are addressed in accordance with the requirements of 10 CFR50.55a(b)(2)(ix). Each ten year ISI interval is divided into three approximately equal inspectionperiods. A minimum of one inspection required by the IWE inspection program is performed duringeach inspection period of the ISI period to meet the program requirements. IWL visualexaminations of accessible concrete containment surfaces are to be completed once every 5 yearswithin the limitations specified in IWL-2410(b), (c), and (d) resulting in at least two IWLexaminations being performed during a 15 year type A and typically scheduled in two of the threeinspection periods of a 10 year ISI interval. Therefore, the frequency of the examinationsperformed in accordance with the IWE / IWL program will satisfy the requirements of NEI 94-01Revision 2A, Section 9.2.3.2, to perform a general visual examination before the Type A test duringat least three other outages before the next Type A test if the interval is extended to 15 years. Thelast ILRT was performed April 2006 and the next 15 year interval will end 12 months after 2R24scheduled for the spring of 2020. The following Tables illustrates the current and plannedinspection intervals for the IP2 first and second IWE inspection intervals:Table 4.4-1IWE InspectionsInspection Inspection Period Start Period End Refuel RefuelInterval Period Date Date Outage Month/YearSeptember September 2R13 Spring 19971 1 9,1996 9, 2001 2R14 Spring 2001September Jan 9, 2005 Spring 20021 2 9, 2001 2R151 3 Jan 10, 2005 Feb 28,2007* 2R16 Spring 20042R17 Spring 20062R18 Spring 20082 1 March 1, 2007 May 31, 2010 2R19 Spring 20102 2 June 1, 2010 May 31,2013 2R20 Spring 20122 3 June 1,2013 May 31, 2016 2R21 Spring 20142R22 Spring 2016* Based upon this extended First Period that ended on September 9, 2001, the First 10-YrInterval for IP2 Containment ISI was originally scheduled to end on May 9, 2010, but wasshortened to align with the Third ISI Interval.

NL-14-128Docket No. 50-247Attachment 1Page 11 of 19The IWL inspections are performed per the following schedule:Table 4.4.2IWL InspectionsInspection Interval Inspection Period IWL Inspection Dates1 1 June 20001 2 June 20051 3 June 20102 1 June 20152 2 June 2020For IP2 the First Interval CII Program Plan was originally effective from September 9, 1996,through and including May 9, 2010. This time period has been shortened to end on February 28,2007. IWE Containment inservice examinations scheduled for the first 40-month period werecompleted during the Third Period of the Third ISI Inspection Interval. These examinations nowserve the same purpose as pre-service baseline examinations. The required IWL inserviceexaminations were also completed and re-inspections are scheduled at 5 year frequency.The Second Ten-Year Interval for IWE Containment ISI inspections at IP2 will commence onMarch 1, 2007 coincident with the start of the Fourth 10-Year ISI Program Interval. Therefore, boththe ISI and the CII IWE & IWL Program Plans will be aligned with the Fourth Interval ISI Programschedule and ASME Code requirements.The following information provides the IP2 IWE examination results of the containment metal linercompleted during refuel outages 2R18 (2008), 2R20 (2012) and 2R21 (2014) and the IWLexamination results for the containment concrete visual inspections completed in 2005 and 2010(these are not always completed in an outage). The next IWE examination is scheduled for 2R23(2018) prior to the proposed date for the next ILRT. The next IWL examination is scheduled for2016 and the inspection will also be scheduled prior to the proposed date for the next ILRT 2R24(2020). Corrective Actions identified by these inspections are provided with the discussions. Thereare no primary containment surface areas that require augmented examination in accordance withASME Section XI, IWE-1240.4.4.1 IWE ExaminationsIP2 IWE containment inspection for the current fourth ISI interval was performed on 2008 -2R18outage, 2012 -2R20 outage and 2014- 2R21 outage.Refueling Outage 2R18 (2008) Containment Inservice InspectionExaminations were performed for the Containment Surfaces (Containment Vessel AccessibleSurface Areas, Bolted connections, Wall and Dome Liner and Moisture Barriers) during 2R18 in2008. There were some deficiencies noted such as general surface corrosion, minor coatingpeeling/flaking, blistered paint, loose stainless steel insulation panels and buckling stainless steelinsulation panels (VC liner inaccessible) at columns 10 and 11 elevation 68'.The general surface corrosion, minor coating peeling/flaking and blistered paint were previouslyidentified and evaluated. These conditions were a repeat of previous inspections and were minorwith no change and therefore acceptable.

NL-14-128Docket No. 50-247Attachment 1Page 12 of 19The condition of the buckling locations and looseness on the VC liner plate insulation wasdocumented in the Corrective Action Program as Condition Report CR-IP2-2008-01892. CivilEngineering performed an inspection of the stainless steel insulation jacket and has determinedthat all but 2 of the insulation jacket issues are acceptable. The two areas not acceptable wererepaired during 2R1 8 outage.Refueling Outage 2R20 (2012) Containment Inservice InspectionExaminations were performed for the Containment Surfaces (Containment Vessel AccessibleSurface Areas, Bolted connections, Wall and Dome Liner and Moisture Barriers) during 2R20 in2012. Most of the findings such as surface corrosion and minor coating flaking and peeling were arepeat of previous inspections and were minor with no change and therefore acceptable. Therewere also some deficiencies noted on the Electrical penetration #69 of the Containment Buildingpenetrations; there was observed water seeping adjacent to penetration #69. This condition wasdocumented in IP2 Corrective Action Program under Condition Report CR-1P2-2012-01760. CivilDesign engineering walked down the penetration and the water seepage is from areas wherecrack/delimitation repairs where performed back in 2000. The water seepage observed has noadverse effect on the penetration as it is not emanating from the penetration sleeve. The sealaround the penetration is intact and the inside of the penetration itself is dry. This penetration wasalso looked at from the inside of the VC during the Maintenance Rule Inspection and no anomalieswere observed.All of the conditions noted during this inspection did not result in any structural degradation thatadversely affects the ability of the containment to perform its design function of maintainingintegrity during accident conditions.Refueling Outage 2R21 (2014) Containment Inservice InspectionExaminations were performed for the Containment Surfaces (Containment Vessel AccessibleSurface Areas, Bolted connections, Wall and Dome Liner and Moisture Barriers) during 2R21 in2014. Most of the findings were a repeat of previous inspections and were minor with no changeand therefore acceptable. All NDE examination reports were accepted during the 2014containment inspection therefore no condition reports were generated.4.4.2 IWL ExaminationsThe inspections are general visual inspections performed in accordance with the requirements ofthe ASME Boiler and Pressure Vessel Code, 1998 Edition,Section XI. Division 1, SubsectionIWL as required and modified by NRC, Code of Federal Regulation, Title 10, Part 50,Section 55a, "Codes and Standards,"(10 CFR 50.55a -1999). When needed, opticalenhancement equipment with zoom capabilities are used as visual aids during the inspections.All of the inspections are performed under the direction of the IWL Responsible Engineer(RE). The RE is the Civil/Structural Design Engineering Supervisor at IPEC and a NewYork State Registered Professional Engineer in accordance with the IWL Procedure. TheResponsible Engineer has knowledge of the Design and Construction Codes as well as othercriterion used in IP2's Containment. Degreed engineers perform the inspections under thedirection of the RE and are knowledgeable and trained in the design, evaluation andperformance requirements of structures and qualified to perform visual examination eitherdirectly or remotely, with adequate illumination, to detect evidence of degradation.

NL-14-128Docket No. 50-247Attachment 1Page 13 of 19The second period of the first interval of the IP2 IWL containment inspection was performed in thespring of 2005 and documented in IP-RPT-06-00019. Visual examinations were performed of allaccessible areas of the containment building exterior concrete including areas visible from insideother surrounding buildings. The concrete exhibited signs of normal weathering that are to beexpected for the time period that it has been in service. These indications include minor crackingto due pressurization, and minor areas of spalling with exposed rebar and cadwelds. The spallingat the cadwelds appears to be due to lack of concrete cover as a result of the cadwelds havingtwice the diameter as the rebar. There were also some locations of efflorescence which weredetermined to be unchanged since the previous inspection and thus deemed inactive. Severalareas of rust bleeding were identified but easily attributed to the lightning arrestors and the ductwork and have no impact on the structural capacity of the containment building. All together therewere 91 recordable indications identified during the inspection however all of them have beenevaluated and are not structural concerns. None of the indications reduce the structural capacityor ability of the containment structure to perform its safety function. Based on condition ofinspected areas it was not deemed necessary to inspect non-accessible areas. No conditionreports or work orders were required as a result of the inspection.The third period of the first interval of the IP2 IWL containment inspection was performed in thespring of 2010 and documented in IP-RPT-10-00027. Visual examinations were performed of allaccessible areas of the containment building exterior concrete including areas visible from insideother surrounding buildings. The concrete exhibited signs of normal weathering that are to beexpected for the time period that it has been in service. These indications include minor crackingto due pressurization, and minor areas of spalling with exposed rebar and cadwelds. The spallingat the cadwelds appears to be due to lack of concrete cover as a result of the cadwelds havingtwice the diameter as the rebar. There were also some locations of efflorescence which weredetermined to be unchanged since the previous inspection and thus deemed inactive. Severalareas of rust bleeding were identified but easily attributed to the lightning arrestors and the ductwork and have no impact on the structural capacity of the containment building. All together therewere 125 recordable indications identified during the inspection which increased from the 91identified in the previous inspection. This is partially attributed to the ILRT performed in 2006which caused several of the previous identified areas of potential future spalling to indeed spall. Inthe fall of 2009 several of the previously identified areas were cleaned and a coating was appliedto protect the exposed steel from future corrosion. All of the recordable indications identifiedduring the inspection have been evaluated and are not structural concerns. None of theindications reduce the structural capacity or ability of the containment structure to perform its safetyfunction. Based on condition of inspected areas it was not deemed necessary to inspect non-accessible areas. No condition reports or work orders were required as a result of the inspection.4.5 Confirmatory Analysis4.5.1 MethodologyAn evaluation has been performed to assess the risk impact of extending the IP2 ILRT interval fromthe current ten years to 15 years. This plant-specific risk assessment followed the guidance in NEI94-01, Revision 2A, the methodology outlined in EPRI TR-1 04285, August 1994 and TR-1 009325,Revision 2A, and the NRC regulatory guidance outlined in RG 1.174 on the use of Probabilistic RiskAssessment (PRA) findings and risk insights in support of a request to change the licensing basis ofthe plant. In addition, the methodology used for Calvert Cliffs Nuclear Power Plant to estimate thelikelihood and risk implication of corrosion-induced leakage of steel containment liners goingundetected during the extended ILRT interval was also used for sensitivity analysis.

NL-14-128Docket No. 50-247Attachment 1Page 14 of 19In their June 25, 2008, SE, the NRC concluded that a 15 year extension to the Type A ILRT intervalwas acceptable and that the methodology in EPRI TR-1009325, Revision 2, is acceptable forreferencing in a proposal to amend TS to extend the ILRT surveillance interval to 15 years. Thisapproval was subject to the limitations and conditions noted in Section 4.0 of the SE. The followingTable 4.5-1 lists the SE Section 4.2 Limitations and Conditions and a description of how the IP2analysis complies with those four limitations and conditionsTable 4.5 -1Limitations and Conditions of Risk IP2 ComplianceAssessmentThe licensee submits documentation The technical adequacy of the IP2 PRA andindicating that the technical adequacy of their consistency with the RG 1.200 requirementsPRA is consistent with the requirements of relevant to the ILRT extension are discussed inRG 1.200 relevant to the ILRT extension Section 4.5.2 and detailed in Appendix A ofapplication. Attachment 3.The licensee submits documentation The IP2 risk evaluation is summarized inindicating that the estimated risk increase Section 4.5.3 and described in detail inassociated with permanently extending the Attachment 3. The results of thatILRT surveillance interval to 15 years is small, evaluation demonstrate that the estimatedand consistent with the clarification provided risk increase is small and consistent within Section 3.2.4.5 of this SE. Specifically, a the criteria discussed in the SE.small increase in population dose should bedefined as an increase in population dose ofless than or equal to either 1.0 person-remper year or 1 percent of the total populationdose, whichever is less restrictive. In addition,a small increase in CCFP should be definedas a value marginally greater than thataccepted in previous one-time 15-year ILRTextension requests. This would require thatthe increase in CCFP be less than or equal to1.5 percentage point. While acceptable forthis application, the NRC staff is notendorsing these threshold values for otherapplications. Consistent with this limitationand condition, EPRI Report No. 1009325 willbe revised in the "-A" version of the report, tochange the population dose acceptanceguidelines and the CCFP guidelines.The methodology in EPRI Report No. The IP2 analysis used a pre-existing containment1009325, Revision 2, is acceptable except for leak rate of 1 0OLa to calculate the increase inthe calculation of the increase in expected population dose for the large leak rate accidentpopulation dose (per year of reactor case (EPRI Class 3b) .(Attachment 3, Sectionoperation). In order to make the methodology 1.3).acceptable, the average leak rate for the pre-existing containment large leak rate accidentcase (accident case 3b) used by thelicensees shall be 100 La instead of 35 La.

NL-14-128Docket No. 50-247Attachment 1Page 15 of 19Table 4.5 -1Limitations and Conditions of Risk IP2 ComplianceAssessmentA LAR is required in instances where Containment overpressure is not relied upon forcontainment over-pressure is relied upon for ECCS performance (Attachment 3, Section 5.8).ECCS performance.4.5.2 PRA QualityThe risk assessment performed for the IP2 ILRT extension request is based on the current Level 1and Level 2 PRA model of record, which was released in November 2011. Information developedfor the license renewal effort to support the Level 2 release categories is also used in this analysissupplemented by additional calculations to more appropriately represent the intact containmentcase in the ILRT extension risk assessment. A discussion of the Entergy model update process,the peer review performed on the IP2 model, the results of that peer review and the potentialimpact of peer review findings on the ILRT extension risk assessment are provided in Attachment3, Section A.2.It should be noted that, while the analysis presented in Attachment 3 was performed for both IP2and IP3, this submittal only addresses a LAR for IP2. The IP2 information presented in Attachment3 is therefore informational only and not part of the basis for the current LAR.4.5.3 Summary of Plant-Specific Risk Assessment ResultsThe findings of the IP2 risk assessment confirm the general findings of previous studies that therisk impact associated with extending the ILRT interval to one in 15 years is small. The IP2 plant-specific results for extending the ILRT interval to 15 years, taken from Attachment 3, Section 7.0,Conclusions, are summarized below.1. Reg. Guide 1.174 provides guidance for determining the risk impact of plant-specific changesto the licensing basis. Reg. Guide 1.174 defines "very small" changes in risk as resulting inincreases of CDF below 1.OE-06/yr and increases in LERF below 1.OE-07/yr. "Small" changesin risk are defined as increases in CDF below 1.0E-05/yr and increases in LERF below 1.OE-06/yr. Since the ILRT extension was demonstrated to have no impact on CDF for IP2, therelevant criterion is LERF. The increase in internal events LERF resulting from a change in theType A ILRT test interval for the base case with corrosion included for IP2 is estimated at9.84E-08 /yr (see Attachment 3, Table 5.6-1A), which is within the small change region of theacceptance guidelines in Reg. Guide 1.174. In using the EPRI Expert Elicitation methodology,the change is estimated as 1.05E-08 /yr (see Attachment 3, Table 6.2-2A), which is within thevery small change region of the acceptance guidelines in Reg. Guide 1.174.2. The change in dose risk for changing the Type A test frequency from three-per-ten years toonce-per-fifteen-years, measured as an increase to the total integrated dose risk for all internalevents accident sequences is 0.584 person-rem/yr (0.62%) using the EPRI guidance with thebase case corrosion case (Attachment 3, Table 5.6-1A). The change in dose risk drops to0.111 person-rem/yr when using the EPRI Expert Elicitation methodology (Attachment 3, Table6.2-2A).

NL-14-128Docket No. 50-247Attachment 1Page 16 of 193. The increase in the conditional containment failure frequency from the three in ten year intervalto one in fifteen years including corrosion effects using the EPRI guidance (see Section 5.5) is0.84% for IP2. This value drops to less that 0.10% for IP2 using the EPRI Expert Elicitationmethodology (see Attachment 3 Table 6.2-2A). This is below the acceptance criteria of lessthan 1.5% defined Attachment 3 in Section 1.3.4. To determine the potential impact from external events, a bounding assessment from the riskassociated with external events utilizing information from the IP2 IPEEEs similar to theapproach used in the License Renewal SAMA analysis. As shown in Attachment 3 Table 5.7-2A the total increase in LERF for IP2 due to internal events and the bounding external eventsassessment is 5.20E-07/yr. This value is in Region II of the Reg. Guide 1.174 acceptanceguidelines.5. As shown in Attachment 3, Table 5.7-4, the same bounding analysis indicates that the totalLERF from both internal and external risks is 6.78E-06/yr for IP2, which is less than the Reg.Guide 1.174 limit of 1.OE-05/yr given that the ALERF is in Region II (small change in risk).6. Finally, since the external events assessment led to exceeding one of the two alternativeacceptance criteria (i.e. greater than 1.0 person-rem/yr, an alternative detailed boundingexternal events assessment was also performed to demonstrate that the alternate 1.0%person-rem/yr criterion and the other acceptance criteria could still be met. In this case, asshown in Attachment 3, Table 5.7-7 for IP2, the total change in LERF from both internal andexternal events was 5.52E-7/yr, the change in person-rem/yr was 3.28/yr representing 0.59%of the total, and the change in the CCFP was 0.89%. All of these calculated changes meet theacceptance criteria. As shown in Attachment 3, Table 5.7-8, this assessment indicates that thetotal LERF from both internal and external risks is 2.65E-06/yr for IP2, which is less than theReg. Guide 1.174 limit of 1.OE-05/yr given that the ALERF is in Region II (small change in risk).7. Including age-adjusted steel liner corrosion effects in the ILRT assessment was demonstratedto be a small contributor to the impact of extending the ILRT interval for IP2.Therefore, increasing the ILRT interval on a permanent basis to a one-in-fifteen yearfrequency is not considered to be risk significant. Details of the IP2 risk assessment are containedin Attachment 3.4.6 ConclusionNEI 94-01, Revision 2A, describes an NRC-accepted approach for implementing theperformance-based requirements of 10 CFR 50, Appendix J, Option B. It incorporates theregulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals to15 years. NEI 94-01, Revision 2A delineates a performance-based approach for determiningType A, Type B, and Type C containment leakage rate surveillance test frequencies. IP2 isproposing to adopt the guidance of NEI 94-01, Revision 2A for the 10 CFR 50, Appendix J, testingprogram plan and the ANSI/ANS 56.8 -2002 standard for Type A, B and C tests..Based on the previous ILRT tests conducted at IP2, supplemented by risk analysis studies,including the IP2 risk analysis provided in Attachment 3, it may be concluded thatextension of the containment ILRT interval from ten to 15 years represents minimal riskperformed in accordance with Option B and inspected per the guidance NEI-94-01 Revision 2A.

NL-14-128Docket No. 50-247Attachment 1Page 17 of 195.0 REGULATORY ANALYSIS5.1 No Significant Hazards ConsiderationEntergy has evaluated the safety significance of the proposed change to the IP2 TS which reviseIP2 TS 3.5.15, "Containment Leakage Rate Testing Program," to allow a permanent extension tothe frequency of Type A testing based upon performance criteria. The proposed changes havebeen evaluated according to the criteria of 10 CFR 50.92, "Issuance of Amendment". Entergy hasdetermined that the subject changes do not involve a Significant Hazards Consideration, asdiscussed below1. Does the proposed amendment involve a significant increase in the probabilityor consequences of an accident previously evaluated?Response: No.The proposed amendment involves changes to the IP2 containment leakage rate testingprogram. The proposed amendment does not involve a physical change to the plant or achange in the manner in which the plant is operated or controlled. The primarycontainment function is to provide an essentially leak tight barrier against the uncontrolledrelease of radioactivity to the environment for postulated accidents. As such, thecontainment itself and the testing requirements to periodically demonstrate the integrity ofthe containment exist to ensure the plant's ability to mitigate the consequences of anaccident do not involve any accident precursors or initiators. Therefore, the probability ofoccurrence of an accident previously evaluated is not significantly increased bythe proposed amendment.The proposed amendment adopts the NRC accepted guidelines of NEI 94-01, Revision2A, for development of the IP2 performance-based testing program for the Type A testing.Implementation of these guidelines continues to provide adequate assurance that duringdesign basis accidents, the primary containment and its components would limit leakagerates to less than the values assumed in the plant safety analyses. The potentialconsequences of extending the ILRT interval to 15 years have been evaluated byanalyzing the resulting changes in risk. The increase in risk in terms of person-rem peryear within 50 miles resulting from design basis accidents was estimated to be acceptablysmall and determined to be within the guidelines published in RG 1.174. Additionally, theproposed change maintains defense-in-depth by preserving a reasonable balance amongprevention of core damage, prevention of containment failure, and consequencemitigation. Entergy has determined that the increase in conditional containment failureprobability due to the proposed change would be very small. Therefore, it is concludedthat the proposed amendment does not significantly increase the consequences of anaccident previously evaluated.Therefore, the proposed change does not involve a significant increase in theprobability or consequences of an accident previously evaluated.

NL-14-128Docket No. 50-247Attachment 1Page 18 of 192. Does the proposed amendment create the possibility of a new or differentkind of accident from any accident previously evaluated?Response: No.The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 2A,for the development of the IP2 performance-based leakage testing program, andestablishes a 15-year interval for the performance of the containment ILRT. Thecontainment and the testing requirements to periodically demonstrate the integrity of thecontainment exist to ensure the plant's ability to mitigate the consequences of an accidentdo not involve any accident precursors or initiators. The proposed change does not involvea physical change to the plant (i.e., no new or different type of equipment will be installed)or a change to the manner in which the plant is operated or controlled.Therefore, the proposed change does not create the possibility of a new or different kindof accident from any previously evaluated.3. Does the proposed amendment involve a significant reduction in a margin ofsafety?Response: No.The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 2A,for the development of the IP2 performance-based leakage testing program, and establishesa 15-year interval for the performance of the containment ILRT. This amendment does notalter the manner in which safety limits, limiting safety system setpoints, or limiting conditionsfor operation are determined. The specific requirements and conditions of the containmentleakage rate testing program, as defined in the TS, ensure that the degree of primarycontainment structural integrity and leak-tightness that is considered in the plant's safetyanalysis is maintained. The overall containment leakage rate limit specified by the TS ismaintained, and the Type A containment leakage tests would be performed at the frequenciesestablished in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 2A withno change to the 60 month frequencies of Type B, and Type C tests.Containment inspections performed in accordance with other plant programs serve to providea high degree of assurance that the containment would not degrade in a manner that is notdetectable by an ILRT. A risk assessment using the current IP2 PSA model concluded thatextending the ILRT test interval from ten years to 15 years results in a very small change to therisk profile.Therefore, the proposed change does not involve a significant reduction in a margin ofsafety.Based on the above, Entergy concludes that the proposed amendment to the Indian Point 2Technical Specifications presents no significant hazards consideration under the standards setforth in 10 CFR 50.92(c), and accordingly, a finding of 'no significant hazards consideration' isjustified.

NL-14-128Docket No. 50-247Attachment 1Page 19 of 195.2 Applicable Regulatory Requirements / CriteriaThe NRC Order of February 11, 1980 required an evaluation of the degree of compliance with theGDC at the time. This section discusses continued compliance with certain of those criteria.The plant will continue to meet Criterion 1 of 10 CFR 50.36 which says "Structures, systems andcomponents important to safety shall be designed, fabricated, erected, and tested to qualitystandards commensurate with the importance of the safety functions to be performed. Wheregenerally recognized codes and standards are used, they shall be identified and evaluated todetermine their applicability, adequacy, and sufficiency and shall be supplemented or modified asnecessary to assure a quality product in keeping with the required safety function. A qualityassurance program shall be established and implemented in order to provide adequate assurancethat these structures, systems and components will satisfactorily perform their safety functions.Appropriate records of the design, fabrication, erection, and testing of structures, systems andcomponents important to safety shall be maintained by or under the control of the nuclear powerplant licensee throughout the life of the unit' and Criterion 3 which says "Structures, systems, andcomponents important to safety shall be designed to withstand the effects of natural phenomenasuch as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capabilityto perform their safety functions. The design bases for these structures, systems and componentsshall reflect: (1) appropriate consideration of the most severe of the natural phenomena that havebeen historically reported for the site and surrounding area, with sufficient margin for the limitedaccuracy, quantity, and period of time in which the historical data have been accumulated, (2)appropriate combinations of the effects of normal and accident conditions with the effects of thenatural phenomena and (3) the importance of the safety functions to be performed."The extension of the duration of the ILRT for the containment will not affect the design, fabrication,or construction of the containment structure and the design will continue to account for the effectsof natural phenomena. The ILRT of the containment will continue to be done in accordance with10 CFR 50 Appendix J using 10 CFR 50 Appendix B quality standards. The frequency of the ILRTis being changed in accordance with standards reviewed and approved as compliant withAppendix J. Therefore there will be no instances where the applicable regulatory criteria are notmet.5.3 Environmental ConsiderationsThe proposed changes to the IP2 TS do not involve (i) a significant hazards consideration, (ii) asignificant change in the types or significant increase in the amounts of any effluent that may bereleased offsite, or (iii) a significant increase in individual or cumulative occupational radiationexposure. Accordingly, the proposed amendment meets the eligibility criterion for categoricalexclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), noenvironmental impact statement or environmental assessment need be prepared in connectionwith the proposed amendment.PRECEDENCEThis request is similar in nature to the license amendment authorized by the NRC on April 22,2012 for the Palisades Nuclear Plant (TAC No. ME5997, Accession Number ML1 20740081).

ATTACHMENT 2 TO NL-14-128MARKED UP TECHNICAL SPECIFICATIONS PAGES FOR PROPOSEDCHANGES REGARDING 15 YEAR CONTAINMENT ILRTChanges indicated by lineout for deletion and Bold/Italics for additionsUnit 2 Affected Pages:5.5-14ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NO. 2DOCKET NO. 50-247 Programs and Manuals5.55.5 Programs and Manuals5.5.13 Safety Function Determination Program (SFDP) (continued)The SFDP identifies where a loss of safety function exists. If a loss of safetyfunction is determined to exist by this program, the appropriate Conditions andRequired Actions of the LCO in which the loss of safety function exists are requiredto be entered. When a loss of safety function is caused by the inoperability of asingle Technical Specification support system, the appropriate Conditions andRequired Actions to enter are those of the support system.5.5.14 Containment Leakage Rate Testing Programa. A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordance withNEI 94-01, Revision 2A, "Industry Guidelines for ImplementingPerformance-Based Option of 10 CFR Part 50, Appendix J," October2008.the guidlinS containd in r .egulator; Guide 1.163, "P,,f.rman...Based Cont-ainmont Leak Toct Program," dated Soptombor, 1995.b. The calculated peak containment internal pressure for the design basis loss ofcoolant accident, Pa, is assumed to be the containment design pressure of47 psig.c. The maximum allowable containment leakage rate, La, at P,, and 271 OF shallbe 0.1% of containment steam air weight per day.d. Leakage rate acceptance criteria:1. Containment leakage rate acceptance criterion is 1.0 La. During the firstunit startup following testing in accordance with this program, theleakage rate acceptance criteria are < 0.60 La for the Type B and C testsand 0.75 La for Type A tests.2. Air lock testing acceptance criteria shall be established to ensure thatlimits for Type B and C testing in Technical Specification 5.5.14.d.1 aremet.(continued)INDIAN POINT 25.5- 14Amendment No. 262 ATTACHMENT 3 TO NL-14-128RISK IMPACT OF EXTENDING THE ILRT INTERVAL ASSOCIATEDWITH THE PROPOSED TECHNICAL SPECIFICATION CHANGESENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NO. 2DOCKET NO. 50-247

-allRISK ASSESSMENT FOR INDIAN POINTREGARDING THE ILRT (TYPE A)PERMANENT EXTENSION REQUESTPrepared for:0U-EntergyEntergy Services, Inc.1340 Echelon Parkway, M-ECH-492Jackson, MS 39213October 2013glneerlpg and Research, Znc.158 West Gay StreetSuite 400West Chester, PA 19380(610) 431-8260 RISK ASSESSMENT FOR INDIAN POINT REGARDING THEILRT (TYPE A) PERMANENT EXTENSION REQUESTRevision 0Prepared for:IEntergEntergy Services, Inc.1340 Echelon Parkway, M-ECH-492Jackson, MS 39213Prepared by:158 West Gay Street, Suite 400West Chester, PA 19380(610) 431-8260Document No. 0247-13-0002-4722Prepared by:Reviewed by:Approved by:Donald E. VanoverDonald E. MacLeodJeff R. GaborDate: 016 /201 3Date: //// --) 61-3Date:

Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE OF CONTENTSSection Page1.0 PURPO SE O F A NA LYSIS ................................................................................ 1-11 .1 P U R PO S E ......................................................................................... 1-11.2 BA C K G R O U N D .................................................................................. 1-11.3 ACCEPTANCE CRITERIA ........................................ 1-22 .0 M ET H O D O LO G Y .......................................................................................... 2-13 .0 G R O U N D R U LES .......................................................................................... 3-14 .0 IN P U T S ...................................................................................................... 4 -14.1 GENERAL RESOURCES AVAILABLE ....................................................... 4-14.2 PLANT-SPECIFIC INPUTS .................................................................... 4-64.3 IMPACT OF EXTENSION ON DETECTION OF COMPONENT FAILURESTHAT LEAD TO LEAKAGE (SMALL AND LARGE) ...................................... 4-134.4 IMPACT OF EXTENSION ON DETECTION OF STEEL LINER CORROSIONTHAT LEADS TO LEAKAGE ................................................................. 4-155 .0 R E S U LT S ................................................................................................... 5 -15.1 STEP 1 -QUANTIFY THE BASE-LINE RISK IN TERMS OF FREQUENCYPER REA CTO R YEA R ........................................................................... 5-25.2 STEP 2 -DEVELOP PLANT-SPECIFIC PERSON-REM DOSE (POPULATIONDOSE) PER REACTOR YEAR ................................................................. 5-65.3 STEP 3 -EVALUATE RISK IMPACT OF EXTENDING TYPE A TESTINTERVAL FROM 10-TO-15 YEARS ...................................................... 5-135.4 STEP 4 -DETERMINE THE CHANGE IN RISK IN TERMS OF LARGEEARLY RELEASE FREQUENCY ............................................................. 5-225.5 STEP 5 -DETERMINE THE IMPACT ON THE CONDITIONALCONTAINMENT FAILURE PROBABILITY ................................................ 5-225.6 SUMMARY OF INTERNAL EVENTS RESULTS .......................................... 5-235.7 EXTERNAL EVENTS CONTRIBUTION .................................................... 5-265.7.1 Indian Point 2 External Events Discussion ............................... 5-265.7.2 Indian Point 3 External Events Discussion ............................... 5-295.7.3 Additional Seism ic Risk Discussion ......................................... 5-315.7.4 External Events Impact Sum mary .......................................... 5-315.7.5 External Events Impact on ILRT Extension Assessment ............. 5-325.7.6 Alternative Approach for External Events Impact on ILRT ExtensionA ssessm ent ......................................................................... 5-365.8 CONTAINMENT OVERPRESSURE IMPACTS ON CDF ................................ 5-476 .0 S EN S IT IV IT IES ........................................................................................... 6-16.1 SENSITIVITY TO CORROSION IMPACT ASSUMPTIONS ............................ 6-16.2 EPRI EXPERT ELICITATION SENSITIVITY .............................................. 6-47 .0 C O N C LU S IO N S ........................................................................................... 7-18 .0 R E FE R E N C ES .............................................................................................. 8 -1APPENDIX A PRA TECHNICAL ADEQUACYP0247130002-4722 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyList of TablesTable 4.1-1 EPRI/NEI Containment Failure Classifications ........................................... 4-4Table 4.2-1 Level 2 Release Category Frequencies for IP2 and IP3 ............................... 4-7Table 4.2-2 Release Category Definitions from the License Renewal Effort ..................... 4-8Table 4.2-3 Population Dose per License Renewal Release Category for IP2 and IP3 ....... 4-8Table 4.2-4 Population Dose for Intact Containment Cases for IP2 and IP3 .................... 4-9Table 4.2-5 Weighted Average Population Dose for Intact Containment Case for IP2a n d IP 3 ............................................................................................. 4 -1 0Table 4.2-6a IP2 Population Dose and Population Dose Risk Organized by EPRIRelease C ategory ................................................................................ 4-11Table 4.2-6b IP3 Population Dose and Population Dose Risk Organized by EPRIRelease C ategory ................................................................................ 4-12Table 4.4-1 Steel Liner Corrosion Base Case ........................................................... 4-17Table 5.0-1 A ccident C lasses .................................................................................. 5-1Table 5.1-1 Radionuclide Release Frequencies As A Function Of Accident Class (IP2and IP3 Base C ase) ............................................................................... 5-6Table 5.2-1 IP2 and IP3 Population Dose for Population Within 50 Miles ....................... 5-8Table 5.2-2a IP2 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 3 in 10 Year ILRT Frequency .............................................. 5-9Table 5.2-2b IP3 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 3 in 10 Year ILRT Frequency ............................................ 5-11Table 5.3-1a IP2 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 1 in 10 Year ILRT Frequency ............................................ 5-14Table 5.3-1b IP3 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 1 in 10 Year ILRT Frequency ............................................ 5-16Table 5.3-2a IP2 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 1 in 15 Year ILRT Frequency ............................................ 5-18Table 5.3-2b IP3 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 1 in 15 Year ILRT Frequency ............................................ 5-20Table 5.5-1 IP2 and IP3 ILRT Conditional Containment Failure Probabilities ................. 5-23Table 5.6-1a IP2 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (IncludingAge Adjusted Steel Liner Corrosion Likelihood) ........................................ 5-24Table 5.6-1b IP3 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (IncludingP0247130002-4722ii Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyAge Adjusted Steel Liner Corrosion Likelihood) ........................................ 5-25Table 5.6-2 IP2 and IP3 ILRT Extension Comparison to Acceptance Criteria ................. 5-26Table 5.7-1 External Events Contributor Summary [20] ........................................... 5-32Table 5.7-2a IP2 3b (LERF/YR) as a Function of ILRT Frequency for Internal andExternal Events (Including Age Adjusted Steel Liner Corrosion Likelihood) .. 5-33Table 5.7-2b IP3 3b (LERF/YR) as a Function of ILRT Frequency for Internal andExternal Events (Including Age Adjusted Steel Liner Corrosion Likelihood) .. 5-33Table 5.7-3 Comparison to Acceptance Criteria Including External EventsContribution for IP2 and IP3 ................................................................. 5-35Table 5.7-4 Impact of 15-yr ILRT Extension on LERF for IP2 and IP3 .......................... 5-36Table 5.7-5a Population Dose Risk As A Function Of Accident Class (IP2 AlternativeExternal Events Base Case) .................................................................. 5-41Table 5.7-5b Population Dose Risk As A Function Of Accident Class (IP3 AlternativeExternal Events Base Case) .................................................................. 5-42Table 5.7-6a Population Dose Risk As a Function of Accident Class (IP2 AlternativeExternal Events Evaluation Characteristic of Conditions For 1 in 15 YearILRT Frequency) ................................................................................. 5-4 3Table 5.7-6b Population Dose Risk As A Function Of Accident Class (IP3 AlternativeExternal Events Evaluation Characteristic of Conditions For 1 in 15 YearILRT Frequency) ................................................................................. 5-44Table 5.7-7 Comparison to Acceptance Criteria Including Alternative External EventsEvaluation Contribution for IP2 and IP3 .................................................. 5-45Table 5.7-8 Impact of 15-yr ILRT Extension on LERF for IP2 and IP3 .......................... 5-46Table 6.1-1a Steel Liner Corrosion Sensitivity Cases for IP2 ........................................ 6-1Table 6.1-1b Steel Liner Corrosion Sensitivity Cases for IP3 ........................................ 6-3Table 6.2-1 EPRI Expert Elicitation Results ................................................................ 6-4Table 6.2-2a IP2 ILRT Cases: 3 in 10 (Base Case), 1 in 10, and 1 in 15 Yr intervals(Based on EPRI Expert Elicitation Leakage Probabilities) ............................. 6-6Table 6.2-2b IP3 ILRT Cases: 3 in 10 (Base Case), 1 in 10, and 1 in 15 Yr intervals(Based on EPRI Expert Elicitation Leakage Probabilities) ............................. 6-7Table A.2-1 Summary of Industry Peer Review Findings for the IP2 Internal EventsPRA M odel U pdate ................................................................................. A -7Table A.2-2 Summary of Industry Peer Review Findings for the IP3 Internal EventsPRA M odel Update ............................................................................ A-18P0247130002-4722iii Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy1.0 PURPOSE OF ANALYSIS1.1 PURPOSEThe purpose of this analysis is to provide an assessment of the risk associated withimplementing a permanent extension of the Indian Point Units 2 and 3 (IP2 and IP3)containment Type A integrated leak rate test (ILRT) interval from ten years to fifteen years.The risk assessment follows the guidelines from NEI 94-01 [1], the methodology outlined inEPRI TR-104285 [2], the EPRI Risk Impact Assessment of Extended Integrated Leak RateTesting Intervals [3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment(PRA) findings and risk insights in support of a request for a plant's licensing basis as outlinedin Regulatory Guide (RG) 1.174 [4], and the methodology used for Calvert Cliffs to estimatethe likelihood and risk implications of corrosion-induced leakage of steel liners goingundetected during the extended test interval [5]. The format of this document is consistentwith the intent of the Risk Impact Assessment Template for evaluating extended integratedleak rate testing intervals provided in the October 2008 EPRI final report [3].1.2 BACKGROUNDRevisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the IntegratedLeak Rate Test (ILRT) Type A surveillance testing requirements from three-in-ten years to atleast once per ten years. The revised Type A frequency is based on an acceptableperformance history defined as two consecutive periodic Type A tests at least 24 months apartin which the calculated performance leakage was less than the normal containment leakage of1.OLa (allowable leakage).The basis for a 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, andwas established in 1995 during development of the performance-based Option B to Appendix J.Section 11.0 of NEI 94-01 states that NUREG-1493 [6], "Performance-Based ContainmentLeak Test Program," provides the technical basis to support rulemaking to revise leakage ratetesting requirements contained in Option B to Appendix J. The basis consisted of qualitativeand quantitative assessments of the risk impact (in terms of increased public dose) associatedwith a range of extended leakage rate test intervals. To supplement the NRC's rulemakingbasis, NEI undertook a similar study. The results of that study are documented in ElectricPower Research Institute (EPRI) Research Project Report TR-104285 [2].The NRC report on performance-based leak testing, NUREG-1493, analyzed the effects ofcontainment leakage on the health and safety of the public and the benefits realized from thecontainment leak rate testing. In that analysis, it was determined for a representative PWRP0247130002-47221-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacyplant (i.e., Surry) that containment isolation failures contribute less than 0.1 percent to thelatent risks from reactor accidents. Because ILRTs represent substantial resourceexpenditures, it is desirable to show that extending the ILRT interval will not lead to asubstantial increase in risk from containment isolation failures to support a reduction in thetest frequency for IP2 and IP3.Earlier ILRT frequency extension submittals have used the EPRI TR-104285 [2] methodologyto perform the risk assessment. In October 2008, EPRI 1018243 [3] was issued to develop ageneric methodology for the risk impact assessment for ILRT interval extensions to 15 yearsusing current performance data and risk informed guidance, primarily NRC Regulatory Guide1.174 [4]. This more recent EPRI document considers the change in population dose, largeearly release frequency (LERF), and containment conditional failure probability (CCFP),whereas EPRI TR-104285 considered only the change in risk based on the change in populationdose. This ILRT interval extension risk assessment for IP2 and IP3 employs the EPRI 1018243methodology, with the affected System, Structure, or Component (SSC) being the primarycontainment boundary.1.3 ACCEPTANCE CRITERIAThe acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanentextension of the Type A test interval beyond that established during the Option B rulemakingof Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines asincreases in core damage frequency (CDF) less than 1.OE-06 per reactor year and increases inlarge early release frequency (LERF) less than 1.OE-07 per reactor year. Note that a separatediscussion in Section 5.8 confirms that the CDF is not impacted by the proposed change for IP2and IP3. Therefore, since the Type A test does not impact CDF for IP2 and IP3, the relevantcriterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 1.OE-06per reactor year, provided that the total LERF from all contributors (including external events)can be reasonably shown to be less than 1.OE-05 per reactor year. RG 1.174 discussesdefense-in-depth and encourages the use of risk analysis techniques to help ensure and showthat key principles, such as the defense-in-depth philosophy, are met. Therefore, the increasein the conditional containment failure probability (CCFP) is also calculated to help ensure thatthe defense-in-depth philosophy is maintained.With regard to population dose, examinations of NUREG-1493 and Safety Evaluation Reports(SERs) for one-time interval extension (summarized in Appendix G of [3]) indicate a range ofincremental increases in population dose1 that have been accepted by the NRC. The range of1 The one-time extensions assumed a large leak (EPRI class 3b) magnitude of 35La, whereas thisP0247130002-47221-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacyincremental population dose increases is from _<0.01 to 0.2 person-rem/yr and 0.002 to 0.46%of the total accident dose. The total doses for the spectrum of all accidents (Figure 7-2 ofNUREG-1493) result in health effects that are at least two orders of magnitude less than theNRC Safety Goal Risk. Given these perspectives, the NRC SER on this issue [7] defines a smallincrease in population dose as an increase of 5 1.0 person-rem per year, or 51 0% of the totalpopulation dose, whichever is less restrictive for the risk impact assessment of the extendedILRT intervals. This definition has been adopted by the IP2/IP3 analysis.The acceptance criteria are summarized below.1. The estimated risk increase associated with permanently extending the ILRTsurveillance interval to 15 years must be demonstrated to be small. (Note thatRegulatory Guide 1.174 defines very small changes in risk as increases in CDFless than 1.OE-6 per reactor year and increases in LERF less than 1.OE-7 perreactor year. Since the type A ILRT test is not expected to impact CDF forIndian Point, the relevant risk metric is the change in LERF. Regulatory Guide1.174 also defines small risk increase as a change in LERF of less than 1.OE-6reactor year.) Therefore, a small change in risk for this application is definedas a LERF increase of less than 1.OE-6.2. Per the NRC SE, a small increase in population dose is also defined as anincrease in population dose of less than or equal to either 1.0 person-rem peryear or 1 percent of the total population dose, whichever is less restrictive.3. In addition, the SE notes that a small increase in Conditional ContainmentFailure Probability (CCFP) should be defined as a value marginally greater thanthat accepted in previous one-time 15-year ILRT extension requests (typicallyabout 1% or less, with the largest increase being 1.2%). This would requirethat the increase in CCFP be less than or equal to 1.5 percentage points.analysis uses lOOLa.P0247130002-47221-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy2.0 METHODOLOGYA simplified bounding analysis approach consistent with the EPRI methodology is used forevaluating the change in risk associated with increasing the test interval to fifteen years [3].The analysis uses results from a Level 2 analysis of core damage scenarios from the currentIP2 and IP3 PRA analyses of record and the subsequent containment responses to establishthe various fission product release categories including the release size.The six general steps of this assessment are as follows:1. Quantify the baseline risk in terms of the frequency of events (per reactor year) foreach of the eight containment release scenario types identified in the EPRI report [3].2. Develop plant-specific population dose rates (person-rem per reactor year) for each ofthe eight containment release scenario types from plant specific consequence analyses.3. Evaluate the risk impact (i.e., the change in containment release scenario typefrequency and population dose) of extending the ILRT interval to fifteen years.4. Determine the change in risk in terms of Large Early Release Frequency (LERF) inaccordance with RG 1.174 and compare this change with the acceptance guidelines ofRG 1.174 [4].5. Determine the impact on the Conditional Containment Failure Probability (CCFP)6. Evaluate the sensitivity of the results to assumptions in the liner corrosion analysis andto variations in the fractional contributions of large isolation failures (due to linerbreach) to LERF.Furthermore," Consistent with the previous industry containment leak risk assessments, the IP2and IP3 assessment uses population dose as one of the risk measures. The otherrisk measures used in the IP2 and IP3 assessment are the conditional containmentfailure probability (CCFP) for defense-in-depth considerations, and change in LERF todemonstrate that the acceptance guidelines from RG 1.174 are met." This evaluation for IP2 and IP3 uses ground rules and methods to calculate changesin the above risk metrics that are consistent with those outlined in the current EPRImethodology [3].P0247130002-47222-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy3.0 GROUND RULESThe following ground rules are used in the analysis:" The IP2 and IP3 Level 1 and Level 2 internal events PRA models providerepresentative core damage frequency and release category frequency distributionsto be utilized in this analysis." It is appropriate to use the IP2 and IP3 internal events PRA model as a gauge toeffectively describe the risk change attributable to the ILRT extension. It isreasonable to assume that the impact from the ILRT extension (with respect topercent increases in population dose) will not substantially differ if external eventswere to be included in the calculations; however, external events have beenaccounted for in the analysis based on the available information from the IP2 and IP3IPEEEs [8, 9] as reported and used in the IP2 and IP3 SAMA analysis performed aspart of the License Renewal efforts as described in Section 5.7." Dose results for the containment failures modeled in the PRA can be characterized byinformation that was prepared to support the SAMA analysis as part of the LicenseRenewal effort [10]. This information is supplemented with revised calculations [11]for the base case containment intact scenarios which are critical for use in the ILRTextension assessment.* Accident classes describing radionuclide release end states and their definitions areconsistent with the EPRI methodology [3] and are summarized in Section 4.2." The representative containment leakage for Class 1 sequences is 1La. Class 3accounts for increased leakage due to Type A inspection failures." The representative containment leakage for Class 3a is 10 La and for Class 3bsequences is 10OLa, based on the recommendations in the latest EPRI report [3] andas recommended in the NRC SE on this topic [7]. It should be noted that this ismore conservative than the earlier previous industry ILRT extension requests, whichutilized 35La for the Class 3b sequences." Based on the EPRI methodology and the NRC SE, the Class 3b sequences arecategorized as LERF and the increase in Class 3b sequences is used as a surrogatefor the ALERF metric." The impact on population doses from containment bypass scenarios is not altered bythe proposed ILRT extension, but is accounted for in the EPRI methodology as aseparate entry for comparison purposes. Since the containment bypass contributionto population dose is fixed, no changes on the conclusions from this analysis willresult from this separate categorization." The reduction in ILRT frequency does not impact the reliability of containmentisolation valves to close in response to a containment isolation signal.* The use of the estimated 2035 population data from the MACCS2 off-siteconsequence runs [10, 11] is appropriate for this analysis. This assumption isconsistent with that made in the SAMA analysis.* An evaluation of the risk impact of the ILRT on shutdown risk is addressed using thegeneric results from EPRI TR-105189 [12].P0247130002-47223-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy4.0 INPUTSThis section summarizes the general resources available as input (Section 4.1) and the plantspecific resources required (Section 4.2).4.1 GENERAL RESOURCES AVAILABLEVarious industry studies on containment leakage risk assessment are briefly summarized here:1. NUREG/CR-3539 [13]2. NUREG/CR-4220 [14]3. NUREG-1273 [15]4. NUREG/CR-4330 [16]5. EPRI TR-105189 [12]6. NUREG-1493 [6]7. EPRI TR-104285 [2]8. Calvert Cliffs liner corrosion analysis [5]9. EPRI 1018243 [3]10. NRC Final Safety Evaluation [7]The first study is applicable because it provides one basis for the threshold that could be usedin the Level 2 PRA for the size of containment leakage that is considered significant and to beincluded in the model. The second study is applicable because it provides a basis of theprobability for significant pre-existing containment leakage at the time of a core damageaccident. The third study is applicable because it is a subsequent study to NUREG/CR-4220that undertook a more extensive evaluation of the same database. The fourth study providesan assessment of the impact of different containment leakage rates on plant risk. The fifthstudy provides an assessment of the impact on shutdown risk from ILRT test intervalextension. The sixth study is the NRC's cost-benefit analysis of various alternative approachesregarding extending the test intervals and increasing the allowable leakage rates forcontainment integrated and local leak rate tests. The seventh study is an EPRI study of theimpact of extending ILRT and LLRT test intervals on at-power public risk. The eighth studyaddresses the impact of age-related degradation of the containment liners on ILRT evaluations.EPRI 1018243 complements the previous EPRI report and provides the results of an expertelicitation process to determine the relationship between pre-existing containment leakageprobability and magnitude. Finally, the NRC Safety Evaluation (SE) documents the acceptanceby the NRC of the proposed methodology with a few exceptions. These exceptions (associatedwith the ILRT Type A tests) were addressed in the Revision 2-A of NEI 94-01 and the finalversion of the updated EPRI report [3], which was used for this application.P0247130002-47224-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyNUREG/CR-3539 [131Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leakrates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 [31] asthe basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rateson LWR accident risks is relatively small.NUREG/CR-4220 [141NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985.The study reviewed over two thousand LERs, ILRT reports and other related records tocalculate the unavailability of containment due to leakage. It assessed the "large" containmentleak probability to be in the range of 1E-3 to 1E-2, with 5E-3 identified as the point estimatebased on 4 events in 740 reactor years and conservatively assuming a one-year duration foreach event.NUREG-1273 r151A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of theNUREG/CR-4220 database. This assessment noted that about one-third of the reported eventswere leakages that were immediately detected and corrected. In addition, this study notedthat local leak rate tests can detect "essentially all potential degradations" of the containmentisolation system.NUREG/CR-4330 [161NUREG/CR-4330 is a study that examined the risk impacts associated with increasing theallowable containment leakage rates. The details of this report have no direct impact on themodeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakagerate and the ILRT test interval extension study focuses on the frequency of testing intervals.However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 andother similar containment leakage risk studies:"...the effect of containment leakage on overall accident risk is small since risk isdominated by accident sequences that result in failure or bypass ofcontainment."P0247130002-47224-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyEPRI TR-105189 r121The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessmentbecause this EPRI study provides insight regarding the impact of containment testing onshutdown risk. This study performed a quantitative evaluation (using the EPRI ORAMsoftware) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT andLLRT test intervals on shutdown risk.The result of the study concluded that a small but measurable safety benefit (shutdown CDFreduced by 1.OE-8/yr to 1.0E-7/yr) is realized from extending the test intervals from 3 per 10years to 1 per 10 years.NUREG-1493 [6]NUREG-1493 is the NRC's cost-benefit analysis for proposed alternatives to reducecontainment leakage testing frequencies and/or relax allowable leakage rates. The NRCconclusions are consistent with other similar containment leakage risk studies:" Reduction in ILRT frequency from 3 per 10 years to 1 per 20 years results in an"imperceptible" increase in risk." Given the insensitivity of risk to the containment leak rate and the small fraction ofleak paths detected solely by Type A testing, increasing the interval betweenintegrated leak rate tests is possible with minimal impact on public risk.EPRI TR-104285 r2lExtending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study),the EPRI TR-104285 study is a quantitative evaluation of the impact of extending IntegratedLeak Rate Test (ILRT) and (Local Leak Rate Test) LLRT test intervals on at-power public risk.This study combined IPE Level 2 models with NUREG-1150 [17] Level 3 population dosemodels to perform the analysis. The study also used the approach of NUREG-1493 [6] incalculating the increase in pre-existing leakage probability due to extending the ILRT and LLRTtest intervals.EPRI TR-104285 used a simplified Containment Event Tree to subdivide representative coredamage sequences into eight categories of containment response to a core damage accident:1. Containment intact and isolated2. Containment isolation failures due to support system or active failures3. Type A (ILRT) related containment isolation failures4. Type B (LLRT) related containment isolation failures5. Type C (LLRT) related containment isolation failuresP0247130002-47224-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy6. Other penetration related containment isolation failures7. Containment failure due to core damage accident phenomena8. Containment bypassConsistent with the other containment leakage risk assessment studies, this study concluded:"These study results show that the proposed CLRT [containment leak ratetests] frequency changes would have a minimal safety impact. The change inrisk determined by the analyses is small in both absolute and relative terms..."Release Category DefinitionsTable 4.1-1 defines the accident classes used in the ILRT extension evaluation, which isconsistent with the EPRI methodology [3]. These containment failure classifications are usedin this analysis to determine the risk impact of extending the Containment Type A test intervalas described in Section 5 of this report.TABLE 4.1-1EPRI/NEI CONTAINMENT FAILURE CLASSIFICATIONSCLASS] DESCRIPTION1 Containment remains intact including accident sequences that do not lead tocontainment failure in the long term. The release of fission products (andattendant consequences) is determined by the maximum allowable leakagerate values La, under Appendix J for that plant2 Containment isolation failures (as reported in the IPEs) include those accidentsin which there is a failure to isolate the containment.3 Independent (or random) isolation failures include those accidents in which thepre-existing isolation failure to seal (i.e., provide a leak-tight containment) isnot dependent on the sequence in progress.4 Independent (or random) isolation failures include those accidents in which thepre-existing isolation failure to seal is not dependent on the sequence inprogress. This class is similar to Class 3 isolation failures, but is applicable tosequences involving Type B tests and their potential failures. These are theType B-tested components that have isolated but exhibit excessive leakage.5 Independent (or random) isolation failures include those accidents in which thepre-existing isolation failure to seal is not dependent on the sequence inprogress. This class is similar to Class 4 isolation failures, but is applicable tosequences involving Type C tests and their potential failures.6 Containment isolation failures include those leak paths covered in the planttest and maintenance requirements or verified per in service inspection andtesting (ISI/IST) program.7 Accidents involving containment failure induced by severe accidentphenomena. Changes in Appendix J testing requirements do not impact theseaccidents.P0247130002-47224-4 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.1-1EPRI/NEI CONTAINMENT FAILURE CLASSIFICATIONSCLASS DESCRIPTION8 Accidents in which the containment is bypassed (either as an initial conditionor induced by phenomena) are included in Class 8. Changes in Appendix Jtesting requirements do not impact these accidents.Calvert Cliffs Liner Corrosion Analysis [51This submittal to the NRC describes a method for determining the change in likelihood, due toextending the ILRT, of detecting liner corrosion, and the corresponding change in risk. Themethodology was developed for Calvert Cliffs in response to a request for additionalinformation regarding how the potential leakage due to age-related degradation mechanismswas factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffsanalysis was performed for a concrete cylinder and dome and a concrete basemat, each with asteel liner. IP2 and IP3 have a similar type of containment.EPRI 1018243 [31This report presents a risk impact assessment for extending integrated leak rate test (ILRT)surveillance intervals to 15 years. This risk impact assessment complements the previousEPRI report, TR-104285, Risk Impact Assessment of Revised Containment Leak Rate TestingIntervals. The earlier report considered changes to local leak rate testing intervals as well aschanges to ILRT testing intervals. The original risk impact assessment considers the change inrisk based on population dose, whereas the revision considers dose as well as large earlyrelease frequency (LERF) and conditional containment failure probability (CCFP). This reportdeals with changes to ILRT testing intervals and is intended to provide bases for supportingchanges to industry and regulatory guidance on ILRT surveillance intervals.The risk impact assessment using the Jeffrey's Non-Informative Prior statistical method isfurther supplemented with a sensitivity case using expert elicitation performed to addressconservatisms. The expert elicitation is used to determine the relationship between pre-existing containment leakage probability and magnitude. The results of the expert elicitationprocess from this report are used as a separate sensitivity investigation for the IP2 and IP3analysis presented here in Section 6.2.P0247130002-47224-5 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyNRC Safety Evaluation Report r7]This SE documents the NRC staff's evaluation and acceptance of NEI TR 94-01, Revision 2, andEPRI Report No. 1009325, Revision 2, subject to the limitations and conditions identified in theSE and summarized in Section 4.0 of the SE. These limitations (associated with the ILRT TypeA tests) were addressed in the Revision 2-A of NEI 94-01 which are also included in Revision3-A of NEI 94-01 [1] and the final version of the updated EPRI report [3]. Additionally, the SEclearly defined the acceptance criteria to be used in future Type A ILRT extension riskassessments as delineated previously in the end of Section 1.3.4.2 PLANT-SPECIFIC INPUTSThe IP2 and IP3 specific information used to perform this ILRT interval extension riskassessment includes the following:* Level 1 and Level 2 PRA model quantification results [18, 19]* Population dose within a 50-mile radius for various release categories [10, 11]IP2 and IP3 Internal Events Core Damage FrequenciesThe current IP2 and IP3 Internal Events PRA analyses of record are based on an event tree /linked fault tree model characteristic of the as-built, as-operated plant. Based on the resultsfound in Tables J1.6-2 of Reference [18] and Reference [19], the internal events Level 1 PRAcore damage frequency (CDF) is 1.17E-05/yr for IP2 and 1.48E-05/yr for IP3.IP2 and IP3 Internal Events Release Category FrequenciesThe Level 2 release category frequencies were developed from the contributions to CDF forthose analyzed containment failure modes that were documented in Tables J1.6-2 and TablesJ1.7-4 for IP2 and IP3 of Reference [18] and Reference [19], respectively. Table 4.2-1summarizes the pertinent IP2 and IP3 results in terms of end-states where a representativerelease category is assigned for each end-state. The total Large Early Release Frequency(LERF) in Table 4.2-1 is 1.16E-06/yr for IP2 and 1.25E-06/yr for IP3. The individual releasecategory frequencies are utilized here to provide the necessary delineation for the ILRT riskassessment with the corresponding EPRI class for each release category. A discussion of theavailable population dose information for various release categories follows this table.P0247130002-47224-6 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.2-1LEVEL 2 RELEASE CATEGORY FREQUENCIES FOR IP2 AND IP3RELEASE CATEGORY DESCRIPTION INDIAN POINT 2 INDIAN POINT 3(FREQUENCY/YR) (FREQUENCY/YR)No Containment Failure 7.86E-06 1.13E-05Late Release 2.71E-06 2.17E-06Low to Moderate Early Release 4.66E-09 1.17E-07High Early Release (LERF) 1.16E-06 1.25E-06LERF: Containment Bypass (SGTRInitiating Events) 9.58E-07 9.19E-07LERF: Containment Bypass (ISLOCA) 2.77E-08 1.93E-07LERF: Containment Bypass (InducedSGTR events) 8.72E-08 5.78E-08LERF: Containment Isolation Failure 1.11E-08 3.99E-09LERF: Energetic Containment Failures 6.90E-08 7.14E-08Total: 1.17E-05 1.48E-05IP2 and IP3 Population Dose InformationIn the License Renewal analysis for IP2 and IP3 [20], the release categories considered themagnitude of the radionuclide release, e.g., concentration of cesium iodide (CsI), and the timeof the release. Table 4.2-2 shows how the different release categories were organized for thelicense renewal effort. While that breakdown was appropriate for that submittal, thebreakdown in Table 4.2-1 is sufficient for this ILRT extension risk assessment.P0247130002-47224-7 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.2-2RELEASE CATEGORY DEFINITIONS FROM THE LICENSE RENEWAL EFFORTRELEASE SEVERITY SOURCE TERMRELEASE TIMING RELEASE FRACTIONCLASSIFICATION TIME OF RELEASE CLASSIFICATION PERCENT CSI INCATEGORY (NOBLE GASES OR CATEGORY RELEASECSI)Late (L) > 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> High (H) > 10Moderate (M) 1 to 10Early (E) < 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Low (L) 0.1 to 1Low-Low (LL) 0.01 to 0.1No Containment < 0.01 (Little to NoFailure (NCF) Release)The population dose results from latest relevant License Renewal submittal [10] form the basisof the initial ILRT assessment using the latest available release category frequency informationas described above. The results for IP2 are taken from Table 5 of Reference [10] and theresults for IP3 are taken from Table 6 of Reference [10]. Those population dose results arereproduced in Table 4.2-3 converted to the corresponding values in person-rem (i.e., 100 *person-sv) used for this analysis.TABLE 4.2-3POPULATION DOSE PER LICENSE RENEWAL RELEASE CATEGORY FOR IP2 AND IP3RELEASE CATEGORY DESCRIPTION INDIAN POINT 2 INDIAN POINT 3(PERSON-REM) (PERSON-REM)No Containment Failure (NCF) 4.75E+03 8.04E+03Early High 6.51E+07 5.08E+07Early Medium 1.94E+07 2.OOE+07Early Low 7.93E+06 5.21E+06Late High 1.63E+07 1.63E+07Late Medium 6.87E+06 6.85E+06Late Low 1.61E+06 1.61E+06Late Low-Low 1.38E+06 1.38E+06P0247130002-47224-8 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacySince the ILRT methodology is based on multipliers to a bounding case which is representativeof an allowable leakage of 1.OLa, the NCF case from the License Renewal effort, whichrepresents a best estimate release, could not be used. As a result, additional analyses wererequired for the ILRT assessment to be consistent with the methodology employed. Table4.2-4 shows the results of four different potential case runs to provide a representative 1.0Larelease [11]. Note that for the containment intact case, given the similarities between IP2 andIP3, the results are assumed to be applicable to both units. These case results arerepresentative of the 1.OLa release as required by the ILRT methodology.TABLE 4.2-4POPULATION DOSE FOR INTACT CONTAINMENT CASES FOR IP2 AND IP3RELEASE CATEGORY DESCRIPTION INDIAN POINT 2 INDIAN POINT 3(PERSON-REM) (PERSON-REM)Intact Scenario #1 (Vessel Breach Occurs,Containment Fan Coolers Available) 8.28E+04 8.28E+04Intact Scenario #2 (Vessel Breach Occurs,Containment Sprays Available) 1.59E+04 1.59E+04Intact Scenario #3 (Vessel Breach Occurs,Fan Coolers and Sprays Available) 1.32E+04 1.32E+04Intact Scenario #4 (No Vessel Breach,Containment Fan Coolers Available) 2.94E+04 2.94E+04Based on a review of cutsets associated with the intact containment end state, anapportionment of the intact containment associated release categories was made. First, it wasnoted that containment sprays were not failed in more than 99% of the intact containmentcases for both IP2 and IP3, but their use could only be definitively declared in Medium andLarge LOCA scenarios or when vessel breach occurs (i.e., other cases with fan coolers availableand no vessel breach are unlikely to reach the automatic containment spray initiation set pointof 24 psig for IP2 and 22 psig for IP3). For IP2 about 68% of the intact containment casesalso involved no vessel breach, and for IP3 about 63% of the intact containment casesinvolved no vessel breach. For IP2 and IP3, the medium and large LOCA contribution to theintact containment case was about 10%. Therefore, it was conservatively assumed that just10% of the intact containment cases could be represented by a case with containment spraysavailable (i.e., intact scenario #2 from Table 4.2-4). Of the remaining 90%, based on theP0247130002-47224-9 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacycontribution from no vessel breach scenarios noted above, it was assumed that about 60% ofthe cases involved scenarios with no vessel failure and about 30% involved scenarios wherevessel failure occurred for both IP2 and IP3. Intact scenario #4 from Table 4.2-4 is then usedas a representative case for the no vessel failure scenarios, and intact scenario #1 is thenconservatively used as a representative case for the remaining vessel failure scenarios.Although sprays are likely available in those scenarios, the SAMG procedures may limit theiruse based on hydrogen detonation concerns. This leads to an overall weighted averagepopulation dose for the intact containment case as shown in Table 4.2-5. This weightedaverage population dose of 4.41E+04 person-rem is used in the remainder of the calculationsusing the ILRT methodology.TABLE 4.2-5WEIGHTED AVERAGE POPULATION DOSE FOR INTACT CONTAINMENT CASE FORIP2 AND IP3RELEASE CATEGORY DESCRIPTION PERCENT POPULATION DOSECONTRIBUTION (PERSON-REM)Intact Scenario #1 (Vessel Breach Occurs,Containment Fan Coolers Available) 30% 8.28E+04Intact Scenario #2 (Vessel Breach Occurs,Containment Sprays Available) 10% 1.59E+04Intact Scenario #3 (Vessel Breach Occurs,Fan Coolers and Sprays Available) N/A 1.32E+04Intact Scenario #4 (No Vessel Breach,Containment Fan Coolers Available) 60% 2.94E+040.3 * (8.28E+04) +Weighted Average 0.1 * (1.59E+04) +0.6 * (2.94E+04) 4.41E+04Population Dose Risk CalculationsThe next step is to take the frequency information from Table 4.2-1, assign each category tothe relevant EPRI release category class from Table 4.1-1, and then associate a representativepopulation dose from Table 4.2-3 or Table 4.2-5 for each release category. Table 4.2-6a liststhe population dose risk and average population dose organized by EPRI release category forIP2, including the delineation of early and late frequencies for Class 7, and a delineation ofSGTR and ISLOCA frequencies for Class 8. Note that the population dose risk (Column 4 ofTable 4.2-6a) was found by multiplying the release category frequency (Column 2 of TableP0247130002-47224-10 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy4.2-6a) by the associated population dose (Column 3 of Table 4.2-6a). The correspondinginformation for IP3 is shown in Table 4.2-6b. Note that only the applicable EPRI releasecategories at this point are shown in the tables (i.e., the Class 3 frequencies are derived laterand the Class 4, 5, and 6 frequencies are not utilized in the EPRI methodology for the ILRTextension risk assessment).IP2 POPULATIONTABLE 4.2-6ADOSE AND POPULATION DOSE RISK ORGANIZEDBY EPRI RELEASE CATEGORYEPRI RELEASE CATEGORY RELEASE ASSIGNED POPULATION DOSEAND DESCRIPTION FREQUENCY POPULATION RISK (PERSON-(1/YR) DOSE (PERSON- REM/YR)REM)1: Containment intact 7.86E-06 4.41E+04 3.47E-01[Weighted AverageFrom Table 4.2-5]2: Large containment 1.11E-08 6.51E+07 7.23E-01isolation failures [Early High FromTable 4.2-3]7-CFE: Phenomena-induced 4.66E-09 1.94E+07 9.04E-02containment failures [Early Medium From(Early-non LERF) Table 4.2-3]7-CFE: Phenomena-induced 6.90E-08 6.51E+07 4.49E+00containment failures [Early High From(Early LERF) Table 4.2-3]7-CFL: Phenomena- 2.71E-06 6.87E+06 1.86E+01induced containment [Late Medium Fromfailures (Late) Table 4.2-3](1)8-SGTR: Containment 1.05E-06 6.51E+07 6.80E+01bypass (SGTR) [Early High FromTable 4.2-3]8-ISLOCA: Containment 2.77E-08 6.51E+07 1.80E+00bypass (ISLOCA) [Early High FromI_ Table 4.2-3]Total: 1.17E-05 94.12) Although the current model does not distinguish between the different late release categories,the weighted average late release from the License Renewal was within 10% of the LateMedium population dose. The use of the Late Medium population dose for this releasecategory was therefore deemed appropriate for the ILRT assessment.P0247130002-47224-11 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.2-6BIP3 POPULATION DOSE AND POPULATION DOSE RISK ORGANIZEDBY EPRI RELEASE CATEGORYEPRI RELEASE CATEGORY RELEASE ASSIGNED POPULATION DOSEAND DESCRIPTION FREQUENCY POPULATION RISK (PERSON-(1/YR) DOSE (PERSON- REM/YR)REM)1: Containment intact 1.13E-05 4.41E+04 4.98E-01[Weighted AverageFrom Table 4.2-5]2: Large containment 3.99E-09 5.08E+07 2.03E-01isolation failures [Early High FromTable 4.2-3]7-CFE: Phenomena-induced 1.17E-07 2.OOE+07 2.34E+00containment failures [Early Medium From(Early-non LERF) Table 4.2-3]7-CFE: Phenomena-induced 7.14E-08 5.08E+07 3.63E+00containment failures [Early High From(Early LERF) Table 4.2-3]7-CFL: Phenomena-induced 2.17E-06 6.85E+06 1.49E+01containment failures [Late Medium From(Late) Table 4.2-3](1)8-SGTR: Containment 9.77E-07 5.08E+07 4.96E+01bypass (SGTR) [Early High FromI Table 4.2-3]8-ISLOCA: Containment 1.93E-07 5.08E+07 9.80E+00bypass (ISLOCA) [Early High FromTable 4.2-3]Total: 1.48E-05 80.96(1) Although the current model does not distinguish between the different late release categories,the weighted average late release from the License Renewal was within 10% of the LateMedium population dose. The use of the Late Medium population dose for this releasecategory was therefore deemed appropriate for the ILRT assessment.The frequencies for the severe accident classes defined in Table 4.1-1 are developed for IP2and IP3 based on the assignments shown above in Tables 4.2-6a and 4.2-6b. Then, thefrequencies for Classes 3a and 3b can be determined with that portion removed from Class 1.This step in the process is described in Section 4.3. Furthermore, adjustments are made tothe Class 3b as well as Class 1 frequencies to account for the impact of undetected corrosion ofthe steel liner per the methodology described in Section 4.4.P0247130002-47224-12 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy4.3 IMPACT OF EXTENSION ON DETECTION OF COMPONENT FAILURES THAT LEAD TOLEAKAGE (SMALL AND LARGE)The ILRT can detect a number of component failures such as liner breach and failure of somesealing surfaces, which can lead to leakage. The proposed ILRT test interval extension mayinfluence the conditional probability of detecting these types of failures. To ensure that thiseffect is properly accounted for, the EPRI Class 3 accident class as defined in Table 4.1-1 isdivided into two sub-classes representing small and large leakage failures. These subclassesare defined as Class 3a and Class 3b, respectively.The probability of the EPRI Class 3a failures may be determined, consistent with the latestEPRI guidance [3], as the mean failure estimated from the available data (i.e., 2 "small"failures that could only have been discovered by the ILRT in 217 tests leads to a2/217=0.0092 mean value). For Class 3b, consistent with latest available EPRI data, a non-informative prior distribution is assumed for no "large" failures in 217 tests (i.e., 0.5/(217+1)= 0.0023).The EPRI methodology contains information concerning the potential that the calculated deltaLERF values for several plants may fall above the "very small change" guidelines of the NRCregulatory guide 1.174. This information includes a discussion of conservatisms in thequantitative guidance for delta LERF. EPRI describes ways to demonstrate that, using plant-specific calculations, the delta LERF is smaller than that calculated by the simplified method.The methodology states:"The methodology employed for determining LERF (Class 3b frequency)involves conservatively multiplying the CDF by the failure probability for thisclass (3b) of accident. This was done for simplicity and to maintainconservatism. However, some plant-specific accident classes leading tocore damage are likely to include individual sequences that either mayalready (independently) cause a LERF or could never cause a LERF, and arethus not associated with a postulated large Type A containment leakagepath (LERF). These contributors can be removed from Class 3b in theevaluation of LERF by multiplying the Class 3b probability by only thatportion of CDF that may be impacted by type A leakage."The application of this additional guidance to the analysis for IP2 and IP3 (as detailed inSection 5) means that the Class 2, Class 7, and Class 8 LERF sequences are subtracted fromthe CDF that is applied to Class 3b. To be consistent, the same change is made to the Class3a CDF, even though these events are not considered LERF. Note that Class 2 events refer tosequences with a large pre-existing containment isolation failure that lead to LERF, a subset ofClass 7 events are LERF sequences due to an early containment failure from energeticphenomena, and Class 8 event are containment bypass events that contribute to LERF.P0247130002-47224-13 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyConsistent with the EPRI methodology [3], the change in the leak detection probability can beestimated by comparing the average time that a leak could exist without detection. Forexample, the average time that a leak could go undetected with a three-year test interval is1.5 years (3 yr / 2), and the average time that a leak could exist without detection for a ten-year interval is 5 years (10 yr / 2). This change would lead to a non-detection probability thatis a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRTtesting, given a 10-year vs. a 3-yr interval. Correspondingly, an extension of the ILRT intervalto fifteen years can be estimated to lead to about a factor of 5.0 (7.5/1.5) increase in the non-detection probability of a leak.IP2 and IP3 Past ILRT ResultsThe surveillance frequency for Type A testing in NEI 94-01 under option B criteria is at leastonce per ten years based on an acceptable performance history (i.e., two consecutive periodicType A tests at least 24 months apart) where the calculated performance leakage rate was lessthan 1.OLa, and in compliance with the performance factors in NEI 94-01, Section 11.3. Basedon the successful completion of two consecutive ILRTs at IP2 and IP3, the current ILRT intervalis once per ten years. Note that the probability of a pre-existing leakage due to extending theILRT interval is based on the industry-wide historical results as noted in the EPRI guidancedocument [3].EPRI MethodoloqyThis analysis uses the approach outlined in the EPRI Methodology [3]. The six steps of themethodology are:1. Quantify the baseline (three-year ILRT frequency) risk in terms of frequency perreactor year for the EPRI accident classes of interest.2. Develop the baseline population dose (person-rem, from the plant PRA or IPE, orcalculated based on leakage) for the applicable accident classes.3. Evaluate the risk impact (in terms of population dose rate and percentile change inpopulation dose rate) for the interval extension cases.4. Determine the risk impact in terms of the change in LERF and the change in CCFP.5. Consider both internal and external events.6. Evaluate the sensitivity of the results to assumptions in the liner corrosion analysis.The first three steps of the methodology deal with calculating the change in dose. The changein dose is the principal basis upon which the Type A ILRT interval extension was previouslygranted and is a reasonable basis for evaluating additional extensions. The fourth step in theP0247130002-47224-14 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacymethodology calculates the change in LERF and compares it to the guidelines in RegulatoryGuide 1.174. Because there is no change in CDF for IP2 and IP3, the change in LERF formsthe quantitative basis for a risk informed decision per current NRC practice, namely RegulatoryGuide 1.174. The fourth step of the methodology calculates the change in containment failureprobability, referred to as the conditional containment failure probability, CCFP. The NRC hasidentified a CCFP of less than 1.5% as the acceptance criteria for extending the Type A ILRTtest intervals as the basis for showing that the proposed change is consistent with the defensein depth philosophy [7]. As such, this step suffices as the remaining basis for a risk informeddecision per Regulatory Guide 1.174. Step 5 takes into consideration the additional risk due toexternal events, and Step 6 investigates the impact on results due to varying the assumptionsassociated with the liner corrosion rate and failure to visually identify pre-existing flaws.4.4 IMPACT OF EXTENSION ON DETECTION OF STEEL LINER CORROSION THAT LEADSTO LEAKAGEAn estimate of the likelihood and risk implications of corrosion-induced leakage of the steelliners occurring and going undetected during the extended test interval is evaluated using themethodology from the Calvert Cliffs liner corrosion analysis [5]. The Calvert Cliffs analysis wasperformed for a concrete cylinder and dome and a concrete basemat, each with a steel liner.IP2 and IP3 have similar containment types.The following approach is used to determine the change in likelihood, due to extending theILRT, of detecting corrosion of the containment steel liner. This likelihood is then used todetermine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the followingissues are addressed:* Differences between the containment basemat and the containment cylinder anddome" The historical steel liner flaw likelihood due to concealed corrosion* The impact of aging" The corrosion leakage dependency on containment pressure" The likelihood that visual inspections will be effective at detecting a flawP0247130002-47224-15 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyAssumptions" A half failure is assumed for the basemat concealed liner corrosion due to lack ofidentified failures.* The two corrosion events over a 5.5 year data period are used to estimate the linerflaw probability in the Calvert Cliffs analysis and are assumed to be applicable to theIP2 and IP3 containment analysis. These events, one at North Anna Unit 2 and oneat Brunswick Unit 2, were initiated from the non-visible (backside) portion of thecontainment liner. It is noted that two additional events have occurred in recentyears (based on a data search covering approximately 9 years documented inReference [21]). In November 2006, the Turkey Point 4 containment building linerdeveloped a hole when a sump pump support plate was moved. In May 2009, a holeapproximately 3/8" by 1" in size was identified in the Beaver Valley 1 containmentliner. For risk evaluation purposes, these two more recent events occurring over a 9year period are judged to be adequately represented by the two events in the 5.5year period of the Calvert Cliffs analysis incorporated in the EPRI guidance (SeeTable 4.4-1, Step 1)." Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumedto double every five years. This is based solely on judgment and is included in thisanalysis to address the increased likelihood of corrosion as the steel liner ages (SeeTable 4.4-1, Steps 2 and 3). Sensitivity studies are included that address doublingthis rate every two years and every ten years.* In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reachingthe outside atmosphere given that a liner flaw exists was estimated as 1.11% for thecylinder and dome region, and 0.11% (10% of the cylinder failure probability) for thebasemat. These values were determined from an assessment of the probability ofcontainment failure versus containment pressure, and the selected values areconsistent with a pressure that corresponds to the ILRT target pressure of 37 psig.For IP2 and IP3, the containment failure probabilities are less than these values at47 psig, which is the containment design pressure [18, 19]. The probabilities of 1%for the cylinder and dome, and 0.1% for the basemat, albeit conservative, are usedin this analysis. Sensitivity studies are included that increase and decrease theprobabilities by an order of magnitude (See Table 4.4-1, Step 4).* Consistent with the Calvert Cliffs analysis, a 5% visual inspection detection failurelikelihood given the flaw is visible and a total detection failure likelihood of 10% isused for the containment cylinder and dome. For the containment basemat, 100% isassumed unavailable for visual inspection. To date, all liner corrosion events havebeen detected through visual inspection (See Table 4.4-1, Step 5). Sensitivitystudies are included that evaluate total detection failure likelihood of 5% and 15%,respectively.* Consistent with the Calvert Cliffs analysis, all non-detectable containment failuresare assumed to result in early releases. This approach avoids a detailed analysis ofcontainment failure timing and operator recovery actions.P0247130002-47224-16 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.4-1STEEL LINER CORROSION BASE CASESTEP DESCRIPTION CONTAINMENT CONTAINMENTCYLINDER AND DOME BASEMATHistorical Steel Liner Events: 2 Events: 0 (assume half aFlaw Likelihood failure)Failure Data: Containment 2/(70

• 5.5) = 5.2E-3 0.5/(70

• 5.5) = 1.3E-3location specific(consistent with CalvertCliffs analysis).2 Age Adjusted Steel Year Failure Rate Year Failure RateLiner Flaw Likelihood 1 2.1E-3 1 5.OE-4During 15-year interval, avg 5-10 5.2E-3 avg 5-10 1.3E-3assume failure rate 15 1.E-2 15 3.5E-3doubles every five years(14.9% increase per year). 15 year average = 15 year average -The average for 5th to 10th 6.27E-3 1.57E-3year is set to the historicalfailure rate (consistentwith Calvert Cliffsanalysis).3 Flaw Likelihood at 3, 0.71% (1 to 3 years) 0.18% (1 to 3 years)10, and 15 years 4.06% (1 to 10 years) 1.04% (1 to 10 years)Uses age adjusted liner 9.40% (1 to 15 years) 2.42% (1 to 15 years)flaw likelihood (Step 2), (Note that the Calvert Cliffs (Note that the Calvertassuming failure rate analysis presents the delta Cliffs analysis presents thedoubles every five years between 3 and 15 years of delta between 3 and 15(consistent with Calvert 8.7% to utilize in the years of 2.2% to utilize inCliffs analysis -See Table estimation of the delta- the estimation of the delta-6 of Reference [5]). LERF value. For this LERF value. For thisanalysis, the values are analysis, however, valuescalculated based on the 3, are calculated based on10, and 15 year intervals.) the 3, 10, and 15 yearintervals.)P0247130002-47224-17 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.4-1STEEL LINER CORROSION BASE CASESTEP DESCRIPTION CONTAINMENT CONTAINMENTCYLINDER AND DOME BASEMAT4 Likelihood of Breach in 1% 0.10/0Containment GivenSteel Liner FlawThe failure probability ofthe containment cylinderand dome is assumed tobe 1% (compared to 1.1%in the Calvert Cliffsanalysis). The basematfailure probability isassumed to be a factor often less, 0.1% (comparedto 0.11% in the CalvertCliffs analysis).5 Visual Inspection 100/% 100%Detection Failure 5% failure to identify visual Cannot be visuallyLikelihood flaws plus 5% likelihood inspected.Utilize assumptions that the flaw is not visibleconsistent with Calvert (not through-cylinder butCliffs analysis. could be detected by ILRT)All events have beendetected through visualinspection. 5% visiblefailure detection is aconservative assumption.6 Likelihood of Non- 0.000710/o (at 3 years) 0.000180/a (at 3 years)Detected Containment =0.71%

• 1%

• 10% =0.18%

• 0.1%

• 100%Leakage(Steps 3

• 4

• 5) 0.00406%/o (at 10 0.001040/a (at 10years) years)=4.06%

• 1%/a

• 10% =1.04%/a

• 0.1%/a

• 100%0.0094% (at 15 years) 0.00242% (at 15=9.40%

• 1%

• 10% years)=2.42%

• 0.1%

• 100%P0247130002-47224-18 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyThe total likelihood of the corrosion-induced, non-detected containment leakage that issubsequently added to the EPRI Class 3b contribution is the sum of Step 6 for the containmentcylinder and dome, and the containment basemat:At 3 years : 0.00071% + 0.00018% = 0.00089%At 10 years: 0.00406% + 0.00104% = 0.00510%At 15 years: 0.0094% + 0.00242% = 0.01182%P0247130002-47224-19 Risk Impact Assessment of Extending the Indian Point ILRT Intervals5.0 RESULTSThe application of the approach based on EPRI Guidance [3] has led to the following results.The results are displayed according to the eight accident classes defined in the EPRI report.Table 5.0-1 lists these accident classes.TABLE 5.0-1ACCIDENT CLASSESACCIDENTCLASSES(CONTAINMENTRELEASE TYPE) DESCRIPTION1 Containment Intact2 Large Isolation Failures (Failure to Close)3a Small Isolation Failures (liner breach)3b Large Isolation Failures (liner breach)4 Small Isolation Failures (Failure to seal -Type B)5 Small Isolation Failures (Failure to seal-Type C)6 Other Isolation Failures (e.g., dependent failures)7 Failures Induced by Phenomena (Early and Late)8 Bypass (SGTR and Interfacing System LOCA)CDF All CET End states (including very low and no release)The analysis performed examined IP2 and IP3 specific accident sequences in which thecontainment remains intact or the containment is impaired. Specifically, the categorization ofthe severe accidents contributing to risk was considered in the following manner:" Core damage sequences in which the containment remains intact initially and in thelong term (EPRI Class 1 sequences).* Core damage sequences in which containment integrity is impaired due to randomisolation failures of plant components other than those associated with Type B orType C test components. For example, liner breach or bellows leakage, if applicable.(EPRI Class 3 sequences)." Core damage sequences in which containment integrity is impaired due tocontainment isolation failures of pathways left "opened" following a plant post-maintenance test. (For example, a valve failing to close following a valve stroketest. (EPRI Class 6 sequences). Consistent with the EPRI Guidance, this class is notspecifically examined since it will not significantly influence the results of thisanalysis.P0247130002-47225-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAccident sequences involving containment bypass (EPRI Class 8 sequences), largecontainment isolation failures (EPRI Class 2 sequences), and small containmentisolation "failure-to-seal" events (EPRI Class 4 and 5 sequences) are accounted for inthis evaluation as part of the baseline risk profile. However, they are not affected bythe ILRT frequency change.Class 4 and 5 sequences are impacted by changes in Type B and C test intervals;therefore, changes in the Type A test interval do not impact these sequences.The steps taken to perform this risk assessment evaluation are as follows:Step 1 Quantify the base-line risk in terms of frequency per reactor year for each of theaccident classes presented in Table 5.0-1.Step 2 Develop plant-specific person-rem dose (population dose) per reactor year foreach of the accident classes.Step 3 Evaluate risk impact of extending Type A test interval from 3 to 15 and 10 to 15years.Step 4 Determine the change in risk in terms of Large Early Release Frequency (LERF)in accordance with RG 1.174.Step 5 Determine the impact on the Conditional Containment Failure Probability(CCFP).5.1 STEP 1 -QUANTIFY THE BASE-LINE RISK IN TERMS OF FREQUENCY PER REACTORYEARThis step involves the review of the IP2 and IP3 Level 2 release category frequency results [18,19]. As described in Section 4.2, the release categories were assigned to the EPRI classes asshown in Table 4.2-6a for IP2 and in Table 4.2-6b for IP3. This application combined with theIP2 and IP3 dose risk (person-rem/yr) also shown in Tables 4.2-6a and 4.2-6b, respectivelyforms the basis for estimating the increase in population dose risk.For the assessment of the impact on the risk profile due to the ILRT extension, the potentialfor pre-existing leaks is included in the model. These pre-existing leak events are representedby the Class 3 sequences in EPRI 1018243 [3]. Two failure modes were considered for theClass 3 sequences, namely Class 3a (small breach) and Class 3b (large breach).The determination of the frequencies associated with each of the EPRI categories listed inTable 5.0-1 is presented next.P0247130002-47225-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsClass 1 SequencesThis group represents the frequency when the containment remains intact (modeled asTechnical Specification Leakage). The frequency per year for these sequences is 7.74E-06/yrfor IP2 and 1.11E-05/yr for IP3 (refer to Table 5.1-1 for Containment Release Type 1) and isdetermined by subtracting all containment failure end states including the EPRI/NEI Class 3aand 3b frequency calculated below, from the total CDF. For this analysis, the associatedmaximum containment leakage for this group is iLa, consistent with an intact containmentevaluation. Note that the values for this Class reported in Table 5.1-1 are slightly lower thanthat reported in Tables 4.2-6a and 4.2-6b since the 3a and 3b frequencies are now subtractedfrom Class 1.Class 2 SequencesThis group consists of large containment isolation failures. For IP2, this frequency is1.11E-08/yr (refer to Table 5.1-1, Containment Release Type 2). For IP3, this frequency is3.99E-09/yr (refer to Table 5.1-1, Containment Release Type 2).Class 3 SequencesThis group represents pre-existing leakage in the containment structure (e.g., containmentliner). The containment leakage for these sequences can be either small (2La to 10OLa) orlarge (>1OOLa). In this analysis, a value of 1OLa was used for small pre-existing flaws and10OLa for relatively large flaws.The respective frequencies per year are determined as follows:PROBciass_3a = probability of small pre-existing containment liner leakage= 0.0092 (see Section 4.3)PROBciass_3b = probability of large pre-existing containment liner leakage= 0.0023 (see Section 4.3)As described in Section 4.3, additional consideration is made to not apply these failureprobabilities to those cases that are already considered LERF scenarios (i.e., the Class 2, Class7, and Class 8 LERF contributions). This adjustment is made for based on the frequencyinformation from Tables 4.2-6a and 4.2-6b for IP2 and IP3, respectively as shown below.P0247130002-47225-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsFor IP2:Class_3a = 0.0092 * [CDF -(Class 2 + Class 7 LERF + Class 8 SGTR + Class 8 ISLOCA)]= 0.0092 * [1.17E-05 -(1.11E-08 + 6.90E-08 + 1.05E-06 + 2.77E-08)]= 9.73E-08/yrClass_3b = 0.0023 * [CDF -(Class 2 + Class 7 LERF + Class 8 SGTR + Class 8 ISLOCA)]= 0.0023 * [1.17E-05 -(1.11E-08 + 6.90E-08 + 1.05E-06 + 2.77E-08)]= 2.43E-08/yrFor IP3:Class_3a = 0.0092 * [CDF -(Class 2 + Class 7 LERF + Class 8 SGTR + Class 8 ISLOCA)]= 0.0092 * [1.48E-05 -(3.99E-09 + 7.14E-08 + 9.77E-07 + 1.93E-07)]= 1.25E-07/yrClass_3b = 0.0023 * [CDF -(Class 2 + Class 7 LERF + Class 8 SGTR + Class 8 ISLOCA)]= 0.0023 * [1.48E-05 -(3.99E-09 + 7.14E-08 + 9.77E-07 + 1.93E-07)]= 3.13E-08/yrFor this analysis, the associated containment leakage for Class 3a is 1OLa and 10OLa for Class3b, which is consistent with the latest EPRI methodology [3] and the NRC SE [7].Class 4 SequencesThis group represents containment isolation failure-to-seal of Type B test components.Because these failures are detected by Type B tests which are unaffected by the Type A ILRT,this group is not evaluated any further in this analysis.Class 5 SequencesThis group represents containment isolation failure-to-seal of Type C test components.Because these failures are detected by Type C tests which are unaffected by the Type A ILRT,this group is not evaluated any further in this analysis.Class 6 SequencesThis group is similar to Class 2. These are sequences that involve core damage with a failure-to-seal containment leakage due to failure to isolate the containment. These sequences aredominated by misalignment of containment isolation valves following a test/maintenanceevolution. Consistent with the EPRI guidance, this accident class is not explicitly consideredsince it has a negligible impact on the results.P0247130002-47225-4 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsClass 7 SequencesThis group represents containment failure induced by early and late severe accidentphenomena. From Table 4.2-6a for IP2, the frequency for early Class 7 sequences is4.66E-09/yr + 6.90E-08/yr = 7.37E-08/yr, and the frequency for the late Class 7 sequences is2.71E-06/yr. From Table 4.2-6b for IP3, the frequency for early Class 7 sequences is1.17E-07/yr + 7.14E-08/yr = 1.88E-07/yr, and the frequency for the late Class 7 sequences is2.17E-06/yr.Class 8 SeauencesThis group represents sequences where containment bypass occurs (SGTR or ISLOCA). Fromthe frequency information provided in Table 4.2-6a for IP2, the total SGTR contribution to coredamage is 1.05E-06/yr and the ISLOCA contribution to core damage is 2.77E-08/yr. From thefrequency information provided in Table 4.2-6b for IP3, the total SGTR contribution to coredamage is 9.77E-07/yr and the ISLOCA contribution to core damage is 1.93E-07/yr.Summary of Accident Class FrequenciesIn summary, the accident sequence frequencies that can lead to release of radionuclides to thepublic have been derived in a manner consistent with the definition of accident classes definedin EPRI 1018243 [3] and are shown in Table 5.1-1 for IP2 and for IP3.P0247130002-47225-5 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.1-1RADIONUCLIDE RELEASE FREQUENCIES AS A FUNCTION OFACCIDENT CLASS (IP2 AND IP3 BASE CASE)ACCIDENT DESCRIPTION IP2 IP3CLASS FREQUENCY FREQUENCY(CONTAINMENT (1/YR) (1/YR)RELEASE TYPE)1 Containment Intact 7.74E-06 1.11E-052 Large Isolation Failures (Failure to Close) 1.11E-08 3.99E-093a Small Isolation Failures (liner breach) 9.73E-08 1.25E-073b Large Isolation Failures (liner breach) 2.43E-08 3.13E-084 Small Isolation Failures (Failure to seal -N/A N/AType B)5 Small Isolation Failures (Failure to seal- N/A N/AType C)6 Other Isolation Failures (e.g., dependent N/A N/Afailures)7-CFE Failures Induced by Phenomena (Early) 7.37E-08 1.88E-077-CFL Failures Induced by Phenomena (Late) 2.71E-06 2.17E-068-SGTR Containment Bypass (Steam Generator 1.05E-06 9.77E-07Tube Rupture)8-ISLOCA Containment Bypass (Interfacing System 2.77E-08 1.93E-07LOCA)CDF All CET End States (Including Intact 1.17E-05 1.48E-05Case)5.2 STEP 2 -REACTOR YEARDEVELOP PLANT-SPECIFIC PERSON-REM DOSE (POPULATION DOSE) PERPlant-specific release analyses were performed to estimate the weighted average person-remdoses to the population within a 50-mile radius from the plant. The releases are based on acombination of the information provided by the IP2 and IP3 SAMA re-analysis [10], additionalpopulation dose runs for the intact containment scenarios [11], and the Level 2 containmentfailure release frequencies [18, 19] (see Tables 4.2-6a and 4.2-6b of this analysis). Theresults of applying these releases to the EPRI containment failure classifications areP0247130002-47225-6 Risk Impact Assessment of Extending the Indian Point ILRT Intervalssummarized below. Note that the 7-CFE release category is further refined to be the weightedaverage of the two contributors for moving forward in the ILRT methodology since it is notimpacted by the change to the ILRT interval.For IP2:Class 1Class 2Class 3aClass 3bClass 4Class 5Class 6Class 7-CFEClass 7-CFLClass 8-SGTR= 4.41E+04 person-rem (at 1.OLa)= 6.51E+07 person-rem= 4.41E+04 person-rem x 1OLa = 4.41E+05 person-rem= 4.41E+04 person-rem x 10OLa = 4.41E+06 person-rem= Not analyzed= Not analyzed= Not analyzed= (4.66E-09

• 1.94E+07 + 6.90E-08

• 6.51E+07) /(4.66E-09 + 6.90E-08) = 6.22E+07 person-rem= 6.87E+06 person-rem= 6.51E+07 person-remClass 8-ISLOCA = 6.51E+07 person-remFor IP3:Class 1Class 2Class 3aClass 3bClass 4Class 5Class 6Class 7-CFEClass 7-CFLClass 8-SGTR= 4.41E+04 person-rem (at 1.OLa)= 5.08E+07 person-rem= 4.41E+04 person-rem x 1OLa = 4.41E+05 person-rem= 4.41E+04 person-rem x 10OLa = 4.41E+06 person-rem= Not analyzed= Not analyzed= Not analyzed= (1.17E-07

• 2.OOE+07 + 7.14E-08

• 5.08E+07) /(1.17E-07 + 7.14E-08) = 3.17E+07 person-rem= 6.85E+06 person-rem= 5.08E+07 person-remClass 8-ISLOCA = 5.08E+07 person-remP0247130002-47225-7 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsIn summary, the population dose estimates derived for use in the risk evaluation per the EPRImethodology [3] for all EPRI classes are provided in Table 5.2-1, which includes the valuespreviously presented in Table 4.2-6a and 4.2-6b as well as the Class 3a, 3b, and 7-CFEpopulation doses calculated above.TABLE 5.2-1IP2 AND IP3 POPULATION DOSEFOR POPULATION WITHIN 50 MILESACCIDENT DESCRIPTION IP2 IP3CLASS PERSON- PERSON-(CONTAINMENT REM REMRELEASE TYPE) (0-50 (0-50MILES) MILES)1 Containment Intact 4.41E+04 4.41E+042 Large Isolation Failures (Failure to 6.51E+07 5.08E+07Close)3a Small Isolation Failures (liner breach) 4.41E+05 4.41E+053b Large Isolation Failures (liner breach) 4.41E+06 4.41E+064 Small Isolation Failures (Failure to seal -N/A N/AType B)5 Small Isolation Failures (Failure to seal -N/A N/AType C)6 Other Isolation Failures (e.g., dependent N/A N/Afailures)7-CFE Failures Induced by Phenomena (Early) 6.22E+07 3.17E+077-CFL Failures Induced by Phenomena (Late) 6.87E+06 6.85E+068-SGTR Containment Bypass (Steam Generator 6.51E+07 5.08E+07Tube Rupture)8-ISLOCA Containment Bypass (Interfacing 6.51E+07 5.08E+07System LOCA)The above population doses, when multiplied by the frequency results presented in Table5.1-1, yield the IP2 and IP3 baseline mean dose risk for each EPRI accident class. Theseresults are presented in Table 5.2-2a for IP2 and in Table 5.2-2b for IP3. Note that theadditional contribution to EPRI Class 3b from the corrosion analysis as described in Section 4.4is also included in these tables.P0247130002-47225-8 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.2-2AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON-REM EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES (0-50 PLUS CORROSION CORROSION(CONTAINMENT MILES) (PERSON-RELEASE TYPE) FREQUENCY PERSON- FREQUENCY PERSON- REM/YR) (1)(1/YR) REM/YR (1/YR) REM/YR(0-50 MILES) (0-50MILES)1 Containment 4.41E+04 7.74E-06 3.41E-01 7.74E-06 3.41E-01 -4.14E-06Intact (2)2 Large Isolation 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 9.73E-08 4.29E-02 9.73E-08 4.29E-02 --Failures (linerbreach)3b Large Isolation 4.41E+06 2.43E-08 1.07E-01 2.44E-08 1.08E-01 4.14E-4Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-9 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.2-2AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON-REM EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES (0-50 PLUS CORROSION CORROSION(CONTAINMENT MILES) (PERSON-RELEASE TYPE) FREQUENCY PERSON- FREQUENCY PERSON- REM/YR) (1)(1/YR) REM/YR (1/YR) REM/YR(0-50 MILES) (0-50MILES)7-CFE Failures Induced 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00 --by Phenomena(Early)7-CFL Failures Induced 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01 --by Phenomena(Late)8-SGTR Containment 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01 --Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00 --Bypass(Interfacing_System LOCA) ICDF All CET end 1.17E-05 9.426E+01 1.17E-05 9.426E+01 4.10E-4states) Only release Classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as ILa release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-10 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.2-2BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (PERSON-RELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- RERSO(1(1/YR) REM/YR (1/YR) REM/YR REM/YR)(0-50 MILES) (0-50MILES)1 Containment 4.41E+04 1.11E-05 4.91E-01 1.11E-05 4.91E-01 -5.32E-6Intact (2)2 Large Isolation 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 1.25E-07 5.51E-02 1.25E-07 5.51E-02 --Failures (linerbreach)3b Large Isolation 4.41E+06 3.13E-08 1.38E-01 3.14E-08 1.38E-01 5.32E-4Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-11 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.2-2BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (PERSON-RELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- REM/YR)(1/YR) REM/YR (1I/YR) REM/YR(0-50 MILES) (0-50MILES)7-CFE Failures Induced 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00 --by Phenomena(Early)7-CFL Failures Induced 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01 --by Phenomena(Late)8-SGTR Containment 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01 --Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00 --Bypass(Interfacing_System LOCA) ICDF All CET end 1.48E-05 8.114E+01 1.48E-05 8.115E+01 5.27E-4states(1) Only release Classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-12 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe baseline IP2 and IP3 doses compare reasonably with other plants given the relativepopulation densities surrounding each location:PLANT ANNUAL DOSE REFERENCE(PERSON-REM/YR)Indian Point 2 94.3 [Table 5.2-2a]Indian Point 3 81.1 [Table 5.2-2b]Peach Bottom 2 8.6 [22]Farley Unit 1, 2 1.5, 2.4 [23]Crystal River 1.4 [24]5.3 STEP 3 -EVALUATE RISK IMPACT OF EXTENDING TYPE A TEST INTERVAL FROM 10-TO-15 YEARSThe next step is to evaluate the risk impact of extending the test interval from its current ten-year value to fifteen-years. To do this, an evaluation must first be made of the risk associatedwith the ten-year interval since the base case applies to a 3-year interval (i.e., a simplifiedrepresentation of a 3-in- 10 year interval).Risk Impact Due to 10-year Test IntervalAs previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, therelease magnitude is not impacted by the change in test interval (a small or large breachremains the same, even though the probability of not detecting the breach increases). Thus,only the frequency of Class 3a and 3b sequences is impacted. The risk contribution is changedbased on the EPRI guidance as described in Section 4.3 by a factor of 3.33 compared to thebase case values. The results of the calculation for a 10-year interval are presented in Table5.3-1a for IP2 and in Table 5.3-1b for IP3.Risk Imoact Due to 15-Year Test IntervalThe risk contribution for a 15-year interval is calculated in a manner similar to the 10-yearinterval. The difference is in the increase in probability of not detecting a leak in Classes 3aand 3b. For this case, the value used in the analysis is a factor of 5.0 compared to the 3-yearinterval value, as described in Section 4.3. The results for this calculation are presented inTable 5.3-2a for IP2 and in Table 5.3-2b for IP3.P0247130002-47225-13 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-1AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-RELEASE TYPE) MILES) (1/YR) REM/YR (1/YR) REM/YR REM/YR) (1)(0-50 (0-50MILES) MILES)1 Containment 4.41E+04 7.46E-06 3.29E-01 7.45E-06 3.29E-01 -2.38E-05Intact (2)2 Large Isolation 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 3.24E-07 1.43E-01 3.24E-07 1.43E-01 --Failures (linerbreach)3b Large Isolation 4.41E+06 8.1OE-08 3.57E-01 8.15E-08 3.60E-01 2.38E-3Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-14 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-1AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 FREQUENCY PERSON FREQUENCY PERSON (PERSON-RELEASE TYPE) MILES) (1/YR) REM/YR (1/YR) REM/YR REM/YR)(1)(0-S0 (0-50MILES) MILES)7-CFE Failures Induced 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00by Phenomena(Early)7-CFL Failures Induced 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01by Phenomena(Late)8-SGTR Containment 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00Bypass(InterfacingSystem LOCA) ICDF All CET end 1.17E-05 9.460E+01 1.17E-05 9.460E+01 2.35E-3states(1) Only release classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 1L. release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3aand 3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-15 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-1BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 FREQUENCY PERSON FREQUENCY PERSON (PERSON-RELEASE TYPE) MILES) (1/YR) REM/YR (1/YR) REM/YR REM/YR)(0-50 (0-50MILES) MILES)1 Containment 4.41E+04 1.08E-05 4.75E-01 1.08E-05 4.75E-01 -3.05E-5Intact (2)2 Large Isolation 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 4.16E-07 1.84E-01 4.16E-07 1.84E-01 --Failures (linerbreach)3b Large Isolation 4.41E+06 1.04E-07 4.59E-01 1.05E-07 4.62E-01 3.05E-3Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-16 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-1BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 FREQUENCY PERSON FREQUENCY PERSON (PERSON-RELEASE TYPE) MILES) (1I/YR) REM/YR (1/YR) REM/YR REM/YR)(0-50 (0-50MILES) MILES)7-CFE Failures Induced 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00by Phenomena(Early)7-CFL Failures Induced 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01by Phenomena(Late)8-SGTR Containment 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00Bypass(InterfacingSystem LOCA) ICDF All CET end 1.48E-05 8.158E+01 1.48E-05 8.158E+01 3.02E-3states(1) Only release classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 11 release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3aand 3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-17 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-2AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSIONS(CONTAINMENT (0-50 (ESNRELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-(1/YR) REM/YR (1/YR) REM/YR REM/YR)(0-50 (0-50MILES) MILES)1 Containment Intact 4.41E+04 7.25E-06 3.20E-01 7.25E-06 3.20E-01 -5.51E-05(2)2 Large Isolation 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 4.86E-07 2.15E-01 4.86E-07 2.15E-01 --Failures (linerbreach)3b Large Isolation 4.41E+06 1.22E-07 5.36E-01 1.23E-07 5.42E-01 5.51E-3Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-18 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-2AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (ESNRELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-(1/YR) REM/YR (1/YR) REM/YR REM/YR)(1)(0-50 (0-50MILES) MILES)7-CFE Failures Induced 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00 --by Phenomena(Early)7-CFL Failures Induced 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01 --by Phenomena(Late)8-SGTR Containment 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01 --Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00 --Bypass(InterfacingSystem LOCA)CDF All CET end 1.17E-05 9.484E+01 1.17E-05 9.484E+01 5.46E-3states(1) Only release classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in asmall reduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3aand 3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-19 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-2BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (ESNRELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-(1/YR) REM/YR (1/YR) REM/YR REM/YR)(1)(0-50 (0-50MILES) MILES)1 Containment 4.41E+04 1.05E-05 4.64E-01 1.05E-05 4.64E-01 -7.08E-5Intact (2)2 Large Isolation 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 6.25E-07 2.76E-01 6.25E-07 2.76E-01 --Failures (linerbreach)3b Large Isolation 4.41E+06 1.56E-07 6.89E-01 1.58E-07 6.96E-01 7.08E-3Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependentfailures)P0247130002-47225-20 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-2BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (ESNRELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-(1/YR) REM/YR (1/YR) REM/YR REM/YR)(1)(0-S0 (0-50MILES) MILES)7-CFE Failures Induced 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00 --by Phenomena(Early)7-CFL Failures Induced 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01 --by Phenomena(Late)8-SGTR Containment 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01 --Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00 --Bypass(InterfacingSystem LOCA)CDF All CET end 1.48E-05 8.189E+01 1.48E-05 8.190E+01 7.01E-3statesIII(1) Only release classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-21 Risk Impact Assessment of Extending the Indian Point ILRT Intervals5.4 STEP 4 -DETERMINE THE CHANGE IN RISK IN TERMS OF LARGE EARLY RELEASEFREQUENCYRegulatory Guide 1.174 provides guidance for determining the risk impact of plant-specificchanges to the licensing basis. RG 1.174 defines very small changes in risk as resulting inincreases of core damage frequency (CDF) below 1E-06/yr and increases in LERF below1E-07/yr, and small changes in LERF as below 1E-06/yr. Because the ILRT does not impactCDF for IP2 and IP3, the relevant metric is LERF.For IP2 and IP3, 100% of the frequency of Class 3b sequences can be used as a conservativefirst-order estimate to approximate the potential increase in LERF from the ILRT intervalextension (consistent with the EPRI guidance methodology and the NRC SE). Based on theoriginal 3-in-10 year test interval assessment from Tables 5.2-2a and 5.2-2b, the Class 3bfrequency is 2.44E-08/yr for IP2 and 3.14E-08/yr for IP3, which includes the corrosion effect ofthe containment liner. Based on a ten-year test interval from Tables 5.3-1a and 5.3-1b, theClass 3b frequency is 8.15E-08/yr for IP2 and 1.05E-07/yr for IP3; and, based on a fifteen-year test interval from Tables 5.3-2a and 5.3-2b, it is 1.23E-07/yr for IP2 and 1.58E-07/yr forIP3. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is dueto increasing the ILRT test interval from 3 to 15 years (including corrosion effects) is 9.84E-08/yr for IP2 and 1.26E-07/yr for IP3. Similarly, the increase in LERF due to increasing theinterval from 10 to 15 years (including corrosion effects) is 4.13E-08/yr for IP2 and 5.31E-08/yr for IP3. As can be seen, even with the conservatisms included in the evaluation (per theEPRI methodology), the estimated change in LERF is well within Region II of Figure 4 ofReference [4] (i.e., the acceptance criteria for small changes in LERF) when comparing the 15year results to the original 3-in-10 year requirement.5.5 STEP 5 -DETERMINE THE IMPACT ON THE CONDITIONAL CONTAINMENT FAILUREPROBABILITYAnother parameter that can provide input into the decision-making process is the change inthe conditional containment failure probability (CCFP). The change in CCFP is indicative of theeffect of the ILRT on all radionuclide releases, not just LERF. The CCFP can be calculated fromthe results of this analysis. One of the difficult aspects of this calculation is providing adefinition of the "failed containment." In this assessment, the CCFP is defined such thatcontainment failure includes all radionuclide release end states other than the intact state and,consistent with the EPRI guidance, the small isolation failures (Class 3a). The conditional partof the definition is conditional given a severe accident (i.e., core damage).P0247130002-47225-22 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe change in CCFP can be calculated by using the method specified in the EPRI methodology[3]. The NRC SE has noted a change in CCFP of <1.5% as the acceptance criterion to be usedas the basis for showing that the proposed change is consistent with the defense-in-depthphilosophy. Table 5.5-1 shows the CCFP values that result from the assessment for thevarious testing intervals including corrosion effects in which the flaw rate is assumed to doubleevery five years.TABLE 5.5-1IP2 AND IP3 ILRT CONDITIONAL CONTAINMENT FAILURE PROBABILITIESUNIT CCFP CCFP CCFP3 IN 10 1 IN 10 1 IN 15 ACCFP15-3 ACCFP15-10YRS YRS YRSIndian Point 2 33.19% 33.67% 34.03% 0.84% 0.35%Indian Point 3 24.03% 24.52% 24.88% 0.85% 0.36%CCFP = [1 -(Class 1 frequency + Class 3a frequency)/CDF] x 100%The change in CCFP of less than 1% as a result of extending the test interval to 15 years fromthe original 3-in-10 year requirement is judged to be relatively insignificant, and is less thanthe NRC SE acceptance criteria of <. 1.5%.5.6 SUMMARY OF INTERNAL EVENTS RESULTSTable 5.6-1a summarizes the internal events results of this ILRT extension risk assessment forIP2. Table 5.6-1b summarizes the internal events results of this ILRT extension riskassessment for IP3. The results between the 3-in-10 year interval and the 15 year intervalcompared to the acceptance criteria are then shown in Table 5.6-2 for IP2 and IP3, and it isdemonstrated that the acceptance criteria are met.P0247130002-47225-23 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.6-1AIP2 ILRT CASES:BASE, 3 TO 10, AND 3 TO 15 YR EXTENSIONS(INCLUDING AGE ADJUSTED STEEL LINER CORROSION LIKELIHOOD)EPRI DOSE BASE CASE EXTEND TO EXTEND TOCLASS PER-REM 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARSCDF PERSON- CDF PERSON- CDF PERSON-(1/YR) REM/YR (1/YR) REM/YR (1/YR) REM/YR1 4.41E+04 7.74E-06 3.41E-01 7.45E-06 3.29E-01 7.25E-06 3.20E-012 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 1.11E-08 7.23E-013a 4.41E+05 9.73E-08 4.29E-02 3.24E-07 1.43E-01 4.86E-07 2.15E-013b 4.41E+06 2.44E-08 1.08E-01 8.15E-08 3.60E-01 1.23E-07 5.42E-017-CFE 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00 7.37E-08 4.58E+007-CFL 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01 2.71E-06 1.86E+018-SGTR 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01 1.05E-06 6.80E+018-ISLOCA 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00 2.77E-08 1.80E+00Total [_1.17E-05 9.426E+01 1. 17E-0-9 19.4 .484E+01ILRT Dose Rate 1.51E-01 5.02E-01 7.56E-01(person-rem/yr) from3a and 3bDelta From 3 yr --- 3.39E-01 5.84E-01TotalIDose From 10 yr 2.45E-01DoseRate*1)3b Frequency (LERF) 2.44E-08 8.15E-08 1.23E-07Delta 3b From 3 yr --- 5.71E-08 9.84E-08LERF From 10 yr ......_4.13E-08CCFP % 33.19% 33.67% 34.03%Delta From 3 yr --- 0.49% 0.84%CCFP %From 10 yr ...0.35%( The overall difference in total dose rate is less than the difference of only the 3a and 3bcategories between two testing intervals. This is due to the fact that the Class 1 person-rem/yr decreases when extending the ILRT frequency.P0247130002-47225-24 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.6-1BIP3 ILRT CASES:BASE, 3 TO 10, AND 3 TO 15 YR EXTENSIONS(INCLUDING AGE ADJUSTED STEEL LINER CORROSION LIKELIHOOD)EPRI DOSE BASE CASE EXTEND TO EXTEND TOCLASS PER-REM 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARSCDF PERSON- CDF PERSON- CDF PERSON-(1/YR) REM/YR (1/YR) REM/YR (1/YR) REM/YR1 4.41E+04 1.11E-05 4.91E-01 1.08E-05 4.75E-01 1.05E-05 4.64E-012 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 3.99E-09 2.03E-013a 4.41E+05 1.25E-07 5.51E-02 4.16E-07 1.84E-01 6.25E-07 2.76E-013b 4.41E+06 3.14E-08 1.38E-01 1.05E-07 4.62E-01 1.58E-07 6.96E-017-CFE 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00 1.88E-07 5.97E+007-CFL 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01 2.17E-06 1.49E+018-SGTR 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01 9.77E-07 4.96E+018-ISLOCA 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00 1.93E-07 9.80E+00Total 1.48E-05 8.115E+01 I 1.48E-05 18.158E+011 1.48E-05 18.190E+01ILRT Dose Rate 1.93E-01 6.46E-01 9.72E-01(person-rem/yr) from3a and 3bDelta From 3 yr --- 4.36E-01 7.51E-01TotalDose From 10 yr --- 3.15E-01DoseRate(l)3b Frequency (LERF) 3.14E-08 1.05E-07 1.58E-07Delta 3b From 3 yr --- 7.34E-08 1.26E-07LERFtFrom 10 yr ...... 5.31E-08CCFP % 24.03% 24.52% 24.88%Delta From 3 yr --- 0.49% 0.85%CCFP %From 10 yr --- 0.36%(1) The overall difference in total dose rate is less than the difference of only the 3a and 3bcategories between two testing intervals. This is due to the fact that the Class 1 person-rem/yr decreases when extending the ILRT frequency.P0247130002-47225-25 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.6-2IP2 AND IP3 ILRT EXTENSION COMPARISON TO ACCEPTANCE CRITERIAUnit ALERF APerson-rem/yr ACCFPIndian Point 2 9.84E-8/yr 0.584/yr (0.62%) 0.84%Indian Point 3 1.26E-7/yr 0.751/yr (0.93%) 0.85%Acceptance < 1.OE-6/yr <1.0 person- <1.50/oCriteria rem/yr or <1.0%5.7 EXTERNAL EVENTS CONTRIBUTIONSince the risk acceptance guidelines in RG 1.174 are intended for comparison with a full-scopeassessment of risk, including internal and external events, a bounding analysis of the potentialimpact from external events is presented here.The method chosen to account for external events contributions is similar to that used in theSAMA analysis [20] in which a multiplier was applied to the internal events results based oninformation from the IPEEE [8, 9]. Similar to that provided in the SAMA analysis, a descriptionof the external events contribution to risk at IP2 and IP3 is provided below.5.7.1 Indian Point 2 External Events DiscussionThe IP2 Individual Plant Examination of External Events (IPEEE) included quantitative CDFresults for high winds, seismic, and fire contributors. Each of these is discussed below.A high wind analysis was performed for the IP2 IPEEE. Conservative assumptions in the highwind PRA analysis included the following.* Offsite power was assumed to be lost for all high wind events." Building frame failures were assumed to cause failure of all equipment within thebuilding.* Missile (high wind projectile) impact on a structure was assumed to cause failure ofall equipment within that structure.* Likelihood of missile (high wind projectile) strikes was assumed to be independent ofthe intensity of the hazard.* Both onsite and offsite alternate power sources (gas turbines) were assumed to failgiven failure of a more robust structure.P0247130002-47225-26 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe core damage frequency contribution associated with high wind events was estimated to be3.03E-05/yr. As described above, this is a conservative value. In addition, plant changes,improved equipment performance data, and modeling improvements since the issuance of theIP2 IPEEE have demonstrated that the response of plant systems as modeled at that time wasconservative. This can be seen from the reduction in internal events CDF from 2.85E-05/yr atthe time the IPEEE was developed to the present value of 1.17E-05/yr. Although conservative,consistent with the SAMA analysis, the wind risk contribution of 3.03E-05/yr is maintained todetermine the potential external events impact in the ILRT extension assessment.A seismic PRA analysis was performed for the seismic portion of the IP2 IPEEE. The seismicPRA analysis was a conservative analysis. Therefore, its results should not be compareddirectly with the best-estimate internal events results. Conservative assumptions in the seismicPRA analysis included the following.* Sequences in the seismic PRA involving loss of off-site power were assumed to beunrecoverable. If off-site power was recovered following a seismic event, there wouldbe many more systems available to maintain core cooling and containment integritythan were credited for those sequences.* A single, conservative, surrogate element whose failure leads directly to coredamage was used in the seismic risk quantification to model the most seismicallyrugged components.* Seismic-induced ATWS was considered in the analysis, but no credit was included formanual scram or mitigation of ATWS using the boration system. This conservativelyresulted in most seismic-induced ATWS events leading to consequential coredamage.* Redundant components were conservatively assumed to be completely correlated bytreating them as if they were one component for the purpose of determining theprobability of seismic induced failures." Several systems were assumed to be unavailable during a seismic event, including:a. the city water system, which can be used to supply backup cooling to thecharging pumps if CCW is lost, as an alternate source of suction to the AFWpumps and to provide alternate cooling to the RHR and SI pumps;b. the primary water system, which can also be used as a backup to CCW tosupply cooling to the RHR and SI pumps; andc. the onsite and offsite gas turbine generators, which can provide alternatestation power.* No credit was taken for recovery of power through the alternate safe shutdownsystem (ASSS).The seismic CDF in the IPEEE was originally estimated to be 1.46E-05/yr. As a result of anIPEEE recommendation, the CCW surge tank hold-down bolts were upgraded, reducing theseismic CDF to 1.06E-05/yr. Although it remains conservative, consistent with the SAMAP0247130002-47225-27 Risk Impact Assessment of Extending the Indian Point ILRT Intervalsanalysis, the seismic risk contribution of 1.06E-05/yr is maintained to determine the potentialexternal events impact in the ILRT extension assessment.The conservative EPRI FIVE methodology was used for initial screening of fire zones in the IP2IPEEE fire analysis. Unscreened fire zones were then analyzed in more detail using a fire PRAapproach. The sum of the resulting fire zone CDF values is approximately 1.84E-05/yr.Conservative assumptions in the IP2 IPEEE fire analysis include the following." The frequency and severity of fires were generally conservatively overestimated inthe generic IPEEE fire analysis methods. A revised NRC fire events databaseindicates a trend toward lower frequency and less severe fires. This trend reflectsimproved housekeeping, reduction in transient fire hazards, and other improved fireprotection steps at utilities.* Cable failure due to fire damage was assumed to arise from open circuits, hot shortcircuits, and short circuits to ground. In damaging a cable, the analysis addressedthe ability of the fire to induce the conductor failure mode of concern. Hot shortswere conservatively assigned a probability of 0.1, which was applied to all singlephase, AC control circuit or DC power and control circuit cases regardless of whetherthe wires were in the same multi-conductor." A plant trip was assumed for all fires, including those for which immediate operatoractions are not specified in emergency response procedures." PORV block valves were assumed to be in the more limiting position (open or closed)to maximize the impact of the fire.* The main feedwater and condensate systems were assumed to be unavailable in allscenarios, even when their power source was not impacted by the fire scenario. Useof these systems for recovery, following a failure of AFW, is addressed in currentplant procedures.* All sequences involving induced RCP seal LOCAs were assumed to lead to completeseal failure. Although casualty cables exist for powering ECCS pumps from the ASSSpower source, the ASSS was assumed to be ineffective in mitigating induced LOCAs.* The currently accepted RCP seal LOCA methodology is more detailed and providessequences with varying leakage rates. Under that current methodology, a majority ofseal LOCAs remain within the capability of a charging pump (which has hardwiredASSS transfer capability) to provide makeup.As noted previously, plant changes, improved equipment performance data and modelingimprovements since the issuance of the IP2 IPEEE have demonstrated that the response ofplant systems as modeled at that time was conservative. This can be seen from the reductionin internal events CDF from 2.85E-05/yr at the time the IPEEE was developed to the presentvalue of 1.17E-5/yr., a reduction factor of 2.4. Factoring in the additional conservatisms in thefire analysis noted above, an overall reduction factor of 2 is reasonable which is consistent withthe assumption used in the SAMA analysis [20]. The IPEEE fire CDF value, reduced by a factorof two, is 9.20E-06/yr.P0247130002-47225-28 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe IP2 Individual Plant Examination of External Events (IPEEE) concluded for "Other" externalevents, with the exception of "high wind" events as noted above, that no undue risks arepresent that might contribute to CDF with a predicted frequency in excess of 1.OE-06/yr. Asthese events are not dominant contributors to external event risk and quantitative analysis ofthese events is not practical, they are considered negligible in estimation of the external eventsimpact on the ILRT extension assessment.In summary, the combination of the IPEEE high wind CDF and the reduced seismic and fireCDF values described above results in an external events risk estimate of 5.01E-05/yr, which is4.3 times higher than the internal events CDF (1.17E-05/yr).5.7.2 Indian Point 3 External Events DiscussionThe IP3 Individual Plant Examination of External Events (IPEEE) concluded for high winds,floods, and "Other" external events that no undue risks are present that might contribute toCDF with a predicted frequency in excess of 1.OE-06/yr. Note that at IP3 (compared to IP2),the EDGs are in separate concrete bunkered cells and as such are not susceptible to highwinds. In any event, as these other events are not dominant contributors to external eventrisk and quantitative analysis of these events is not practical, they are considered negligible inestimation of the external events impact on the ILRT extension assessment. The IPEEEanalyses using the seismic PRA and fire PRA provided quantitative, but conservative, results.Therefore, the results were combined as described below to represent the total external eventsrisk.A seismic PRA analysis was performed for the seismic portion of the IP3 IPEEE. The seismicPRA analysis is a conservative analysis. Therefore, its results should not be compared directlywith the best-estimate internal events results. Conservative assumptions in the seismic PRAanalysis included the following." Each of the sequences in the seismic PRA assumes unrecoverable loss of off-sitepower. If off-site power was maintained, or recovered, following a seismic event,there would be many more systems available to maintain core cooling andcontainment integrity than were credited in the analysis.* Seismic events were assumed to induce a small loss of coolant accident (LOCA) inaddition to a loss of offsite power." A single, conservative, surrogate element whose failure leads directly to coredamage was used in the seismic risk quantification to model the most seismicallyrugged components." Redundant components were conservatively assumed to be completely correlated bytreating them as if they were one component for the purpose of determining theprobability of seismic induced failures.P0247130002-47225-29 Risk Impact Assessment of Extending the Indian Point ILRT Intervals* The ATWS event tree was conservatively simplified so that all conditions which leadto a failure to trip result in core damage, without the benefit of emergency borationor other mitigating systems." Because there is little industry experience with crew actions following seismic events,human actions were conservatively characterized.The seismic CDF in the IPEEE was conservatively estimated to be 4.40E-05/yr. As describedabove, this is a conservative value. The seismic PRA CDF has been re-evaluated to reflectupdated random component failure probabilities and to model recovery of onsite power andlocal operation of the turbine-driven AFW pump. The updated seismic CDF is 2.65E-05/yr.Although it remains conservative, consistent with the SAMA analysis, the seismic riskcontribution of 2.65E-05/yr is maintained to determine the external events impact on the ILRTextension assessment.The EPRI Fire PRA Implementation Guide was followed for the IP3 IPEEE fire analysis. The EPRIFire Induced Vulnerability Evaluation (FIVE) method was used for the initial screening, fortreatment of transient combustibles, and as the source of fire frequency data. The sum of theresulting fire zone CDF values is approximately 5.58E-05/yr. Conservatisms in the IP3 IPEEEfire analysis include the following.* The frequency and severity of fires were generally conservatively overestimated. Arevised NRC fire events database indicates a trend toward lower frequency and lesssevere fires. This trend reflects improved housekeeping, reduction in transient firehazards, and other improved fire protection steps at utilities." There is little industry experience with crew actions following fires. This led toconservative characterization of crew actions in the IPEEE fire analysis. Because CDFis strongly correlated with crew actions, this conservatism has a profound effect onfire results.* Hot gas layer temperature timing calculations were based on simplified analyses(versus more detailed calculations such as GOTHIC or even COMPBURN) which arebelieved to result in more severe timing (i.e., shorter time to equipment failure).* Heat and combustion products from a fire within a zone were assumed to beconfined within the zone. Heat loss through separating zones was not considered;nor was heat loss through open equipment hatches, ladder ways, open doorways, orunsealed penetrations." Cable failure due to fire damage was assumed to arise from open circuits, hot shortscircuits, and short circuits to ground. In damaging a cable, the fire was alwaysassumed to induce the conductor failure mode of concern." A plant trip was assumed for all fires, including those for which immediate operatoractions are not specified in emergency response procedures." For several fire zones, a minimum heat requirement for target damage wasestimated." Propagation of fires in cable spreading room trays and electrical tunnels was modeledusing a maximum heat release rate. This results in a shorter time to damage thanP0247130002-47225-30 Risk Impact Assessment of Extending the Indian Point ILRT Intervalsthe five-minute delay using heat release rate scaling factors as a function of distancerecommended in the EPRI fire PRA implementation guide.Implementation of the IP3 IPEEE recommendations reduced the fire risk. The fire suppressionsystem in the 480V switchgear room was restored to automatic actuation, and realignmentand rerouting of the power feeds to the EDG exhaust fans and engine auxiliaries in emergencydiesel generator room 31, emergency diesel generator room 32, and emergency dieselgenerator room 33 significantly reduce the respective fire zone's CDF. In addition, restorationof the 480V switchgear room fire suppression system to automatic actuation results in a similarreduction in the fire zone 14/37A multiple compartment fire CDF. Consequently, the IPEEE fireCDF value was reduced from 5.58E-05/yr to 2.55E-05/yr. Although it remains conservative,consistent with the SAMA analysis, the fire risk contribution of 2.55E-05/yr is maintained todetermine the potential external event impact on the ILRT extension assessment.In summary, combining the reduced seismic and fire CDF values results in an external eventsrisk estimate of 5.20E-05/yr, which is 3.5 times higher than the internal events CDF (1.48E-05/yr).5.7.3 Additional Seismic Risk DiscussionAs an additional consideration, it can be noted that in June 2013, Entergy submittedinformation to the NRC that addressed some conservatisms in the original IPEEE analyses, andindicated that the seismic CDF risk at IP2 and IP3 are both actually less than 1.OE-05/yr [25].However, to maintain consistency with the approach utilized in the SAMA analysis, theadditional information will not be factored into this analysis but is noted here for completeness.5.7.4 External Events Impact SummaryTable 5.7-1 summarizes the external events CDF contribution for IP2 and 1P3. Although notedas conservative, these values are consistent with that used in the SAMA analysis [20].P0247130002-47225-31 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-1EXTERNAL EVENTS CONTRIBUTOR SUMMARY [20]EXTERNAL EVENT INITIATOR GROUP IP2 CDF (1/YR)_ 1 1P3 CDF (1/YR)Seismic 1.06E-05 2.65E-05Internal Fire 9.20E-06 2.55E-05High Winds 3.03E-05 ScreenedOther Hazards Screened ScreenedTotal (for initiators with CDF available) 5.01E-05 5.20E-05Internal Events CDF 1.17E-05 1.48E-05External Events Multiplier 4.28 3.51From Table 5.7-1, the external events multiplier for IP2 is conservatively estimated to be 4.28and for IP3, it is conservatively estimated to be 3.51.5.7.5 External Events Impact on ILRT Extension AssessmentThe EPRI Category 3b frequency for the 3-per-10 year, 1-per-10 year, and 1-per-15 year ILRTintervals are shown in Table 5.6-1a for IP2 as 2.44E-08/yr, 8.15E-08/yr, and 1.23E-07/yr,respectively. Using an external events multiplier of 4.28 for IP2, the change in the LERF riskmeasure due to extending the ILRT from 3-per-l.0 years to 1-per-15 years, including bothinternal and external hazards risk, is estimated as shown in Table 5.7-2a. Similarly, the EPRIClass 3b frequencies shown in Table 5.6-1b for IP3 are 3.14E-08/yr, 1.05E-07/yr, and1.58E-07/yr. Using an external events multiplier of 3.51 for IP3, the change in the LERF riskmeasure due to extending the ILRT from 3-per-10 years to 1-per-15 years, including bothinternal and external hazards risk, is estimated as shown in Table 5.7-2b.P0247130002-47225-32 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-2AIP2 3B (LERF/YR) AS A FUNCTION OF ILRT FREQUENCYFOR INTERNAL AND EXTERNAL EVENTS(INCLUDING AGE ADJUSTED STEEL LINER CORROSION LIKELIHOOD)3B B 3B LERFFREQUENCY FREQUENCY FREQUENCY INCREASE"1)(3-PER-10 (1-PER-10 (1-PER-15YR ILRT) YEAR ILRT) YEAR ILRT)Internal Events 2.44E-08 8.15E-08 1.23E-07 9.84E-08ContributionExternal EventsContribution (Internal 1.05E-07 3.49E-07 5.26E-07 4.22E-07Events CDF x 4.28)Combined (Internal +1.29E-07 4.31E-07 6.49E-7 5.20E-07External)(1) Associated with the change from the baseline 3-per-10 year frequency to the proposed 1-per-15year frequency.Thus for IP2, the total increase in LERF (measured from the baseline 3-per-10 year ILRTinterval to the proposed 1-per-15 year frequency) due to the combined internal and externalevents contribution is estimated as 5.20E-07/yr, which includes the age adjusted steel linercorrosion likelihood.TABLE 5.7-2B1P3 3B (LERF/YR) AS A FUNCTION OF ILRT FREQUENCYFOR INTERNAL AND EXTERNAL EVENTS(INCLUDING AGE ADJUSTED STEEL LINER CORROSION LIKELIHOOD)3B 3B 3B LERFFREQUENCY FREQUENCY FREQUENCY INCREASE"1)(3-PER-10 (1-PER-10 (1-PER-15YR ILRT) YEAR ILRT) YEAR ILRT)Internal Events 3.14E-08 1.05E-07 1.58E-07 1.26E-07ContributionExternal EventsContribution (Internal 1.10E-07 3.67E-07 5.53E-07 4.43E-07Events CDF x 3.51)ombined (Internal + 1.41E-07 4.72E-07 7.11E-7 5.70E-07External) _Associated with the change from the baselineyear frequency.3-per-10 year frequency to the proposed 1-per-15P0247130002-47225-33 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThus for IP3, the total increase in LERF (measured from the baseline 3-per-10 year ILRTinterval to the proposed 1-per-15 year frequency) due to the combined internal and externalevents contribution is estimated as 5.70E-07/yr, which includes the age adjusted steel linercorrosion likelihood.The other acceptance criteria for the ILRT extension risk assessment can be similarly derivedusing the multiplier approach. The results between the 3-in-10 year interval and the 15 yearinterval compared to the acceptance criteria are shown in Table 5.7-3. As can be seen, theimpact from including the external events contributors would not change the conclusion of therisk assessment. That is, the acceptance criteria are all met such that the estimated riskincrease associated with permanently extending the ILRT surveillance interval to 15 years hasbeen demonstrated to be small. Note that a bounding analysis for the total LERF contributionfollows Table 5.7-3 to demonstrate that the total LERF value for IP2 and IP3 is less than1.OE-5/yr consistent with the requirements for a "Small Change" in risk of the RG 1.174acceptance guidelines.P0247130002-47225-34 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-3COMPARISON TO ACCEPTANCE CRITERIA INCLUDING EXTERNALEVENTS CONTRIBUTION FOR IP2 AND IP3Contributor ALERF APerson-rem/yr ACCFPIP2 Internal 9.84E-8/yr 0.584/yr (0.62%) 0.84%EventsIP2 External 4.22E-7/yr 2.50/yr (0.62%) 0.84%EventsIndian Point 2 5.20E-7/yr 3.09/yr (0.62%) 0.84%TotalIP3 Internal 1.26E-7/yr 0.751/yr (0.93%) 0.85%EventsIP3 External 4.43E-7/yr 2.63/yr (0.93%) 0.85%EventsIndian Point 3 5.70E-7/yr 3.38/yr (0.93%/) 0.850/0TotalAcceptance < 1.OE-6/yr <1.0 person- <1.50/0Criteria rem/yr or <1.0%The 5.20E-07/yr increase in LERF for IP2 and the 5.70E-07/yr increase in LERF for IP3 due tothe combined internal and external events from extending the ILRT frequency from 3-per-10years to 1-per-15 years falls within Region II between 1.OE-7 to 1.OE-6 per reactor year("Small Change" in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when thecalculated increase in LERF due to the proposed plant change is in the "Small Change" range,the risk assessment must also reasonably show that the total LERF is less than 1.OE-5/yr.Similar bounding assumptions regarding the external event contributions that were madeabove are used for the total LERF estimate.From Table 4.2-1, the total LERF due to postulated internal event accidents is 1.16E-06/yr forIP2 and 1.25E-06/yr for IP3. Although some of the LERF contributors may not be applicable toexternal events initiators, the base LERF distribution due to external events is assumed to bethe same as the internal events contribution. The total LERF values for IP2 and IP3 are thenshown in Table 5.7-4.P0247130002-47225-35 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-4IMPACT OF 15-YR ILRT EXTENSION ON LERF FOR IP2 AND IP3LERF CONTRIBUTOR IP2 (1/YR) IP3 (1/YR)Internal Events LERF 1.16E-06 1.25E-064.97E-06 4.38E-06External Events LERF [Internal Events LERF * [Internal Events LERF *4.28] 3.51]Internal Events LERF due to 1.23E-07 1.58E-07ILRT (at 15 years) (1)External Events LERF due to 5.26E-07 5.53E-07ILRT (at 15 years) (1)Total 6.78E-06/yr 6.34E-06/yr) Including age adjusted steel liner corrosion likelihood as reported in Table 5.7-2a for IP2and Table 5.7-2b for IP3.As can be seen, the estimated upper bound LERF for IP2 is estimated as6.78E-06/yr and for IP3 it is 6.34E-06/yr. These values are both less than the RG 1.174requirement to demonstrate that the total LERF due to internal and external events is less than1.OE-5/yr.5.7.6 Alternative Approach for External Events Impact on ILRT Extension AssessmentThe approach above described in Section 5.7.5 for the external events impact is consistentwith that used in the Palisades ILRT extension risk assessment evaluation that was submittedby Entergy [26] and approved by the NRC [27]. As shown, the IP2 and IP3 results fall withinthe value in the NRC SER for a small increase in population dose, as defined by percentincrease in dose (i.e., <1.0% person-rem/yr). However, since the IP2 and IP3 results rely onthat criterion rather than the absolute increase in dose criteria (i.e., < 1.0 person-rem/yr),additional information is provided to further demonstrate that the percent increase in dosecriteria is not exceeded.To do this, a reasonable estimate for the base case dose risk associated with external eventsmust be determined. In this case, each EPRI accident class is re-examined considering thepotential contribution for external events. Since the Class 1 frequency is determined based onremaining contribution not assigned to other classes, the discussion appears in reverse orderstarting with EPRI Class 8 and ending with EPRI Class 1. However, EPRI Class 2 is discussedprior to Class 3 since its value is used in the final determination of the Class 3 frequencies.P0247130002-47225-36 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsClass 8 SequencesThis group represents sequences where containment bypass occurs (SGTR or ISLOCA).ISLOCA and SGTR initiators are deemed inapplicable to the external events assessment so onlyinduced SGTR scenarios need to be considered. From the frequency information provided inTable 4.2-1 for IP2, the induced SGTR contribution to core damage is about 0.75% and for IP3it represented about 0.39%. A value of 0.5% is assumed for the external events contributionfor both IP2 and IP3. A High Early release magnitude dose is assigned.For IP2:Class_8 = 0.005 * [IP2 External Events CDF]= 0.005 * [5.01E-05]= 2.51E-07/yrFor IP3:Class_8 = 0.005 * [IP3 External Events CDF]= 0.005 * [5.20E-05]= 2.60E-07/yrClass 7 SeauencesThis group represents containment failure induced by early and late severe accidentphenomena. From Table 5.1-1 for IP2, the contribution from the early Class 7 sequences isabout 0.6% and for IP3 it represented about 1.3%. A value of 1.0% is assumed for theexternal events contribution for both IP2 and IP3. A High Early release magnitude dose isassigned. From Table 5.1-1 for IP2, the contribution from the late Class 7 sequences is about23% and for IP3 it represented about 15%. However, since the external events contributorsare more dominated by unrecoverable SBO-like scenarios, a value of 50% is assumed for theexternal events contribution for both IP2 and IP3. A High Late release magnitude dose isassigned.P0247130002-47225-37 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsFor IP2:Class_7-CFE = 0.01 * [IP2 External Events CDF]= 0.01 * [5.01E-05]= 5.01E-07/yrClass_7-CFL = 0.50 * [IP2 External Events CDF]= 0.50 * [5.01E-05]= 2.51E-05/yrForlP3:Class_7-CFE = 0.01 * [IP3 External Events CDF]= 0.01 * [5.20E-05]= 5.20E-07/yrClass_7-CFL = 0.50 * [IP3 External Events CDF]= 0.50 * [5.20E-05]= 2.60E-05/yrClass 4, 5. and 6 SequencesSimilar to the internal events assessment, because these failures are unaffected by the Type AILRT, these groups are not evaluated any further in this analysis.Class 2 SequencesThis group consists of large containment isolation failures. From the frequency informationprovided in Table 4.2-1 for IP2, the internal events contribution to this accident class wasapproximately 0.1% of the CDF and for IP3 it represented about 0.03%. Since seismic andfire initiated events would likely be more susceptible to this failure mode, the largercontribution of 0.1% is assumed for both IP2 and IP3. The population doses are assigned thesame as the Class 2 scenarios in the internal events assessment.P0247130002-47225-38 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsForIP2:Class_2 = 0.001 * [IP2 External Events CDF]= 0.001 * [5.01E-05]= 5.01E-08/yrFor IP3:Class_2 = 0.001 * [IP3 External Events CDF]= 0.001 * [5.20E-05]= 5.20E-08/yrClass 3 SequencesSimilar to the internal events assessment, the respective frequencies peras follows:year are determinedPROBciass_3aPROBclass_3b= probability of small pre-existing containment liner leakage= 0.0092 (see Section 4.3)= probability of large pre-existing containment liner leakage= 0.0023 (see Section 4.3)As described in Section 4.3, additional consideration is made to not apply these failureprobabilities to those cases that are already considered LERF scenarios (i.e., the Class 2, Class7, and Class 8 LERF contributions). This adjustment is made for based on the frequencyinformation described above for IP2 and IP3, respectively as shown below.For IP2:Class_3a = 0.0092 * [CDF -(Class 2 + Class 7-CFE + Class 8)]= 0.0092 * [5.01E-05 -(5.01E-08 + 5.01E-07 + 2.51E-07)]= 4.54E-07/yrClass_3b = 0.0023 * [CDF -(Class 2 + Class 7-CFE + Class 8)]= 0.0023 * [5.01E-05 -(5.01E-08 + 5.01E-07 + 2.51E-07)]= 1.13E-07/yrP0247130002-47225-39 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsFor IP3:Class_3a = 0.0092 * [CDF -(Class 2 + Class 7-CFE + Class 8)]= 0.0092 * [5.20E-05 -(5.20E-08 + 5.20E-07 + 2.60E-07)]= 4.71E-07/yrClass_3b = 0.0023 * [CDF -(Class 2 + Class 7-CFE + Class 8)]= 0.0023 * [5.20E-05 -(5.20E-08 + 5.20E-07 + 2.60E-07)]= 1.18E-07/yrFor this analysis, the associated containment leakage for Class 3a is 1OLa and 10OLa for Class3b, which is consistent with the latest EPRI methodology [3] and the NRC SE [7].Class 1 SequencesSimilar to the internal events assessment, the frequency is determined by subtracting allcontainment failure end states including the EPRI/NEI Class 3a and 3b frequency calculatedbelow, from the total CDF. The internal events intact containment dose of 4.41E+04person-rem for IP2 and IP3 is also utilized.Summary of Alternative External Events Base Case Dose AssessmentIn summary, the accident sequence frequencies that can lead to release of radionuclides to thepublic have been derived in a manner consistent with the definition of accident classes definedin EPRI 1018243 [3]. These frequencies have been combined with reasonable assumptionsregarding the population dose associated with each class to determine the base casepopulation dose risk for external events. This information is provided in Table 5.7-5a for IP2and in Table 5.7-5b for IP3. Additionally, following the same EPRI methodology utilized forinternal events to determine the risk impact assessment of extending the ILRT interval, theexternal events accident class frequencies indicative of a 15 year ILRT interval are provided inTable 5.7-6a for IP2 and in Table 5.7-6b for IP3.Table 5.7-7 then shows the changes due to the ILRT extension from 3 year to a 15 yearinterval in the LERF, person-rem/yr, and CCFP figures of merit. When these values are addedto the internal events results, the acceptance criteria are all still met by using this detailedalternative external events evaluation instead of the simple evaluation that was utilized inSection 5.7.5. A comparison to the acceptance criteria is also shown in Table 5.7-7. Note thatthe ALERF, person-rem/yr, and change in CCFCP shown in Table 5.7-7 are all slightly higherthan the corresponding values shown in Table 5.7-3. This is because the simple method inTable 5.7-3 assumes the same distribution of LERF contributors exists between the internalP0247130002-47225-40 Risk Impact Assessment of Extending the Indian Point ILRT Intervalsand external events models whereas the alternative assessment re-apportions the base caseLERF contributions based on more realistic assumptions while conservatively maintaining thetotal CDF value. That is, since the contribution from SGTR initiators and ISLOCA initiators(which contribute to the base LERF value) are not applicable to the external eventscontribution, more of the remaining CDF distribution is potentially affected by the ILRTextension as represented by the Class 3b multiplier on CDF (that is not already LERF).Additionally, the alternative detailed assessment leads to slightly different percent increases inperson-rem/yr which are a function of the base case dose estimates.TABLE 5.7-5APOPULATION DOSE RISK AS A FUNCTION OF ACCIDENT CLASS(IP2 ALTERNATIVE EXTERNAL EVENTS BASE CASE)ACCIDENT DESCRIPTION FREQUENCY DOSE DOSE RISKCLASS (1/YR) (PERSON- (PERSON-(CONTAINMENT REM) REM/YR)RELEASE TYPE)1 Containment Intact 2.37E-05 4.41E+04 1.04E+002 Large Isolation Failures 5.01E-08 6.51E+07 3.26E+00(Failure to Close)3a Small Isolation Failures (liner 4.54E-07 4.41E+05 2.OOE-01breach)3b Large Isolation Failures (liner 1.13E-07 4.41E+06 5.OOE-01breach)4 Small Isolation Failures N/A N/A N/A(Failure to seal -Type B)5 Small Isolation Failures N/A N/A N/A(Failure to seal-Type C)6 Other Isolation Failures (e.g., N/A N/A N/Adependent failures)7-CFE Failures Induced by 5.01E-07 6.51E+07 3.26E+01Phenomena (Early)7-CFL Failures Induced by 2.51E-05 1.63E+07 4.08E+02Phenomena (Late)8-SGTR Containment Bypass (Steam 2.51E-07 6.51E+07 1.63E+01Generator Tube Rupture)8-ISLOCA Containment Bypass O.OOE+00 6.51E+07 O.OOE+00(Interfacing System LOCA)CDF All CET End States 5.01E-05 462.2(Including Intact Case)P0247130002-47225-41 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-5BPOPULATION DOSE RISK AS A FUNCTION OF ACCIDENT CLASS(IP3 ALTERNATIVE EXTERNAL EVENTS BASE CASE)ACCIDENT DESCRIPTION FREQUENCY DOSE DOSE RISKCLASS (1/YR) (PERSON- (PERSON-(CONTAINMENT REM) REM/YR)RELEASE TYPE)1 Containment Intact 2.46E-05 4.41E+04 1.08E+002 Large Isolation Failures 5.20E-08 5.08E+07 2.64E+00(Failure to Close)3a Small Isolation Failures (liner 4.71E-07 4.41E+05 2.08E-01breach)3b Large Isolation Failures (liner 1.18E-07 4.41E+06 5.19E-01breach)4 Small Isolation Failures N/A N/A N/A(Failure to seal -Type B)5 Small Isolation Failures N/A N/A N/A(Failure to seal-Type C)6 Other Isolation Failures (e.g., N/A N/A N/Adependent failures)7-CFE Failures Induced by 5.20E-07 5.08E+07 2.64E+01Phenomena (Early)7-CFL Failures Induced by 2.60E-05 1.63E+07 4.24E+02Phenomena (Late)8-SGTR Containment Bypass (Steam 2.60E-07 5.08E+07 1.32E+01Generator Tube Rupture)8-ISLOCA Containment Bypass O.OOE+00 5.08E+07 O.OOE+00(Interfacing System LOCA)CDF All CET End States 5.20E-05 467.9(Including Intact Case)P0247130002-47225-42 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-6APOPULATION DOSE RISK AS A FUNCTION OF ACCIDENT CLASS (IP2 ALTERNATIVEEXTERNAL EVENTS EVALUATION CHARACTERISTIC OF CONDITIONS FOR 1 IN 15YEAR ILRT FREQUENCY)ACCIDENT DESCRIPTION FREQUENCY DOSE DOSE RISKCLASS (1/YR) (PERSON- (PERSON-(CONTAINMENT REM) REM/YR)RELEASE TYPE)1 Containment Intact 2.14E-05 4.41E+04 9.44E-012 Large Isolation Failures 5.01E-08 6.51E+07 3.26E+00(Failure to Close)3a Small Isolation Failures (liner 2.27E-06 4.41E+05 1.OOE+00breach)3b Large Isolation Failures (liner 5.67E-07 4.41E+06 2.50E+00breach)4 Small Isolation Failures N/A N/A N/A(Failure to seal -Type B)5 Small Isolation Failures N/A N/A N/A(Failure to seal-Type C)6 Other Isolation Failures (e.g., N/A N/A N/Adependent failures)7-CFE Failures Induced by 5.01E-07 6.51E+07 3.26E+01Phenomena (Early)7-CFL Failures Induced by 2.51E-05 1.63E+07 4.08E+02Phenomena (Late)8-SGTR Containment Bypass (Steam 2.51E-07 6.51E+07 1.63E+01Generator Tube Rupture)8-ISLOCA Containment Bypass O.OOE+00 6.51E+07 O.OOE+00_ (Interfacing System LOCA)CDF All CET End States 5.01E-05 [ 464.9_ (Including Intact Case) IP0247130002-47225-43 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-6BPOPULATION DOSE RISK AS A FUNCTION OF ACCIDENT CLASS (IP3 ALTERNATIVEEXTERNAL EVENTS EVALUATION CHARACTERISTIC OF CONDITIONS FOR 1 IN 15YEAR ILRT FREQUENCY)ACCIDENT DESCRIPTION FREQUENCY DOSE DOSE RISKCLASS (1/YR) (PERSON- (PERSON-(CONTAINMENT REM) REM/YR)RELEASE TYPE)1 Containment Intact 2.22E-05 4.41E+04 9.80E-012 Large Isolation Failures 5.20E-08 5.08E+07 2.64E+00(Failure to Close)3a Small Isolation Failures (liner 2.35E-06 4.41E+05 1.04E+00breach)3b Large Isolation Failures (liner 5.88E-07 4.41E+06 2.59E+00breach)4 Small Isolation Failures N/A N/A N/A(Failure to seal -Type B)5 Small Isolation Failures N/A N/A N/A(Failure to seal-Type C)6 Other Isolation Failures (e.g., N/A N/A N/Adependent failures)7-CFE Failures Induced by 5.20E-07 5.08E+07 2.64E+01Phenomena (Early)7-CFL Failures Induced by 2.60E-05 1.63E+07 4.24E+02Phenomena (Late)8-SGTR Containment Bypass (Steam 2.60E-07 5.08E+07 1.32E+01Generator Tube Rupture)8-ISLOCA Containment Bypass O.OOE+00 5.08E+07 O.OOE+00(Interfacing System LOCA)CDF All CET End States I 5.20E-05 470.7(Including Intact Case) IP0247130002-47225-44 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-7COMPARISON TO ACCEPTANCE CRITERIA INCLUDING ALTERNATIVEEXTERNAL EVENTS EVALUATION CONTRIBUTION FOR IP2 AND IP3Contributor ALERF APerson-rem/yr ACCFPIP2 Internal 9.84E-8/yr 0.584/yr (0.62%) 0.84%EventsIP2 External 4.54E-7/yr 2.70/yr (0.58%) 0.91%EventsIndian Point 2 5.52E-7/yr 3.28/yr (0.59%) 0.89%TotalIP3 Internal 1.26E-7/yr 0.751/yr (0.93%) 0.85%EventsIP3 External 4.71E-7/yr 2.80/yr (0.60%) 0.91%EventsIndian Point 3 5.96E-7/yr 3.55/yr (0.65%) 0.89%TotalAcceptance < 1.OE-6/yr <1.0 person- < 1.5%0/Criteria rem/yr or <1.0%The 5.52E-07/yr increase in LERF for IP2 and the 5.97E-07/yr increase in LERF for IP3 due tothe combined internal and external events from extending the ILRT frequency from 3-per-10years to 1-per-15 years falls within Region II between 1.0E-7 to 1.0E-6 per reactor year("Small Change" in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when thecalculated increase in LERF due to the proposed plant change is in the "Small Change" range,the risk assessment must also reasonably show that the total LERF is less than 1.0E-5/yr.From Table 4.2-1, the total LERF due to postulated internal event accidents is 1.16E-06/yr forIP2 and 1.25E-06/yr for IP3. From Table 5.7-5a for IP2, the base external events LERF can bederived from the Class 2, Class 3b, Class 7-CFE, and Class 8 contributions. From the individualcontributions of 5.01E-08/yr + 1.13E-07/yr + 5.01E-07/yr + 2.51E-07/yr, this equates to9.15E-07/yr. From Table 5.7-5b for IP3, the individual contributions of 5.20E-08/yr +1.18E-07/yr + 5.20E-07/yr + 2.60E-07/yr result in a total base case LERF from externalevents of 9.50E-07/yr. The total LERF values for IP2 and IP3 using the alternative externalevents evaluation are then shown in Table 5.7-8.P0247130002-47225-45 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-8IMPACT OF 15-YR ILRT EXTENSION ON LERF FOR IP2 AND IP3LERF CONTRIBUTOR IP2 (1/YR) IP3 (1/YR)Internal Events LERF 1.16E-06 1.25E-06External Events LERF 9.15E-07 9.50E-07Internal Events LERF due to 1.23E-07 1.58E-07ILRT (at 15 years) (1)External Events LERF increasedue to ILRT extension (2) 4.54E-07 4.71E-07Total 2.65E-06/yr 2.83E-06/yr(1) Including age adjusted steel liner corrosion likelihood as reported in Table 5.7-2a for IP2and Table 5.7-2b for IP3.(2) As shown in Table 5.7-7. This did not include the age adjusted steel liner corrosionlikelihood, but this was demonstrated to be a small contributor for IP2 and IP3.As can be seen, the total LERF for IP2 is estimated as 2.65E-06/yr and for IP3 it is2.83E-06/yr. These values are both less than the RG 1.174 requirement to demonstrate thatthe total LERF due to internal and external events is less than 1.OE-5/yr.P0247130002-47225-46 Risk Impact Assessment of Extending the Indian Point ILRT Intervals5.8 CONTAINMENT OVERPRESSURE IMPACTS ON CDFFor IP2 and IP3, ECCS NPSH calculations made in support of the GSI-191 effort [28, 29]confirmed that containment overpressure is not required to obtain adequate NPSH [30]. Thisis consistent with the PRA models which indicate there is no impact on CDF from the ILRTextension risk assessment.In IP-CALC-06-000231 [28], the NPSHA / NPSHR relationship for IP2 ECCS pumpswas being evaluated. For conservatism in obtaining the NPSHA and NPSHR, themaximum volumetric flow rate was used. The greatest volumetric flow rate occurswhen the least dense fluid is being pumped. This is at the highest temperature in therecirculation phase of the accident. For IP2, this temperature was 264.4 F whichoccurs at start of recirculation. Since 264.4 F is higher than 212 F, a boundarycondition pressure of 37.6 psia is inputted. This is close to the saturation pressure at264.4 F so there is essentially no containment overpressure being invoked. In otherwords, 264.4 F and 37.6 psia is basically equivalent to 212 F and 14.7 psia (0 psig).* The same issue was addressed in IP-CALC-07-00054 [29] for the TP3 NPSHA /NPSHR evaluation. Again, to be most conservative with respect to NPSHA andNPSHR, the maximum volumetric flow rate has to be used. This entails that thehighest temperature during recirculation applies. This is 242.8 F at commencementof recirculation. The saturation pressure at 242.8 F is close to 26.1 psia, which is theboundary condition pressure input in the calculation. Again, essentially nocontainment overpressure is being invoked since 242.8 F and 26.1 psia is basicallyequivalent to 212 F and 14.7 psia (0 psig).P0247130002-47225-47 Risk Impact Assessment of Extending the Indian Point ILRT Intervals6.0 SENSITIVITIES6.1 SENSITIVITY TO CORROSION IMPACT ASSUMPTIONSThe results in Tables 5.2-2a(b), 5.3-la(b), and 5.3-2a(b) show that including corrosion effectscalculated using the assumptions described in Section 4.4 does not significantly affect theresults of the ILRT extension risk assessment. In any event, sensitivity cases were developedto gain an understanding of the sensitivity of the results to the key parameters in the corrosionrisk analysis. The time for the flaw likelihood to double was adjusted from every five years toevery two and every ten years. The failure probabilities for the cylinder, dome and basematwere increased and decreased by an order of magnitude. The total detection failure likelihoodwas adjusted from 10% to 15% and 5%. The results are presented in Table 6.1-1a for IP2and in Table 6.1-1b for IP3. In every case, the impact from including the corrosion effects isvery minimal. Even the upper bound estimates with very conservative assumptions for all ofthe key parameters yield increases in LERF due to corrosion of only 3.68E-8/yr for IP2 and4.72E-08/yr for IP3. The results indicate that even with very conservative assumptions, theconclusions from the base analysis would not change.TABLE 6.1-1ASTEEL LINER CORROSION SENSITIVITY CASES FOR IP2AGE CONTAINMENT VISUAL INCREASE IN CLASS 3B(STEP 3 IN THE BREACH INSPECTION FREQUENCY (LERF)CORROSION (STEP 4 IN THE & NON- FOR ILRT EXTENSIONANALYSIS) CORROSION VISUAL FROM 3 IN 10 TO 1 IN 15 YEARSANALYSIS) FLAWS (PER YEAR)(STEP 5 IN THECORROSION TOTAL INCREASE DUE TOANALYSIS) INCREASE CORROSIONBase Case Base Case Base Case 9.84E-08 1.16E-09Doubles every (1.0% Cylinder- (10% Cylinder-5 yrs Dome, Dome,0.1% Basemat) 100% Basemat)Doubles every Base Base 9.99E-08 2.63E-092 yrsDoubles every Base Base 9.83E-08 9.68E-1010 yrsBase Base 15% Cylinder- 9.89E-08 1.62E-09DomeP0247130002-47226-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 6.1-1ASTEEL LINER CORROSION SENSITIVITY CASES FOR IP2AGE CONTAINMENT VISUAL INCREASE IN CLASS 3B(STEP 3 IN THE BREACH INSPECTION FREQUENCY (LERF)CORROSION (STEP 4 IN THE & NON- FOR ILRT EXTENSIONANALYSIS) CORROSION VISUAL FROM 3 IN 10 TO 1 IN 15 YEARSANALYSIS) FLAWS (PER YEAR)(STEP 5 IN THECORROSION TOTAL INCREASE DUE TOANALYSIS) INCREASE CORROSIONBase Base 5% Cylinder- 9.79E-08 6.97E-10DomeBase 10% Cylinder- Base 1.09E-07 1.16E-08Dome,1% BasematBase 0.1% Cylinder- Base 9.74E-08 1.16E-10Dome,0.01% BasematLOWER BOUNDDoubles every 0.1% Cylinder- 5% Cylinder- 9.73E-08 5.81E-1110 yrs Dome, Dome,0.01% Basemat 100% BasematUPPER BOUNDDoubles every 10% Cylinder- 15% Cylinder- 1.34E-07 3.68E-082 yrs Dome, Dome,1% Basemat 100% BasematP0247130002-47226-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 6.1-1BSTEEL LINER CORROSION SENSITIVITY CASES FOR IP3AGE CONTAINMENT VISUAL INCREASE IN CLASS 3B(STEP 3 IN THE BREACH INSPECTION FREQUENCY (LERF)CORROSION (STEP 4 IN THE & NON- FOR ILRT EXTENSIONANALYSIS) CORROSION VISUAL FROM 3 IN 10 TO 1 IN 15 YEARSANALYSIS) FLAWS (PER YEAR)(STEP 5 IN THECORROSION TOTAL INCREASE DUE TOANALYSIS) INCREASE CORROSIONBase Case Base Case Base Case 1.26E-07 1.49E-09Doubles every (1.0% Cylinder- (10% Cylinder-5 yrs Dome, Dome,0.1% Basemat) 100% Basemat)Doubles every Base Base 1.28E-07 3.37E-092 yrsDoubles every Base Base 1.26E-07 1.24E-0910 yrsBase Base 15% Cylinder- 1.27E-07 2.08E-09DomeBase Base 5% Cylinder- 1.26E-07 8.95E-10DomeBase 10% Cylinder- Base 1.40E-07 1.49E-08Dome,1% BasematBase 0.1% Cylinder- Base 1.25E-07 1.49E-10Dome,0.01% BasematLOWER BOUNDDoubles every 0.1% Cylinder- 5% Cylinder- 1.25E-07 7.47E-1110 yrs Dome, Dome,0.01% Basemat 100% BasematUPPER BOUNDDoubles every 100/a Cylinder- 15% Cylinder- 1.72E-07 4.72E-082 yrs Dome, Dome,1% Basemat 100% BasematP0247130002-47226-3 Risk Impact Assessment of Extending the Indian Point ILRT Intervals6.2 EPRI EXPERT ELICITATION SENSITIVITYAn expert elicitation was performed to reduce excess conservatisms in the data associated withthe probability of undetected leaks within containment [3]. Since the risk impact assessmentof the extensions to the ILRT interval is sensitive to both the probability of the leakage as wellas the magnitude, it was decided to perform the expert elicitation in a manner to solicit theprobability of leakage as a function of leakage magnitude. In addition, the elicitation wasperformed for a range of failure modes which allowed experts to account for the range offailure mechanisms, the potential for undiscovered mechanisms, inaccessible areas of thecontainment as well as the potential for. detection by alternate means. The expert elicitationprocess has the advantage of considering the available data for small leakage events, whichhave occurred in the data, and extrapolate those events and probabilities of occurrence to thepotential for large magnitude leakage events.The basic difference in the application of the ILRT interval methodology using the expertelicitation is a change in the probability of pre-existing leakage within containment. The basecase methodology uses the Jeffrey's non-informative prior for the large leak size and theexpert elicitation sensitivity study uses the results from the expert elicitation. In addition,given the relationship between leakage magnitude and probability, larger leakage that is morerepresentative of large early release frequency can be reflected. For the purposes of thissensitivity, the same leakage magnitudes that are used in the base case methodology (i.e.,1OLa for small and 10OLa for large) are used here. Table 6.2-1 illustrates the magnitudes andprobabilities of a pre-existing leak in containment associated with the base case and the expertelicitation statistical treatments. These values are used in the ILRT interval extension for thebase methodology and in this sensitivity case. Details of the expert elicitation process,including the input to expert elicitation as well as the results of the expert elicitation, areavailable in the various appendices of EPRI 1018243 [3].TABLE 6.2-1EPRI EXPERT ELICITATION RESULTSLEAKAGE SIZE (LA) BASE CASE MEAN EXPERT PERCENTPROBABILITY OF ELICITATION MEAN REDUCTIONOCCURRENCE PROBABILITY OFOCCURRENCE [3]10 9.2E-03 3.88E-03 58%100 2.3E-03 2.47E-04 89%P0247130002-47226-4 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe summary of results using the expert elicitation values for probability of containmentleakage is provided in Table 6.2-2a for IP2 and in Table 6.2-2b for 1P3. As mentionedpreviously, probability values are those associated with the magnitude of the leakage used inthe base case evaluation (1OLa for small and 10OLa for large). The expert elicitation processproduces a relationship between probability and leakage magnitude in which it is possible toassess higher leakage magnitudes that are more reflective of large early releases; however,these evaluations are not performed in this particular study.The net effect is that the reduction in the multipliers shown above also leads to a dramaticreduction on the calculated increases in the LERF values. As shown in Table 6.2-2a for IP2, theincrease in the overall value for LERF due to Class 3b sequences that is due to increasing theILRT test interval from 3 to 15 years is just 1.05E-08/yr. Similarly, the increase due toincreasing the interval from 10 to 15 years is just 4.40E-09/yr. As shown in Table 6.2-2b for1P3, the increase in the overall value for LERF due to Class 3b sequences that is due toincreasing the ILRT test interval from 3 to 15 years is just 1.34E-08/yr. Similarly, the increasedue to increasing the interval from 10 to 15 years is just 5.60E-09/yr. As such, if the expertelicitation probabilities of occurrence are used instead of the non-informative prior estimates,the change in LERF for IP2 and IP3 is within the range of a "very small" change in risk whencompared to the current 1-in-10, or baseline 3-in-10 year requirement. Additionally, as shownin Table 6.2-2a for IP2 and Table 6.2-2b for IP3, the increase in dose rate and CCFP aresimilarly reduced to much smaller values. The results of this sensitivity study are judged to bemore indicative of the actual risk associated with the ILRT extension than the results from theassessment as dictated by the values from the EPRI methodology [3], and yet are stillconservative given the assumption that all of the Class 3b contribution is considered to beLERF.P0247130002-47226-5 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 6.2-2AIP2 ILRT CASES:3 IN 10 (BASE CASE), 1 IN 10, AND 1 IN 15 YR INTERVALS(BASED ON EPRI EXPERT ELICITATION LEAKAGE PROBABILITIES)EPRI DOSE BASE CASE EXTEND TO EXTEND TOCLASS PER-REM 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARSCDF PERSON- CDF PERSON- CDF PERSON-(1/YR) REM/YR (1/YR) REM/YR (1/YR) REM/YR1 4.41E+04 7.82E-06 3.45E-01 7.71E-06 3.40E-01 7.64E-06 3.37E-012 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 1.11E-08 7.23E-013a 4.41E+05 4.10E-08 1.81E-02 1.37E-07 6.03E-02 2.05E-07 9.05E-023b 4.41E+06 2.61E-09 1.15E-02 8.70E-09 3.84E-02 1.31E-08 5.76E-027-CFE 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00 7.37E-08 4.58E+007-CFL 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01 2.71E-06 1.86E+018-SGTR 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01 1.05E-06 6.80E+018-ISLOCA 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00 2.77E-08 1.80E+00Total 1.17E-05 9.414E+01 1.17E-05 9.421E+01 1.17E-05 19.425E+01ILRT Dose Rate from 2.96E-02 9.86E-02 1.48E-013a and 3bDelta From 3 yr --- 6.45E-02 1.11E-01TotalDose From 10 yr --- 4.62E-02DoseRate(1)3b Frequency (LERF) 2.61E-09 8.70E-09 1.31E-08Delta 3b From 3 yr --- 6.09E-09 1.05E-08LERF From 10 yr .... -- 4.40E-09CCFP % 33.00% 33.05% 33.09%Delta From 3 yr --- 0.05% 0.09%CCFP %From 10 yr --- 0.04%(1) The overall difference in total dose rate is less than the difference of only the 3a and 3bcategories between two testing intervals. This is due to the fact that the Class 1 person-rem/yr decreases when extending the ILRT frequency.P0247130002-47226-6 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 6.2-2BIP3 ILRT CASES:3 IN 10 (BASE CASE), 1 IN 10, AND 1 IN 15 YR INTERVALS(BASED ON EPRI EXPERT ELICITATION LEAKAGE PROBABILITIES)EPRI DOSE BASE CASE EXTEND TO EXTEND TOCLASS PER-REM 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARSCDF PERSON- CDF PERSON- CDF PERSON-(1/YR) REM/YR (1/YR) REM/YR (1/YR) [ REM/YR1 4.41E+04 1.12E-05 4.96E-01 1.11E-05 4.90E-01 1.10E-05 4.86E-012 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 3.99E-09 2.03E-013a 4.41E+05 5.27E-08 2.32E-02 1.76E-07 7.74E-02 2.64E-07 1.16E-013b 4.41E+06 3.36E-09 1.48E-02 1.12E-08 4.93E-02 1.68E-08 7.40E-027-CFE 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00 1.88E-07 5.97E+007-CFL 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01 2.17E-06 1.49E+018-SGTR 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01 9.77E-07 4.96E+018-ISLOCA 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00 1.93E-07 9.80E+00Total 1.48E-05 8.099E+01 1.48E-05 18.108E+01 I 1.48E-05 18.114E+01ILRT Dose Rate from 3.81E-02 1.27E-01 1.90E-013a and 3bDelta From 3 yr --- 8.29E-02 1.42E-01TotalDose From 10 yr --- 5.94E-02DoseRate*1)3b Frequency (LERF) 3.36E-09 1.12E-08 1.68E-08Delta 3b From 3 yr --- 7.84E-09 1.34E-08LERF IFrom 10 yr ....5.60E-09CCFP % 23.84% 23.89% 23.93%Delta From 3 yr --- 0.05% 0.09%CCFP %From 10 yr --.--- 0.04%( The overall difference in total dose rate is less than the difference of only the 3a and 3bcategories between two testing intervals. This is due to the fact that the Class 1 person-rem/yr decreases when extending the ILRT frequency.P0247130002-47226-7 Risk Impact Assessment of Extending the Indian Point ILRT Intervals

7.0 CONCLUSION

SBased on the results from Section 5 and the sensitivity calculations presented in Section 6, thefollowing conclusions regarding the assessment of the plant risk are associated withpermanently extending the Type A ILRT test frequency to fifteen years:* Reg. Guide 1.174 [4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Reg. Guide 1.174 defines "very small"changes in risk as resulting in increases of CDF below 1.OE-06/yr and increases inLERF below 1.OE-07/yr. "Small" changes in risk are defined as increases in CDFbelow 1.OE-05/yr and increases in LERF below 1.OE-06/yr. Since the ILRT extensionwas demonstrated to have no impact on CDF for IP2 and IP3, the relevant criterion isLERF. The increase in internal events LERF resulting from a change in the Type AILRT test interval for the base case with corrosion included for IP2 is 9.84E-08/yr(see Table 5.6-1a). In using the EPRI Expert Elicitation methodology, the change isestimated as 1.05E-08/yr (see Table 6.2-2a). Both of these values fall within thevery small change region of the acceptance guidelines in Reg. Guide 1.174. For IP3,the increase is estimated at 1.26E-07/yr (see Table 5.6-1b), which is within thesmall change region of the acceptance guidelines in Reg. Guide 1.174. In using theEPRI Expert Elicitation methodology, the change is estimated as 1.34E-08/yr (seeTable 6.2-2b), which is within the very small change region of the acceptanceguidelines in Reg. Guide 1.174.* The change in dose risk for changing the Type A test frequency from three-per-tenyears to once-per-fifteen-years, measured as an increase to the total integrated doserisk for all internal events accident sequences for IP2, is 0.584 person-rem/yr(0.62%) using the EPRI guidance with the base case corrosion case (Table 5.6-1a).The change in dose risk drops to 1.11E-01 person-rem/yr when using the EPRIExpert Elicitation methodology (Table 6.2-2a). For IP3, it is 0.751 person-rem/yr(0.93%) using the EPRI guidance with the base case corrosion case (Table 5.6-1b).The change in dose risk drops to 1.42E-01 person-rem/yr when using the EPRIExpert Elicitation methodology (Table 6.2-2b). The values calculated per the EPRIguidance are all lower than the acceptance criteria of 51.0 person-rem/yr or <1.0%person-rem/yr defined in Section 1.3.* The increase in the conditional containment failure frequency from the three in tenyear interval to one in fifteen years including corrosion effects using the EPRIguidance (see Section 5.5) is 0.84% for IP2 and 0.85% for IP3. This value drops toless that 0.10% for IP2 and IP3 using the EPRI Expert Elicitation methodology (seeTable 6.2-2a and Table 6.2-2b, respectively). This is below the acceptance criteria ofless than 1.5% defined in Section 1.3.* To determine the potential impact from external events, a bounding assessmentfrom the risk associated with external events utilizing information from the IP2 andIP3 IPEEEs similar to the approach used in the License Renewal SAMA analysis wasperformed. As shown in Table 5.7-2a for IP2, the total increase in LERF due tointernal events and the bounding external events assessment is 5.20E-07/yr. Asshown in Table 5.7-2b for IP3, the total increase in LERF due to internal events andthe bounding external events assessment is 5.70E-07/yr. Both of these values are inRegion II of the Reg. Guide 1.174 acceptance guidelines.P0247130002-47227-1 Risk Impact Assessment of Extending the Indian Point ILRT Intervals* As shown in Table 5.7-4, the same bounding analysis indicates that the total LERFfrom both internal and external risks is 6.78E-06/yr for IP2 and 6.34E-06/yr for IP3,which are less than the Reg. Guide 1.174 limit of 1.OE-05/yr given that the ALERF isin Region II (small change in risk)." Finally, since the external events assessment led to exceeding one of the twoalternative acceptance criteria (i.e. greater than 1.0 person-rem/yr, an alternativedetailed bounding external events assessment was also performed to demonstratethat the alternate 1.0% person-rem/yr criterion and the other acceptance criteriacould still be met. In this case, as shown in Table 5.7-7 for IP2, the total change inLERF from both internal and external events was 5.52E-7/yr, the change in person-rem/yr was 3.28/yr representing 0.59% of the total, and the change in the CCFP was0.89%. For IP3, the total change in LERF from both internal and external events was5.97E-7/yr, the change in person-rem/yr was 3.55/yr representing 0.65% of thetotal, and the change in the CCFP was 0.89%. All of these calculated changes meetthe acceptance criteria. As shown in Table 5.7-8, this assessment indicates that thetotal LERF from both internal and external risks is 2.65E-06/yr for IP2 and 2.83E-06/yr for IP3, which are less than the Reg. Guide 1.174 limit of 1.OE-05/yr given thatthe ALERF is in Region II (small change in risk).* Including age-adjusted steel liner corrosion effects in the ILRT assessment wasdemonstrated to be a small contributor to the impact of extending the ILRT intervalfor IP2 and IP3.Therefore, increasing the ILRT interval on a permanent basis to a one-in-fifteen year frequencyis not considered to be significant since it represents only a small change in the IP2 and IP3risk profiles.Previous AssessmentsThe NRC in NUREG-1493 [6] has previously concluded the following:* Reducing the frequency of Type A tests (ILRTs) from three per 10 years to one per20 years was found to lead to an imperceptible increase in risk. The estimatedincrease in risk is very small because ILRTs identify only a few potential containmentleakage paths that cannot be identified by Type B and C testing, and the leaks thathave been found by Type A tests have been only marginally above existingrequirements.* Given the insensitivity of risk to containment leakage rate and the small fraction ofleakage paths detected solely by Type A testing, increasing the interval betweenintegrated leakage-rate tests is possible with minimal impact on public risk. Theimpact of relaxing the ILRT frequency beyond one in 20 years has not beenevaluated. Beyond testing the performance of containment penetrations, ILRTs alsotest the integrity of the containment structure.The findings for IP2 and IP3 confirm these general findings on a plant specific basis consideringthe severe accidents evaluated, the containment failure modes, and the local populationsurrounding IP2 and IP3.P0247130002-47227-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy

8.0 REFERENCES

[1] Nuclear Energy Institute, Industry Guideline for Implementing Performance-BasedOption of 10 CFR Part 50, Appendix J, NEI 94-01, Revision 3-A, July 2012.[2] Electric Power Research Institute, Risk Impact Assessment of Revised ContainmentLeak Rate Testing Intervals, EPRI TR-104285, August 1994.[3] Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals:Revision 2-A of 1009325. EPRI, Palo Alto, CA: October 2008. 1018243.[4] U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic RiskAssessment in Risk-Informed Decisions on Plant-Specific Changes to the LicensingBasis, Regulatory Guide 1.174, Revision 2, May 2011.[5] Letter from Mr. C. H. Cruse (Constellation Nuclear, Calvert Cliffs Nuclear PowerPlant) to U.S. Nuclear Regulatory Commission, Response to Request for AdditionalInformation Concerning the License Amendment Request for a One-Time IntegratedLeakage Rate Test Extension, Accession Number ML020920100, March 27, 2002.[6] U.S. Nuclear Regulatory Commission, Performance-Based Containment Leak-TestProgram, NUREG-1493, September 1995.[7] U.S. Nuclear Regulatory Commission, Final Safety Evaluation for Nuclear EnergyInstitute (NEI) Topical Report (TR) 94-01, Revision 2, "Industry Guideline forImplementing Performance-Based Option Of 10 CFR Part 50, Appendix J" and ElectricPower Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007,"Risk Impact Assessment Of Extended Integrated Leak Rate Testing Intervals" (TACNo. MC9663), Accession Number ML081140105, June 25, 2008.[8] Consolidated Edison Company of New York, Individual Plant Examination for ExternalEvents for Indian Point Unit 2 Nuclear Generating Station, Revision 0, December1995.[9] New York Power Authority, Indian Point Three Nuclear Power Plant Individual PlantExamination for External Events, IP3-RPT-UNSPEC-02182, Revision 0, September1997.[10] Entergy Nuclear, Re-analysis of MACCS2 Models for IPEC, Calculation IP-CALC-09-00265, December 2009.[11] Entergy Nuclear, MAAP/MACCS2 Computer Codes Calculated Dose for IPECContainment Structure Based on Allowable Leakage From an Intact Containment,Calculation IP-CALC-13-00042, September 2013.[12] ERIN Engineering and Research, Shutdown Risk Impact Assessment for ExtendedContainment Leakage Testing Intervals Utilizing ORAMTM, EPRI TR-105189, FinalReport, May 1995.[13] Oak Ridge National Laboratory, Impact of Containment Building Leakage on LWRAccident Risk, NUREG/CR-3539, ORNL/TM-8964, April 1984.[14] Pacific Northwest Laboratory, Reliability Analysis of Containment Isolation Systems,NUREG/CR-4220, PNL-5432, June 1985.P0247130002-47228-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy[15] U.S. Nuclear Regulatory Commission, Technical Findings and Regulatory Analysis forGeneric Safety Issue II.E.4.3 (Containment Integrity Check), NUREG-1273, April1988.[16] Pacific Northwest Laboratory, Review of Light Water Reactor RegulatoryRequirements, NUREG/CR-4330, PNL-5809, Vol. 2, June 1986.[17] U.S. Nuclear Regulatory Commission, Severe Accident Risks: An Assessment for FiveU.S. Nuclear Power Plants, NUREG-1150, December 1990.[18] Entergy Nuclear, Indian Point Unit 2 Probabilistic Safety Assessment (PSA),Calculation IP-RPT-09-00026, Revision 0, November 2011.[19] Entergy Nuclear, Indian Point Unit 3 Probabilistic Safety Assessment (PSA),Calculation IP-RPT-10-00023, Revision 0, November 2012.[20] Entergy Nuclear, Indian Point Units 2 & 3, License Renewal Application, Appendix E,Applicant's Environmental Report, Accession Number ML071210530, April 23, 2007.[21] Letter from P. B. Cowan (Exelon Nuclear, Peach Bottom) to U.S. Nuclear RegulatoryCommission, Response to Request for Additional Information -License AmendmentRequest for Type A Test Extension, Accession Number ML100560433, February 25,2010.[22] Letter from P. B. Cowan (Exelon Nuclear, Peach Bottom) to U.S. Nuclear RegulatoryCommission, License Amendment Request -Type A Test Extension, AccessionNumber ML092440053, August 28, 2009.[23] Letter from Dave Morey (Southern Company, Farley Project) to U.S. NuclearRegulatory Commission, Joseph M. Farley Nuclear Plant Technical SpecificationRevision Request Integrated Leakage Rate Testing Interval Extension, NEL-02-0001,Accession Number ML020990040, April 4, 2002.[24] Letter from D.E. Young (Florida Power, Crystal River) to U.S. Nuclear RegulatoryCommission, License Amendment Request #267, Revision 1, Supplemental Risk-Informed Information in Support of License Amendment Request #267, Revision 0,3F0401-11, Accession Number ML011210207, April 25, 2001.[25] Letter from John A. Ventosa (Entergy, Indian Point Energy Center) to U.S. NuclearRegulatory Commission, Indian Point Nuclear Power Plant Units 2 and 3Reassessment of the Seismic Core Damage Frequency, NL-13-084, AccessionNumber ML13183A279, June 26, 2013.[26] Letter from Thomas P. Kirwin (Entergy, Palisades Nuclear Plant) to U.S. NuclearRegulatory Commission, License Amendment Request to Extend the ContainmentType A Leak Rate Test Frequency to 15 Years, Accession Number ML110970616,April 6, 2011.[27] U.S. Nuclear Regulatory Commission, Palisades Nuclear Plant -Issuance ofAmendment to Extend the Containment Type A Leak Rate Test Frequency to 15Years (TAC No. ME5997), Accession Number ML120740081, April 23, 2012.[28] Westinghouse, Indian Point Unit 2 SI Recirculation (LHSI and HHSI) Performance forthe Containment Sump Program, Entergy Calculation IP-CALC-06-00231, Revision 1,April 2010.P0247130002-47228-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy[29] Westinghouse, Indian Point Unit 3 SI Recirculation (LHSI and HHSI) Performance forthe Containment Sump Program, Entergy Calculation IP-CALC-07-00054, Revision 2,June 2010.[30] E-Mail from D. Gaynor (Entergy) to D. Vanover (ERIN), FW: Inputs for NPSH Calcs,July 24, 2013.[31] U.S. Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October1975.P0247130002-47228-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyAppendix APRA Technical AdequacyP0247130002-4722 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyNote that the information provided in this appendix was provided by Entergy personnel.A. 1 OVERVIEWA technical Probabilistic Risk Assessment (PRA) analysis is presented in this report to helpsupport an extension of the IP2 and IP3 containment Type A test integrated leak rate test(ILRT) interval to fifteen years.The analysis follows the guidance provided in Regulatory Guide 1.200, Revision 2 [A.1], "AnApproach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results forRisk-Informed Activities." The guidance in RG-1.200 indicates that the following steps shouldbe followed to perform this study:1. Identify the parts of the PRA used to support the application" SSCs, operational characteristics affected by the application and how these areimplemented in the PRA model." A definition of the acceptance criteria used for the application.2. Identify the scope of risk contributors addressed by the PRA model* If not full scope (i.e. internal and external), identify appropriate compensatorymeasures or provide bounding arguments to address the risk contributors notaddressed by the model.3. Summarize the risk assessment methodology used to assess the risk of theapplication* Include how the PRA model was modified to appropriately model the risk impact ofthe change request.4. Demonstrate the Technical Adequacy of the PRA" Identify plant changes (design or operational practices) that have been incorporatedat the site, but are not yet in the PRA model and justify why the change does notimpact the PRA results used to support the application." Document peer review findings and observations that are applicable to the parts ofthe PRA required for the application, and for those that have not yet beenaddressed justify why the significant contributors would not be impacted." Document that the parts of the PRA used in the decision are consistent withapplicable standards endorsed by the Regulatory Guide. Provide justification toshow that where specific requirements in the standard are not met, it will notunduly impact the results." Identify key assumptions and approximations relevant to the results used in thedecision-making process.Items 1 through 3 are covered in the main body of this report. The purpose of this appendix isto address the requirements identified in item 4 above. Each of these items (plant changesP0247130002-4722A-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacynot yet incorporated into the PRA model, relevant peer review findings, consistency withapplicable PRA standards and the identification of key assumptions) are discussed in thefollowing sections.The risk assessment performed for the ILRT extension request is based on the current Level 1and Level 2 PRA models of record. Information developed for the license renewal effort tosupport the Level 2 release categories is also used in this analysis supplemented by additionalcalculations to more appropriately represent the intact containment case in the ILRT extensionrisk assessment.Note that for this application, the accepted methodology involves a bounding approach toestimate the change in the LERF from extending the ILRT interval. Rather than exercising thePRA model itself, it involves the establishment of separate evaluations that are linearly relatedto the plant CDF contribution. Consequently, a reasonable representation of the plant CDFthat does not result in a LERF does not require that Capability Category II be met in everyaspect of the modeling if the Category I treatment is conservative or otherwise does notsignificantly impact the results.As further discussed below, the PRA models used for this application are the latest models,which were released in November 2011 (for IP2) and November 2012 (for IP3). There are nosignificant plant changes (design or operational practices) that have not yet been incorporatedin those PRA models.A discussion of the Entergy model update process, the peer reviews performed on the IP2 andIP3 models, the results of those peer reviews and the potential impact of peer review findingson the ILRT extension risk assessment are provided in Section A.2. Section A.3 provides anassessment of key assumptions and approximations used in this assessment and Section A.4briefly summarizes the results of the PRA technical adequacy assessment with respect to thisapplication.A.2 PRA UPDATE PROCESS AND PEER REVIEW RESULTSA.2.1 IntroductionThe Indian Point Unit 2 (IP2) and Unit 3 (IP3) Probabilistic Risk Assessment (PRA) models usedfor this application [A.2 and A.3] are the most recent evaluations of the IP2 and IP3 riskprofiles for internal event challenges. The IP2 and IP3 PRA modeling is highly detailed,including a wide variety of initiating events, modeled systems, operator actions, and commonP0247130002-4722A-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacycause failure events. The PRA model quantification process is based on the event tree andfault tree methodology, which is a well-known methodology in the industry.Entergy employs a multi-faceted approach to establishing and maintaining the technicaladequacy 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 theuse of self-assessments and independent peer reviews. The following information describesthis approach as it applies to the IP2 and IP3 PRA models.A.2.2 PRA Maintenance and UpdateThe Entergy risk management process ensures that the applicable PRA model is an accuratereflection of the as-built and as-operated plant. This process is defined in the Entergy fleetprocedure EN-DC-151, "PSA Maintenance and Update" [A.4]. This procedure delineates theresponsibilities and guidelines for updating the full power internal events PRA models at alloperating Entergy nuclear power plants. In addition, the procedure also defines the processfor implementing regularly scheduled and interim PRA model updates, and for tracking issuesidentified as potentially affecting the PRA models (e.g., due to changes in the plant, industryoperating experience, etc.). To ensure that the current PRA model remains an accuratereflection 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 PRAmodel. Potential PRA model changes resulting from these reviews are entered intothe Model Change Request (MCR) database, and a determination is made regardingthe significance of the change with respect to current PRA model." New engineering calculations and revisions to existing calculations are reviewed fortheir impact on the PRA model.* Plant specific initiating event frequencies, failure rates, and maintenanceunavailabilities are updated approximately every four years, and* Industry standards, experience, and technologies are periodically reviewed to ensurethat any changes are appropriately incorporated into the models.In addition, following each periodic PRA model update, Entergy performs a self-assessment toassure that the PRA quality and expectations for all current applications are met. The EntergyPRA maintenance and update procedure requires updating of all risk informed applications thatmay have been impacted by the update.P0247130002-4722A-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyA.2.3 Regulatory Guide 1.200 PWROG Peer Review of the IP2 and IP3 Internal EventsPRA ModelsBoth the IP2 and IP3 internal events models went through a Regulatory Guide 1.200 PWROwners Group peer review using the NEI 05-04 process.The IP2 PRA internal events model peer review was performed in December 2009, and usedthe American Society of Mechanical Engineers PRA Standard RA-Sb-2005, and RegulatoryGuide 1.200 Revision 1. The IP3 PRA internal events model peer review was performed inDecember 2010. Since the IP3 peer review was later, it used RA-Sa-2009 (the AmericanSociety of Mechanical Engineers / American Nuclear Society Combined PRA Standard) andRegulatory Guide 1.200 Revision 2. As noted in the forward to the combined standard, theprimary purpose, in addition to combining internal and external events into a single standard,was to ensure consistency in format, organization, language, and level of detail. It was alsonoted that, among the criteria observed in assembling the component Standards were:(a) the requirements in the Standards would not be revised or modified(b) no new requirements would be includedAn internal comparison of the ASME standard to the combined ASME / ANS standard confirmedthat there were few substantive changes to the internal events portion of the standard,although the expected level of documentation was increased in some cases.The IP2 and IP3 PRA peer reviews addressed 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)P0247130002-4722A-4 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyDuring the IP2 and IP3 PRA model peer reviews, the technical elements identified above wereassessed with respect to Capability Category II criteria to better focus the SupportingRequirement assessments.A.2.4 Peer Review ResultsThe ASME PRA standards used for the IP2 and IP3 peer reviews each contained a total of 326numbered supporting requirements. A number of the supporting requirements weredetermined to be not applicable to the IP2 or IP3 PRA (e.g., BWR related, multi-site related).Of the applicable supporting requirements, 95% were satisfied at Capability Category II orgreater for IP2, and 97% were satisfied at Capability Category II criteria or greater for IP3.The Facts and Observations (F&Os) for the IP2 PRA peer review are provided in the report,entitled, "RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for theIndian Point 2 Nuclear Power Plant Probabilistic Risk Assessment" [A.5]. Of the 41 Facts andObservations (F&Os) generated by the Peer Review Team, 21 were considered Findings.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 theIndian Point 3 Probabilistic Risk Assessment" [A.6]. Of the 68 Facts and Observations (F&Os)generated by the Peer Review Team, 11 were considered Findings.As a result of the Regulatory Guide 1.200 PWROG peer reviews, all the F&Os (other than bestpractices) were identified as potential improvements to the IP2 and IP3 PRA models ordocumentation and were entered into the Entergy Model Change Request (MCR) database.Tables A.2-1 and A.2-2 contain the findings resulting from the peer review of each unit, thestatus of the resolution for each finding and the potential impact of each finding on thisapplication. In summary, a majority of the findings were related to documentation and have nomaterial impact. As shown, almost all findings have been resolved and incorporated into theupdated model and/or documentation. Resolution of the few open peer review findings isexpected to have, at most, a minor impact on the model and its quantitative results and nosignificant impact on the conclusions of this application.In resolving the IP3 peer review findings, several additional internal flooding sources wereidentified as not being addressed in the original internal flooding analysis report. Most of thosesources involved fire protection piping, but they also included auxiliary component coolingwater (ACCW) piping in the fan house and short sections of component cooling water (CCW)P0247130002-4722A-5 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacypiping in a pipe chase in the foyer outside the charging pump rooms. These additional sourceswere included in the final model used for this application.A.2.5 External EventsAlthough EPRI report 1018243 [A.7] recommends a quantitative assessment of thecontribution of external events (for example, fire and seismic) where a model of sufficientquality exists, it also recognizes that the external events assessment can be taken fromexisting, previously submitted and approved analyses or another alternate method ofassessing an order of magnitude estimate for contribution of the external event to the impactof the changed interval. Since the most current external events models for IP2 and IP3 arethose embodied in the IPEEE, a multiplier was applied to the internal events results based onthe IPEEE, similar to that used in the SAMA analysis [A.8 and A.9]. This is further discussed inSection 5.7 of the risk assessment.A.2.6 SummaryThe IP2 and IP3 PRA technical capability evaluations and the maintenance and updateprocesses described above provide a robust basis for concluding that these PRA models aresuitable for use in the risk-informed process used for this application.P0247130002-4722A-6 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-3 Appendix Al, Section 3.4, "Other IE-A8 Appendix Al, Section 3.4, Document the interviews OPEN No ImpactInitiating Events" states 'Other "Other Initiating Events" states This is a documentation issue. Although This is aplant-specific initiators and event 'Other plant-specific initiators discussions were held with plant personnel, no documentationprecursors were also investigated and event precursors were formal interview form or format was used. This enhancement issue.using an FMEA of plant systems as also Investigated using an remains open as a documentation improvementdiscussed below and this was FMEA of plant systems as item for the next update.reviewed with plant personnel to discussed below and this wasverify expected plant response.' It reviewed with plant personnelis not clear that interviews were to verify expected plantconducted, response.' It is not clear thatinterviews were conducted.1-7 Not met since the frequencies were IE-C5 The SR requires that the IE Weight the initiating event OPEN No significantnot weighted by the fraction of frequencies be weighted by frequency time by the While we agree that the wording in the SR itself impacttime the plant was at power. the plant availability. This has fraction of time the plant indicates that weighting should be done, the The current approachnot been done for IP2 initiating was at power. ASME standard acknowledges that the SR provides a slightlyevents, wording is somewhat unclear and provides a conservative result,detailed note of explanation (Note 1 of the and use of theSR). Entergy believes that using the annual stipulated weightingaverage model, which Note 1 acknowledges approach would haveshould not include the weighting factors, is the no significant impactappropriate baseline model in the absence of an on this application.all modes model. We do agree, as the standardstates, that an all modes model should accountfor the time in each operating state. Entergydoes not have an all modes model at this time.We believe that tying risk values to plantavailability without an all modes model canpotentially provide inappropriate risk insights tonon-PSA personnel. It does not apply any risk toother operating states. Therefore, we believethat at the least, our current model meets theSR, when taken in concert with the associatedNote 1.P0247130002-4722 A-7 Risk Impact Assessment of Extending the Indian Point tLRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTnON ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-8 While the documentation of the SC-Cl The current documentation Provide basis for Resolved No ImpactSuccess Criteria is detaiied with poses a potential problem in parameters, limits, Additional references/basis for parameters, Documentation issuesufficient information to support facilitating PRA applications, setpoints, etc. limits, and setpoints were added to Section -incorporated in finalthe model development, the lack of upgrades, and peer review due 01.3.2, "Level 1 Assumptions" and other project file for thereferences to supporting to the significant amount of pertinent sections of the success criteria analysis model used for thisdocuments for a variety of information included that is notebook, application.assumptions and sections makes not traceable.the review difficult and the abilityto maintain the model based uponplant changes and analysisrevisions very difficult to track andchange.Examples are:1) RCS peak pressure within 120seconds of an ATWS2) The normal relief flow througheach PORV valve is 179,000 lb/hr;the maximum flow is 210,000 lb/hrNote that these are simply a coupleof examples of a more prevalentissue.1-t1 Attachment E summarizes the tE-C4 Attachment E summarizes the Produce a table which Resolved No Impactcalculation of initiating event IE-C5 calculation of initiating event shows the actual Added a table showing a sample calculation to Documentation issuefrequencies but there must be a frequencies but there must be calculations using generic, enhance Appendix At of the update report. The -incorporated in finaltable that shows the actual a table that shows the actual plant-specific, and calculations used to develop the IE frequencies project file for thecalculations using generic, plant- calculations using generic, Bayesian updating are contained in the EXCEL files that are part of model used for thisspecific, and Bayesian updating. It plant-specific, and Bayesian the IP2 model update project files and are application.would be helpful to include this updating. It would be helpful retained for future reviews, updates ortable, to include this table, applications. This issue is only a matter of theextent and the details of the calculationsextracted and made part of the written report.Also note that the methodology used for thesecalculations was discussed in Appendix At,Section 11 and the results were summarized inAttachment E.P0247130002-4722ý_a Risk Impact Assessment of Extending the Indian Point ILRT intervalsAppendi, A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-13 No definition or criteria for the DA-A2 The criteria to establish the Provide documentation Resolved No Impactdefinitions of failure modes, and definitions of SSC boundaries, regarding the failure This is a documentation issue. The Current Documentation issuesuccess criteria were identified in failure modes, and success modes to consider for process satisfies the requirements of this SR. The -incorporated in finalthe review of the Data analysis criteria In a manner consistent evaluation of the data boundaries, failure modes, and success criteria project file for thepackage. with corresponding basic event analysis and the associated considered in the Data Analysis are consistent model used for thisdefinitions in Systems Analysis success criteria. (It Is with those used for each system to match the application.are required per the SR. In noted that Attachment 2 of failure modes, common cause and boundaries ofthis case SSC boundaries were Appendix DO, identifies unavailability events. The data analysis notebookdiscussed and examples many of the issues for discusses this (for example, see Appendix D1,provided. However, there was consideration in relation to sections 1.4 and 3.1 thru 3.3 and 4.1, 4.3 andno similar documentation for this SR.) 4.6) and shows that these are all addressed inthe failure modes and success the updated plant model. App. D1, Attachment Acriteria includes discussions and definitions of componentboundaries related to component failure modesand how this was considered in the data analysis.This is consistent with Appendix E, Table E0.1-3which lists the failure modes and associatedcodes that are used in the model. All modeledbasic events are captured in the fault trees andthe associated model data base with codescorresponding to this table and the Data Analysisis shown to match the failure modes andboundaries of these events. In the associatedSystem Notebook, each fault tree is discussedand the overall system success criteria In themodel are summarized.1-14 Accident sequences that reach and AS-A8 DEFINE the end state of the Rewrite the statement to Resolved No Impactremain in a stable state for 24 accident progression as indicate that the accident The statement referred to in the finding, which Documentation issuehours are assumed to be occurring when either a core sequence is mitigated exists in Section 4 of the main report and in -incorporated in finalsuccessfully mitigated. This can be damage state or a steady state when a stable state without Appendix F1.0, has been revised to read: project file for theinterpreted to mean that the condition has been reached core damage has been model used for thismission time is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reached. The mission time "Accident sequences that reach a stable state application.reaching a stable state. This for this is usually 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and remain in that state for thestatement should indicate that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time after the initiating eventaccident sequence is considered are assumed to be successfully mitigated. It Ismitigated when a stable state assumed that sufficient additional resources existwithout core damage is reached. and sufficient time is available by that time torespond to any additional challenges."Ptla7130002-4722 9~

Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-16 SR is MET, however, three system SY-B8 Walkdowns were documented Provide conclusion of Resolved No Impactpackages in which the section as required for this SR. walkdown in all systems The walhdown records for the systems noted in Documentation issuerelating to spatial dependencies However, this Is a packages. the finding (Control Building HVAC, Primary -incorporated in fSnalhad no conclusion as to whether a documentation issue. Water and AFW Building Ventilation systems) project file for thespatial dependency exists (e.g. have been reviewed and no spatial dependencies model used for thisControl Building HVAC, Primary have been Identified. The conclusion has been application.Water, AFWP Building Ventilation) added to each of those system notebooks underSection 1.5 "LOCATION AND SPATIALDEPENDENCIES". The remaining systemnotebooks already contain this conclusion.1-18 Not Met CC II/III due to the lack of DA-D4 A review of the Update Evaluate the posterior data Resolved No Impactdiscussion and documentation Spreadsheet in support of the in relation to the Revised App. Dt and Data Analysis spreadsheet No change wasrelating to examination of Bayesian analysis reflects a uncertainty bounds of the to follow the same approach used for IP3 and required to theinconsistencies between the prior single failure in which the posterior and prior clarify that the requirement in SR DA-D4 to posterior data set.distribution and the plant-specific posterior mean fell outside the uncertainties to address "check that the posterior distribution isevidence to confirm that they are uncertainty bound of the prior discrepancies and reasonable given the relative weight of evidenceappropriate distribution. document the issue such provided by the prior and the plant-specific data"that the discrepancies (if was performed. The discrepancies between thethey exist) can be generic and the updated means were identifiedexplained or resolved, and evaluated and all were found to bereasonable based on the nature of the Bayesianupdate algorithm, the number of failures and theavailable plant data. Appendix D1, Section 3.6was revised to discuss the approach. Thesestatistical tests satisfy the requirements of DA-D4.1-19 There is no evidence that HR-C2 INCLUDE those modes of Analyze miscalihbration of Resolved No Impactmiscalibration of equipment that unavailability that, following equipment that provided Comment incorporated. Additional pre-initiator Change incorporatedprovided initiation signals for completion of each unscreened initiation signals for hunman failure events (HFEs) were added to the in model used for thisstandby pumps were analyzed. activity, result from failure to standby pumps. model to represent miscalibration errors. See application.restore (b) initiation signal or SAS system notebook, Table 1.2 Pre-tnitiatorSection Ht.0 states: 'This review set point for equipment start- Human Failure Events (HFEs) Screening.did not identify any Human Failure up or realignmentEvents (HFEs) that are not alreadyaccounted for as possible failuremodes in the Human Reliabilityanalysis (HRA).'P0247130002-47122utA-10 Risk Impact AsssseSn et of Extending tire Indian P01W ILRT IntervalsAppendhix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-20 A review of the CCF in the System SY-B4 Naming convention should Correct the naming Resolved No ImpactWork Packages (i.e. AFW) reveals match in all references. This convention in the System The common cause basic event names in the Documentation issuethat the Common Cause names issue does not affect results Packages to match the AFW System Work Packages have been corrected -incorporated in finallisted do not match the common since the model names and model. and now match the basic event names used in project file for thecause names in the model and data the data analysis names are the AFW system fault tree model and data model used for thisanalysis package. consistent. analysis, application.(Example: FW406, FW-CCFS-AFWPM, etc.)1-23 In the Scope of Analysis it is IFSO-A4 For each potential source of Include maintenance Resolved No Impactstated: 'In this analysis, all causes flooding, IDENTIFY the induced flooding in the A search of the IP2 condition reporting system No changes to theof flooding were considered except flooding mechanisms that flood initiator frequencies was performed for a period of 15 years for the flooding frequencyplant-specific maintenance would result in a release. Internal Flooding Analysis. No significant Internal values were required.activities-the contribution of INCLUDE: .flooding events (including maintenance Induced),normal maintenance to flooding is (a) Failure modes of were identified which would significantly alter theincluded in the rupture frequency components such as pipes, generic data.data used.' The flood frequencies in tanks, gaskets, enpansionthe EPRI flood guideline do not joints, fittings, seals, etc.include maintenance. (b) Human-inducedmechanisms that could lead tooverfilling tanks, diversion offlow-through openings\created to performmaintenance; inadvertentactuation of fire-suppressionsystem0c) Other eventsresulting In a release into theflood areaFnlu7t3nnIl.a722 u-tiP0247130002-4722A-11 Risk Impnact Assessment of Extending the Inidan Point ILRT Inte-Is~Appendix A P5.A 7ecthsaI AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-24 IDENTIFY the characteristic of IFSO-A5 There is no documentation Identify the pressure and Open No Impactrelease and the capacity of the IFSO-A6 that identifies the pressure temperature of the source. This is a documentation issue. While Appendix C This is asource. INCLUDE the pressure and and temperature of the source, does not specifically identify the pressure and documentation issue.temperature of the source. temperature of the sources, the analysis did The description indocument that the maximum flow rate resulting Appendix C will befrom a guillotine rupture was determined as well enhanced during theas lesser calculated release rates. A range of next update.release sizes consistent with the available EPRIpipe rupture frequency data were, in fact,considered and a flow rate and frequency ofoccurrence derived for each. By this means, thesize and frequency of possible releases werematched as required for the quantitativedetermination of the consequences of internalflooding. This remains an open finding, pendingenhancement of the documentation regarding thepressures and temperatures of the rupturedsystems to meet the letter of the SR.P0247130002-4722A-12 Risk Impact Assessment of E'tending the Indian Point ILRT interoaisAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-26 Capability categories met. Latest DA-Ct It would be helpful to indicate Provide documentation Resolved No significantversions of recognized generic data instances in which the generic regarding the failure Appendix D1 was revised to clarify that any impactsources were used. Generic data data and the model do not modes to consider for mismatches are due to discrepancies in the At most, this mayfor unavailability were not used. match. As currently evaluation of the data generic data sources. Added the following result in a slightdocumented, it is not clear analysis and the associated wording to section 1.4 to address boundaries and conservatism asNote: The analysts ensured, to the how often this occurs or how success criteria. (It is other Issues; "Consistent with System Analysis noted in theextent possible, that the parameter significant mismatches of this noted that Attachment 2 of requirements, the failure rates, common cause disposition.definitions and component type might be. Note: the EDG Appendix DO, identifies failure events and unavailability events wereboundaries were consistent load output breakers are many of the issues for identified from the system fault trees to bebetween the model and the data identified specifically in the consideration in relation to consistent with corresponding systems analysissource. Appendix D notes that text as being one area of this SR.) definitions, success criteria and boundaries (tomismatches may be present, but mismatch. If this is the only the extent practical considering the differences inthat any such Instances would be instance, then this should be the boundary definitions in the generic andconservative because the generic clarified. common cause databases). Component failuredata would include subcomponents data was matched to corresponding events inthat are treated separately in the system fault trees. Failure modes that are in themodel. system models were mapped to correspondingbasic event Type Codes and other events used inNote: The opening paragraph In CAFTA (common cause failure and maintenanceAttachment 0 indicates: 'The unavailability events)." Also revised Attachmentboundary definitions used in the A, section 1.0, item 2 to add; "Note that themodel may need to be modified boundaries provided below are consistent withdepending on the generic database those used in NUREG/CR-6928, however they areand should be clearly defined so not defined in the same manner or to the samethat the failure modes in the model level of detail as they are in the NRC CCFmatch those in the generic database which may result in overlaps in thedatabases.' Apparently, this was boundaries that could lead to conservativenot done in all cases -as noted estimates for the CCF failures". No additionalabove. documentation or evaluation of the data analysisis required to satisfy this requirement.roianisnoti.t,22 u-tIP0247130002-4722A-13 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-27 Met for CCI but not CCII; Section DA-C13 Appendix D1, section 3.7 says Document the interviews Resolved No Impact3.7 System Unavailability Due to 'If no Maintenance Rule or used to meet this As demonstrated in the EXCEL file used for the This was aTesting and Maintenance discusses plant records were available requirement data update, the population of components for documentation issuethat 5 years of unavailability data for a particular component, which Maintenance Rule (MR) unavailability data since there were nowas collected via the Maintenance generic data from NUREG/CR- did not exist was limited to the Appendix R Diesel additional insightsRule program. If no Maintenance 6928 were used to estimate Generator and a few MR non-risk significant available from plantRule or plant records were unavailability.' systems. The Appendix R diesel has only been in personnel.available for a particular service a limited time and the System Engineercomponent, generic data from confirmed that there were no unavailable hoursNUREG/CR-6928 were used to that could be applied for the update. Theestimate unavailability. Maintenance Rule Coordinator and/or theappropriate System Engineers were queriedregarding the other systems for which MRavailability was not monitored but were unable toprovide reliable estimates due to the lack ofmonitoring data. As a result, generic data wasapplied to these system components.Since the discussions with plant personnel did notyield useful information and could not be used "togenerate estimates" for unavailability, additionaldocumentation of those discussions would be oflittle additional value and was not generated.2-2 .Capability Category I met. DA-C1O Discussion in Appendix D was Add discussion to further Resolved No ImpactDocumentation in Appendix Dl was not explicit enough to know explain whether this SR Appendix Dl, Section 3.4 was enhanced to clarify Documentation issuenot sufficient to determine if It was whether Cat II was met. was met at Cat I1. that failure modes were not decomposed into -incorporated in finalnecessary to decompose sub-elements. Therefore, Appendix D does not project file for thesurveillance test data Into sub- address decomposition of failure modes and it model used for thiselements and whether this was was not necessary to perform additional reviews application.done. of surveillance tests to address sub-elementspecific data.ro 247 1 3000 2.47 22 u-anP02,17130002-4722A-14 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION3-2 Each System Notebook contains SY-B14 TAKE CREDIT for system or Provide analysis that the Resolved No ImpactTable B-2a Supporting SY-A22 component operability only if equipment can function The model only takes credit for component Documentation issueRequirements for HLR-SY-A that an analysis exists to beyond design basis operability based on design or rated capability -incorporated in finalstates under SY-A20 something demonstrate that rated or environment. and does not assume or take credit for operation project file for the-such as this for CCW: 'The design capabilities are not beyond design basis capability unless specific model used for thisComponent Cooling Water System exceeded. calculations and evaluations were available, as application.by its design function removes heat noted in the system notebooks for AFV, CBfrom containment. Therefore, the HVAC, EDGV. Clarification was provided in theComponent Cooling Water System system notebooks, as required, to revisedis fully capable of providing heat wording of "Harsh Environments" under sectionremoval. Therefore, no further t.S and in Table B-2a for how SY-A20 is met (seeanalysis is required to support this the other various system notebooks includingfunction.' CCW, CVCS, HHSI, LHSI, IAS, EDG, SWS).However it is not clear thatanalyses were done to take creditfor equipment associated withrecirc inside containment.3-4 There Is no problem with the DA-D1 Issue centers on the Calculate realistic Resolved No Impactgeneric data or the Bayesian calculation of 'realistic parameter estimates using Revised failure identification to include plant No changes to theupdating process used. The issue parameter estimates' using plant specific data. failures not included in EPIX data as explained in data analysis wereis the calculation of 'realistic plant specific data since only revised Appendix D1, Section 3.5. Entergy fleet required.parameter estimates' using plant EPIX / Maintenance Rule procedures and fleet standards address EPIXspecific data since only EPIX / information was used. reporting and confirm that all Maintenance RuleMaintenance Rule information was (MR) functional failures require an EPIX report.used. They also require all failures of high criticalcomponents to be included in EPIX reporting,which includes failures that may cause a trip orimpact plant operation, even of non-risksignificant operating systems within MR scopethat might be monitored under plant criteria andmight not otherwise be captured. Theserequirements ensure that failures of all modeledcomponents are captured in the EPIX data usedfor the PSA model. The only exceptions arefailures of high critical components that occurredprior to 2007, when these procedures wereimplemented. Those failures were obtained fromspecific plant records and included in the update.No further action is required to satisfy thisrequirement.P024713tOD-4722 -tA-15 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION4-1 Met CC It/ItI Based upon a DA-D4 A review of the Update Evaluate the posterior data Resolved No Impactthorough analysis of the generic Spreadsheet in support of the in relation to the The associated data analysis spreadsheet was No change wasdata using plant specific data for Bayesian analysis reflects a uncertainty bounds of the ievised to allow the discrepancies between the required to theBayesian updating. However, there single failure in which the posterior and prior generic and the updated means to be identified posterior data set.is a lack of discussion and posterior mean fell outside the uncertainties to address and evaluated and all were found to bedocumentation relating to uncertainty bound of the prior discrepancies and reasonable based on the nature of the Bayesianexamination of inconsistencies distribution, document the Issue such update algorithm, the number of failures and thebetween the prior distribution and that the discrepancies can available plant data. Appendix D1, Section 3.6the plant-specific evidence to be explained or resolved. revised to clarify that the requirement in SR DA-confirm these inconsistencies are D4 to "check that the posterior distribution isappropriate reasonable given the relative weight of evidenceprovided by the prior and the plant-specific data"was performed. These statistical tests satisfy therequirements of DA-D4.4-2 This SR is Not MET. The use of DA-D1 It is not apparent that all plant Perform a more extensive Resolved No ImpactEPIX as the basis for plant related DA-D4 specific failures associated review of the plant specific See disposition for finding 3-4. No changes to thefailures associated with PRA with PRA related components failures to ensure that the data analysis weremodeled components is insufficient have been captured in the data is complete. (Note: required.to ensure that all failures are data review for this model should it be determinedcaptured. EPIX captures update. that the Indian Point EPIXMaintenance Rule Functional database does actuallyFailures and Critical component Include all PRA modeledfailures (post 2007). Therefore, component failures, thisthis database is limited in scope. FAO can be dispositionedas such).Also it should be considered thatthe Maintenance Rule will notcapture all failures associated withnon-risk significant systems.Therefore, this data is also notinclusive.4-3 Documentation of the data analysis DA-El Supporting files were provided Incorporate the Resolved No ImpactIs not complete due to the lack of during the review that spreadsheet into the Revised Appendix D1, Section 3.6 to include Documentation issueany reference to the basis for the contained critical information document or as a reference reference to the applicable spreadsheets along -incorporated in finaldata results. It was noted during relating to the data analysis. in order to ensure with discussion of how they are the basis for the project file for thethe review that the data analysis is This Information in the form of traceability of the analysis results. The spreadsheets are also retained in the model used for thisactually calculated using an Excel Spreadsheet is not and inputs for the analysis. project files that are maintained available for PRA application.spreadsheets; however, those Included in the Data Analysis Also include guidance on applications, upgrades, and future reviews. Anspreadsheets are not part of the package and is not referenced the use of the information example of the calculations in the Excel fies wasdata analysis package. by the package. contained In within the added to Appendix D. No further action isspreadsheet, required to satisfy the requirements of this SR.P0247130002-4722A-16 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendi.x A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION4-4 The model uses a single value for QU-B1O The modularnzation of RPS in Incorporate the RPS fault Resolved No ImpactRPS in relation to the ATWS tree the ATWS logic precludes the tree system into the ATWS The RPS is a somewhat unique system, and while Use of a single valueand certain initiating Events. This ability for risk significant logic in a manner that we agree that the modeling of RPS Is not fully for RPS unavailabilityRPS module for the ATWS logic is determinations of support allows results consistent with this SR, we disagree that this has no Impact on thisquantified using the RPS fault tree. systems and components interpretation of individual finding warrants the SR not being met. In application.Although modularization of within RPS. events, particular, the RPS is a fail-safe system. As such,initiating events allows for the loss of a support system does not materiallydetermination of risk significance of impact the reliability of the RPS. Although thethe Initiator, the use of this module shunt trip function does rely on 125V dc power,restricts the usability of the model the increase In unreliability of the RPS associatedfor risk significance determination with unavailability of dc power is negligible. Infor those components associated addition, regarding the modeling of transmitterswith RPS. and trip relays, it should be noted that the RPSfault tree, which is consistent with NUREG/CR-5500 (Volume 2), Is conservative in that it onlycredits two trip signals (overpower delta T andpressurizer high pressure). tndividual testsimpacting the RPS are addressed for onlinemaintenance by adjusting the top event for RPSunreliahility accordingly. Furthermore, thelimited applicability of the Finding should notpreclude the SR from being met.PtlC7t3000Z.47Z2 u-tnP0247130002-4722A 17 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC, BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-11 Appendix C1 of IP-RPT-I0-00023, IFSN-BI Analysis details available to Provide required Resolved No impactRev. A provides a high to medium IFSN-Bi2 the peer review team such as documentation The backup spreadsheets have been obtained as The backup requiredlevel summary of the flood flooding calculations, were not well as the software used for flood level to support futurescenarios, and provides greater sufficient to support upgrades calculations, instructions for use of this software model updates anddepth in some areas. Analysis and would have to be obtained and material that supports its application. This applications is now Indetails available to the peer review or reproduced for future model additional documentation was included in the the projectteam such as flooding calculations, changes. The documentation final model documentation package. Initiator documentation.were not sufficient to support also lacks in reference to specific flag files are contained in the electronicupgrades and would have to be quantification input files Included in the model update documentationobtained or reproduced for future documentation (initiator package. A list of flag files was also added to themodel changes. The documentation specific flag files) internal flooding notebook.also lacks in reference toquantification input documentation(initiator specific flag files)1-12 The walkdown notes in Appendix A tFSN-A5 There is no specific physical For SSCs susceptible to Resolved No impactof IP-RPT-10-00023, Rev. 0, location information found in spray failure (also see FAO Additional discussion was added to the walkdown AdditionalAppendix C.A note the general the documentation for SSCs 2-3), ensure sufficient Appendix to support the spray impacts included information has beenlocation of each SSC with respect to other than flood area and relational location in the model. This includes reference to included in theIts room and elevation as well as its elevation. Therefore, it cannot information between the environmental qualification documents where updated modelsubmergence height. Some be determined which SSCs in target SSC and spray these were used as a basis for stating that documentation.additional general locational any area are susceptible to sources are provided so equipment would not be vulnerable to sprayinformation is sometimes identified spray from any specific spray that a determination can be damage. A conservative separation criterion ofin Section 4.2 of IP-RPT-10-h0f23, source. In the scenario made as to whether the 30 feet was used in examining the potential forRev. 0, Appendix C.t. For example, development it identifies SSCs can be damaged by spray impacts in the analysis. The compositeit may state that a flood source may which equipment is impacted each potential spray piping and general arrangement drawings wereimpact one but not both trains of by spray, but it cannot be source, scrutinized to ascertain whether equipment couldequipment; specifics are not given determined how that be sprayed should a line or other piece ofas to why both cannot be impacted information was obtained or if equipment rupture. The text of the report has(e.g., shielding, curbs, etc.), but the It is correct, been changed to note this. Providing additionalinformation implies the impact of specific location information within the modelspatial information. documentation will be considered to supportfuture updates but is considered a documentationThere is no specific physical location enhancement issue with no expected impact onInformation related to spray type the analysis.failures found in the documentation.SSCs are only identified locationallyby their flood area and elevation. Itcannot be determined which SSCsin any area are susceptible to sprayfrom any specific spray source.P00247 13000 2.47 22A-18 Risk Impact Assessment of Estending the Indian Point ILRT IntervalsAppendix A PRA Techncal AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-15 The initiating event frequencies are IE-C5 The initiating event Include the plant Open No significantnot weighted by the fraction of time frequencies are not weighted availability factor in the While we agree that the wording in the SR itself impactthe plant is at power, by the fraction of time the calculation of initiating indicates that weighting should be done, the The current approachplant is at power. event frequencies. ASME standard acknowledges that the SR provides, at most, aSection 10.9 of Appendix AO wording is somewhat unclear and provides a slightly conservativeprovides guidance to account for detailed note of explanation (Note 1 of the result in comparisonplant availability in initiating event SR). Entergy believes that using the annual to use of thecalculations. Section 4.0 of average model, which Note 1 acknowledges stipulated weightingAppendix At states that the should not include the weighting factors, is the approach and wouldavailability factor for the data appropriate baseline model in the absence of an have no significantupdate period was calculated, all modes model. We do agree, as the standard impact on thisHowever, the calculated value is not states, that an all modes model should account application.incorporated into the initiating event for the time in each operating state. Entergyor final CDF results, does not have an all modes model at this time.We believe that tying risk values to plantavailability without an all modes model canpotentially provide inappropriate risk insights tonon-PSA personnel. It does not apply any risk toother operating states. Therefore, we believethat at the least, our current model meets theSR, when taken in concert with the associatedNote 1.3-7 The effects of the flood on PSFs IFQU-A6 Limited flooding-related Discuss flood effects on Resolved No impactwere not specifically addressed in human actions are included in PSFs and make No short term isolation actions were credited in As discussed in thethe HRA analysis. the HRA discussion in adjustments to the HRA the flooding analysis. The only significant field disposition, the onlyAppendix H, but there is no analysis if needed, action credited in the internal events model that potential for amention of any effects of the could be impacted by the plant conditions flooding impact onflood on PSFs. associated with flooding was alignment of the modeledalternate cooling to the charging pumps on loss operator actions hasof CCW for certain specific CCW failure locations, been addressed inThe model has been updated to address that the updated modelconcern, and assumes that operator action is used for thisprecluded by a break in the location that would application.impact that action.P0207130002-4722A-19 Risk impact Assessment of Eyctending the Indian Point ILRT IntervalsAppendi, A PRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION4-14 Failure modes and success criteria DA-A2 Based on the documents As described in Sections Resolved No significantdefined in Systems Analysis are DA-D6 reviewed and the Issues 5.10 and 6.3.11 of This was a documentation issue. The level of impactconsistent with the Data Analysis. identified, component Appendix DO, assure modeling in the IP3 update required use of As noted, anyThis SR also asks for establishing boundaries are not consistent component boundaries various databases since not all databases differences Inconsistent SSC boundaries between among failure rate, CCF and defined in failure rate and provided data for the components included in the boundary definitionsthe system level analysis and the unavailability data. Plant- CCF data match the PSA model. tn some cases, the databases do not have would at most resultdata analysis, specific features need to be model. Assure the sufficient information to clearly delineate the in a very minorReviewed Appendix E6 and E27 of considered for boundary boundaries used In the test applicable boundaries. The system models and conservatism andthe systems notebooks and definitions. and maintenance data is generic databases were reviewed to confirm that would have noAppendix D for the Data Analysis. It is possible to ensure that consistent with the PSA either there was agreement between the model significant impact onBelow is a list of issues identified: the inconsistent boundary model. Make adjustments and generic database boundaries, or component this application.1. System notebooks do not define definitions result in or provide justification for boundaries In the current model conservativelythe component boundaries. The conservative results, but any mismatch identified. overlap the boundaries shown in the genericcomponent boundaries are defined realistic rather than Review plant-specific CCF databases used for the update. The failure ratesby the generic failure rate data conservative results is Ideal. experience for consistency for these additional components were found to besource with limited discussions on CCF events tend to dominate to meet SY DA-D6 small and inclusion in the model results in, atplant-specific SSC features and system level cutsets and requirements, most, a very minor conservatism in the results.modeling considerations, conservative CCF basic event The model documentation was enhanced to2. The guidance document Appendix values may mask other provide additional detail to clarify the issues withDO Section 5. ce states 'Assure the important components in a the generic database boundaries and the slightlycomponent boundaries established system. conservative modeling approach.in the generic data match thosedefined in the PSA model. Make Regarding the example given of the batteryadjustments or justify differences', chargers, the input and output breakers areAlso, Attachment 4, Section 3.0 of included In the generic database boundarythe same document states that CCF definition for common cause failures whereas theboundaries are dictated by the fault input breakers are not clearly identified to betree modeling. However, the included In the generic independent failure rate.component boundaries defined for The PSA model does not include common causefailure rate and CCF data do not failure of the input or output breakers. Thematch. The justification for using model does conservatively include independentthe data that way is that it is the failure of the input breakers due to specificconservative to do so. It Is true that modeling considerations. This approach isthis approach is conservative for considered appropriate to satisfy the SREmergency Diesel Generators, but it requirement.may not be conservative for othercases like batteries and batterychargers where CCF of outputbreakers are not modeled.P024713aa02-4722A-20 Risk Impact Assessment of Extending the Indlan Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION4-14 (continued) Regarding the test and maintenance boundaries,3. Sections 1.2 and 1.4 of Appendix the tP3 Maintenance Rule Basis documents forDI state that the data analysis each system, which define the functions thepackage is consistent with the system must meet and the interfacing boundariessystem analysis. However, as between systems, were compared to thediscussed in Item number 1 above, maintenance unavailability terms in the updatedsystems analysis only defines the model. The system functions are consistent withsystem boundary and not the the system models. The unavailable hourscomponent boundaries within the monitored under the Maintenance Rule weresystem. assigned to the same major components in the4. Boundaries of the test and model so that the model boundaries agree withmaintenance unavailability events or conservatively overlap the maintenanceare not specifically discussed, but unavailability boundaries.seem to be same as the boundariesfor the failure rates. Data from theMaintenance Rule program is usedfor this case, but It is not clear if thesystem and component boundariesconsidered In this program isconsistent with the PSA modelboundaries. Section 6.3.11 ofAppendix DO discusses this issue,but there Is no evidence that theanalysis done In Appendix Dlconsidered boundaries applies toroutine test and maintenancepractices at IP3.POZO,1301i02-tZ2 u.2P0247130D02-4722A-22 Risk Impact Assessnent of Extending the Indian Point ILRT IntervalsAppendix A PPA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION6-i The justification/statement that the SC-Bi The justification/ statement Perform rigorous Resolved No impactCST inventory is sufficient for AFW SY-B11 that the CST inventory is evaluation/justification of Plant design documentation supports the 24 Documentation issuefor 24 hrs should be enhanced, sufficient for AFW for 24 hrs the CST inventory to mission time for the CST. The Appendix B write- -incorporated inshould be enhanced. IP-RPT- support 24-hour AFW up was revised to reference a June 2004 final project file for10-00023, Rev. 0, Appendix B, operation. Westinghouse calculation in support of IP3 power the model used forSection B1.3.1.3.2 states early uprate project. The results of this calculation this application.that CST inventory is sufficient (along with initial calculation boundaryfor 24 hrm while later reveals conditions) are used to document adequate CSTthat the MAAP analysis shows water inventory supply to support AFW operationinsufficient CST inventory with for secondary-side cooling for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Instatement that alignment to addition, as noted, CST inventory is typicallythe city water supply may be maintained above the minimum inventory level,required. An informal providing additional margin. Final modelcalculation with the minimum documentation was modified to remove theflow requirement in EOP apparent discrepancies.concludes that "it would seemthat there is enough inventoryin the CST to allow the AFWsystem to operate for 24hours". Then in IP-RPT-10-0023, Section Insights statesthat 'As the normal CSTinventory is sufficient tosupply the AFW pumps for the24-hour mission time in thePSA', no credit is taken for thealternate suction path fromcity water supply.P0247130002-4722A-22 Risk Impact Assessment of ES-tendlng the Indian Point ILP7 IntervalsAppendix A PPRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION6-6 Supporting requirement IFSO-A4 is IFSO-A4 This supporting requirement is Identify the flooding Resolved No impactintentionally not met as stated In intentionally not met as mechanisms that would The intent of the statement in the report was to As noted in thetP-RPT-10-00023, Rev. 0, Appendix stated in IP-RPT-10-OOO23, result in a release for each acknowledge that the EPRI data used for the disposition, plantC1, Section 3.3: 'The one Rev. 0, Appendix CI, Section potential source of flooding analysis included all rupture mechanisms that specific conditionsupporting requirement of the ASME 3.3: 'The one supporting to meet the SR. contribute to piping system failures and to note reports werestandard that we have made no requirement of the ASME there are no readily available data that would reviewed forattempt to meet is IF-B2: "for each standard that we have made allow us to distinguish between different release applicable eventspotential source of flooding, identify no attempt to meet is IF-B2: mechanisms. The identification of specific causes involving humanthe mechanisms that would result in "for each potential source of of failure is therefore a documentation issue. The induced floodinga flooding release". In this analysis, flooding, identify the only contributor not included in the EPRI data is events, which wereno distinction was made between mechanisms that would result human induced flooding events. Since no the only events notthe various causes of floods because in a flooding release". In this applicable generic data exists related to human covered by the EPRIthe rupture frequencies used analysis, no distinction was Induced events, plant specific condition reports data. No suchincluded all floods." made between the various were reviewed for applicable events (none were events were foundcauses of floods because the identified) and discussions were held with plant and the frequenciesrupture frequencies used operations personnel. Based on those used remain valid.included all floods." discussions, activities that could challenge The modelsystem integrity such as large scale movements documentation hasof water and plant modifications are typically been modified toperformed during outages and would not specifically discussconstitute significant contributors to flooding risk. both failureNonetheless, the model documentation has been mechanisms and themodified to specifically discuss both failure conclusions of thesemechanisms and the conclusions of these human human inducedinduced failure evaluations. failure evaluations.6-7 As stated in IP-RPT-10-'OO23, Rev. tFSO-A5 As stated in IP-RPT-10-'OO23, Identify the characteristic Resolved No impact0, Appendix C1, Table 3.3.1.1 for Rev. 0, Appendix Cl, Table of release for each source We consider this a documentation issue. While Documentation issueIFSO-A5, maximum flow rate 3.3.1.1 for IFSO-AS, and its identified failure the table mentioned in the finding did state that a -incorporated Inresulting from a guillotine rupture is maximum flow rate resulting mechanism. maximum flow rate resulting from a guillotine final project file fordetermined and used, instead of from a guillotine rupture is rupture was determined, it also noted that the the model used foridentifying the characteristic of determined and used, instead frequency of this and lesser releases were this application.release for different failure of identifying the characteristic calculated. A range of release sizes consistentmechanism, of release for different failure with the available EPRI pipe rupture frequencymechanism. This is in contrary data were, in fact, considered and a flow rate andto the SR. frequency of occurrence derived for each. By thismeans, the size and frequency of possiblereleases were matched as required for thequantitative determination of the consequencesof Internal flooding. The text in the report hasbeen modified to clarify this matter. Additionalinformation regarding the pressures andtemperatures of the ruptured systems has alsoSeen added to the documentation.P0247130002-4722A-23 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Techncal AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION6-8 IP-RPT-1l-00023, Rev. 0, Appendix IFSO-Al IP-RPT-10-00023, Rev. 0, Identify the potential Resolved No impactCt, Section 4.1.3 states that the IFSO-BS Appendix C1, Section 4.1.3 sources of flooding for each All accessible flood areas were included In the Since as noted in thepotential flood sources were IFSO12 states that the potential flood flood area per the plant walkdowns. Appendix A has been revised disposition, all areasidentified by walkdowns and the sources were identified by standard. to include the areas that were previously omitted were, in fact, walkedexamination of drawings, and listed IFSO-A3 walkdowns and the Perform and document from the documentation, including those areas down, this was ain Appendix A, Plant Walkdown. IFSO-A6 examination of drawings, and walkdowns for missed flood mentioned in the finding, documentation issueHowever, Appendix A does not listed in Appendix A, Plant areas. If these areas The statement in the introduction to the and wasprovide adequate information on Walkdown. However, Appendix cannot be walked down for walkdown notes was intended only to Incorporated in finalflood source as (1) some flood areas A does not provide adequate operational or health acknowledge that there might be small bore, field project file for theare not included in the walkdown information on flood source as reasons, other methods of run piping (less than 1 inch diameter) that were model used for thissuch as 3PAB41-1A,43-60A, 46- (1) some flood areas are not obtaining this data (e.g., not shown on system drawings and would not application.73A,55-63A, 3FH72-B, 3FH80-A, included in the walkdown such plant drawings, operator have been confirmed by the waikdown. Suchetc.; (2) Appendix A has stressed as 3PAB41-1A,43-60A, 46- interviews, etc.) should be small bore pipes were not considered to bethat the walkdown notes do NOT 73A,55-63A, 3FH72-B, employed and documented. signifhcant flood sources.provide a definitive listing of all 3FH80-A, etc.; (2) Appendix A Prepare an integrated list ofequipment and lines or other flood has stressed that the the internal flood sources.sources. Also other fluid sources walkdown notes do NOThave not been considered in the provlde a definitive listing ofanalysis. all equipment and lines orother flood sources. Also otherfluid sources have not beenconsidered in the analysis.6-11 IP-RPT-1O-fiOO23, Rev. 0, Appendix IFSO-B1 There is no list of the internal Prepare an integrated list of Resolved No impactC, Section 4.1.3, which is the flood sources in the analysis the internal flood sources. This is documentation issue. A list of internal Documentation issuesection in the main report for flood that may facilitate PRA flooding sources has been developed and was -incorporated insources, just refers Appendix A, applications, upgrades, and included in a new Table 4.2.1.1 in the final final project file forPlant Walkdown for the information. peer review. update report. This table identifies all the the model used forThere is no list of the internal flood It could facilitate applications, flooding sources in each area and identifies this application.sources in the analysis that may update and review if sources adjacent or lower areas through which floodwaterfacilitate PRA applications, were identified in the main might propagate.upgrades, and peer review. report.50247135000-4722A-24 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix 4 PRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION6-12 tP-RPT-1t-01023, Rev. 0, Appendix IFSO-B2 IP-RPT-10-00023, Rev. 0, Provide adequate Resolved No impactC identifies applicable flood sources Appendix C identifies documentation on the Although Section 3.1.2 previously described the Documentation issuein its Appendix A, Plant Walkdown, applicable flood sources in its process used to identify process for identifying flooding sources, -incorporated inwhich is not adequate for process Appendix A, Plant Walkdown, applicable flood sources additional description has been added to that final project file fordocumentation purpose. For which is not adequate for section and an additional table (Table 4.2.1.1) the model used forexample, the walkdown notes process documentation has been added, which provides additional detail this application.stressed that they do NOT provide a purpose. For example, the describing the sources in each flood zone.definitive listing of all equipment walkdown notes stressed that The statement in the introduction to theand lines or other flood sources; they do NOT provide a walkdown notes was intended only tothere is no list of sources to be definitive listing of all acknowles was tere only toexamined. equipment and lines or other achnowledge that there might be small bore, fieldflood sources; there is no list run piping (less than 1 inch diameter) that wereflood sources; there imno, lnot shown on system drawings and would notof sources to be enamined have been confirmed by the walkdown. Suchsmall bore pipes were not considered to besignificant flood sources.P0247130002-4722u-25 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyA.3 IDENTIFICATION OF KEY ASSUMPTIONSThe methodology employed in this risk assessment followed the NEI guidance. The analysisincluded the incorporation of several sensitivity studies and factored in the potential impactsfrom external events in a bounding fashion. None of the sensitivity studies or boundinganalysis indicated any source of uncertainty or modeling assumption that would have resultedin exceeding the acceptance guidelines. Since the accepted process utilizes a bounding analysisapproach which is mostly driven by that CDF contribution which does not already lead to LERF,there are no identified key assumptions or sources of uncertainty for this application (i.e. thosewhich would change the conclusions from the risk assessment results presented here).A.4 SUMMARYA PRA technical adequacy evaluation was performed consistent with the requirements of RG-1.200, Revision 2. This evaluation combined with the details of the results of this analysisdemonstrates with reasonable assurance that the proposed extension to the ILRT interval forIP2 and IP3 to fifteen years satisfies the risk acceptance guidelines in RG 1.174.A.5 REFERENCES[A.1] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy ofProbabilistic Risk Assessment Results for Risk Informed Activities, Revision 2, March2009.[A.2] Engineering Report, IP2-RPT-09-00026, Rev.0, "Indian Point Unit 2 ProbabilisticSafety Assessment (PSA)", November 2011.[A.13] Engineering Report, IP3-RPT-10-00023, Rev.0, "Indian Point Unit 3 ProbabilisticSafety Assessment (PSA)", November 2012.[A.4] Entergy Fleet Procedure EN-DC-151, Revision 2, "PSA Maintenance and Update",January 2011.[A.5] PWR Owners Group LTR-RAM-II-09-092, "RG 1.200 PRA Peer Review Against theASME PRA Standard Requirements for the Indian Point 2 Nuclear Power PlantProbabilistic Risk Assessment," March 2010.[A.6] PWR Owners Group LTR-RAM-I-11-055, "RG 1.200 PRA Peer Review Against theASME PRA Standard Requirements for the Indian Point 3 Probabilistic RiskAssessment," October 2011.[A.7] "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals:Revision 2-A of 1009325", EPRI, Palo Alto, CA: 2008. 1018243.[A.8] Entergy Engineering Report, IP-RPT-07-00007, "IP2 Cost Benefit Analysis of SevereAccident Mitigation Alternatives", Revision 0.[A.9] Entergy Engineering Report, IP-RPT-07-00008, "IP3 Cost Benefit Analysis of SevereAccident Mitigation Alternatives", Revision 0.P0247130002-4722A-26

-EntergaEnteray Nuclear NortheastIndian Point Energy Center450 Broadway, GSBP.O. Box 249Buchanan, NY 10511-0249Tel 914 254 6700Lawrence CoyleSite Vice PresidentNL-14-128December 9, 2014U.S. Nuclear Regulatory CommissionATTN: Document Control Desk11545 Rockville Pike, TWFN-2 F1Rockville, MD 20852-2738

SUBJECT:

Proposed License Amendment Regarding Extending the Containment Type A LeakRate Testing Frequency to 15 yearsIndian Point Unit Number 2Docket No. 50-247License No. DPR-26

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby requests a LicenseAmendment to Operating License DPR-26, Docket No. 50-247 for Indian Point Nuclear GeneratingUnit No. 2 (IP2). The proposed TS change contained herein would revise Appendix A, TechnicalSpecifications (TS), to allow extension of the ten-year frequency of the Type A or Integrated LeakRate Test (ILRT) that is required by Technical Specification (TS) 5.5.14 to 15 years on apermanent basis.Entergy has evaluated the proposed change in accordance with 10 CFR 50.91 (a)(1) using thecriteria of 10 CFR 50.92(c) and Entergy has determined that this proposed change involves nosignificant hazards, as described in Attachment 1. The marked up page showing the proposedchange is provided in Attachment 2. An assessment of the risk impact of extending the ILRTinterval is provided in Attachment 3. A copy of this application and the associated attachments arebeing submitted to the designated New York State official in accordance with 10 CFR 50.91.Entergy requests approval of the proposed amendment in one calendar year and an allowance of30 days for implementation. There are no new commitments being made in this submittal. If youhave any questions or require additional information, please contact Mr. Robert Walpole, Manager,Regulatory Assurance at (914) 254-6710.AD/7 NL-14-128Docket 50-247Page 2 of 2I declare under penalty of perjury that the foregoing is true and correct. Executed on December,2014.Sincerely,LC/spAttachments: 1. Analysis of Proposed Technical Specification Changes Regarding 15Year Containment ILRT2. Marked Up Technical Specifications Page for Proposed ChangesRegarding 15 Year Containment ILRT3. Risk Impact of Extending the ILRT interval Associated with the ProposedTechnical Specification Changescc: Mr. Douglas Pickett, Senior Project Manager, NRC NRR DORLMr. Daniel H. Dorman, Regional Administrator, NRC Region 1NRC Resident InspectorMr. John B. Rhodes, President and CEO, NYSERDAMs. Bridget Frymire, New York State Dept. of Public Service ATTACHMENT 1 TO NL-14-128ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGESREGARDING 15 YEAR CONTAINMENT ILRTENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NO. 2DOCKET NO. 50-247 NL-14-128Docket No. 50-247Attachment 1Page 1 of 191.0 DESCRIPTIONEntergy Nuclear Operations, Inc. (Entergy) is requesting an amendment to Operating LicenseDPR-26, Docket No. 50-247 for Indian Point Nuclear Generating Unit No. 2 (IP2). The proposedTechnical Specification (TS) change contained herein would revise Appendix A, TS, to allowextension of the ten-year frequency of the Type A or Integrated Leak Rate Test (ILRT) that isrequired by TS 5.5.15 to 15 years on a permanent basis.The specific proposed changes are listed in the following section.2.0 PROPOSED CHANGESThe containment leakage rate testing program in Technical Specification 5.5.15 currently says"A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordance withthe guidelines contained in Regulatory Guide 1.163, "Performance-BasedContainment Leak-Test Program," dated September, 1995."The proposed TS 5.5.15 is as follows:"A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordancewith NEI 94-01, Revision 2A, "Industry Guideline for ImplementingPerformance-Based Option of 10 CFR Part 50, Appendix J," October 2008."3.0 BACKGROUNDThe testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from thecontainment, including systems and components that penetrate the containment, do not exceedthe allowable leakage values specified in the TS. Furthermore, the requirements ensure thatperiodic surveillance of the containment, containment penetrations and isolation valves isperformed so that proper maintenance and repairs are made during the service life of thecontainment, the systems and penetrations. The limitation on containment leakage providesassurance that the containment would perform its design function following an accident up to andincluding the plant design basis accident. Appendix J identifies three types of required tests: (1)Type A tests, intended to measure the containment overall integrated leakage rate; (2) Type Btests, intended to detect local leaks and to measure leakage across pressure-containing orleakage limiting boundaries (other than valves) for containment penetrations; and (3) Type C tests,intended to measure containment isolation valve leakage. Type B and C tests identify the vastmajority of potential containment leakage paths. Type A tests identify the overall integratedcontainment leakage rate and serve to ensure continued leakage integrity of the containmentstructure by evaluating those structural parts of the containment not covered by Type B and Ctesting.

NL-14-128Docket No. 50-247Attachment 1Page 2 of 19In 1995, 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," was amended to provide a performance-based Option B for thecontainment leakage testing requirements. Option B requires that test intervals for Type A, TypeB, and Type C testing be determined by using a performance-based approach. Performance-based test intervals are based on consideration of the operating history of the component andresulting risk from its failure. The use of the term "performance-based' in 10 CFR 50, Appendix Jrefers to both the performance history necessary to extend test intervals as well as to the criterianecessary to meet the requirements of Option B.Regulatory Guide (RG) 1.163 was also issued in 1995. The RG endorsed NEI 94-01, Revision 0,"Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," withcertain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successfulType A tests) to reduce the test frequency from the containment Type A (ILRT) test from threetests in ten years to one test in ten years. This relaxation was based on an NRC risk assessmentcontained in NUREG-1493, "Performance-Based Containment Leak-Test Program," and ElectricPower Research Institute (EPRI) TR-1 04285, "Risk Impact Assessment of Revised ContainmentLeak Rate Testing Intervals." These documents illustrated that the risk increase associated withextending the ILRT surveillance interval was very small.By letter dated August 7, 1996, Indian Point Unit 2 submitted a TS change request, supplementedby letter dated March 12, 1997, to implement 10 CFR 50, Appendix J, Option B. The NRCapproved this request as Amendment 190 issued in NRC letter of April 10, 1997. The NRC notedthe proposed TS changes were in compliance with the requirements of Option B, and areconsistent with the guidance in RG 1.163. With the approval of the amendment, IP2 transitioned toa performance-based ten year frequency for the Type A tests.Entergy submitted an Amendment request to extend the ILRT interval one time from ten years to15 years in a letter dated July 13, 2001 that was supplemented by letters dated November 30,2001 March 13, April 3, May 30, and June 13, 2002. This one-time extension was approved bythe NRC, as license Amendment 232 on August 5, 2002.By letter dated August 31, 2007, NEI submitted NEI 94-01, Revision 2, and EPRI report No.1009325, Revision 2, "Risk Impact Assessment of Extended Integrated Leak Rate TestingIntervals," to the NRC Staff for review. NEI 94-01, Revision 2, describes an approach forimplementing the optional performance-based requirements of Option B, which includes provisionsfor extending Type A intervals to up to 15 years and incorporates the regulatory positions stated inRG 1.163. It delineates a performance-based approach for determining Type A, Type B, and TypeC containment leakage rate surveillance testing frequencies. This method uses industryperformance data, plant-specific performance data, and risk insights in determining the appropriatetesting frequency. NEI 94-01, Revision 2, also discusses the performance factors that licenseesmust consider in determining test intervals.The NEI guideline does not address how to perform the tests because these details are included inreferenced industry documents (e.g., American National Standards institute/American NuclearSociety (ANSI/ANS) 56.8-2002).The NRC final Safety Evaluation (SE) issued by letter dated June 25, 2008, documents theevaluation and acceptance of NEI 94-01, Revision 2, subject to the specific limitations andconditions listed in Section 4.1 of the SE. The accepted version of NEI 94-01 Rev 2A was NL-14-128Docket No. 50-247Attachment 1Page 3 of 19issued as Revision 2A dated October 2008.EPRI Report No. 1009325, Revision 2, provides a risk impact assessment for optimized ILRTintervals of up to 15 years, using current industry performance data and risk-informed guidance,primarily Revision 1 of RG 1.174, "An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." The NRC's final SE issuedby letter dated June 25, 2008, documents the evaluation and acceptance of EPRI Report No.1009325, Revision 2, subject to the specific limitations and conditions listed in Section 4.2 of theSE. An accepted version of EPRI Report No. 1009325 has subsequently been issued asRevision 2A (also identified as Technical Report TR-1 018243) dated October 2008.The proposed amendment would revise TS 5.5.14, "Containment Leakage Rate Testing Program,"by replacing the reference to Regulatory Guide (RG) 1.163, "Performance-Based ContainmentLeak Test Program," with a reference to Nuclear Energy institute (NEI) topical report NEI 94-01,"Industry Guideline for implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"Revision 2A, dated October 2008, as the implementation document used by Entergy to developthe Indian Point 2 performance-based leakage testing program in accordance with Option B of 10CFR 50, Appendix J (Option B).Revision 2A of NEI 94-01 describes an approach for implementing the optional performance-basedrequirements of Option B, including provisions for extending primary containment integrated leakrate test (ILRT) intervals to 15 years, and incorporates the regulatory positions stated in RG 1.163.In the SE issued by NRC letter dated June 25, 2008, the NRC concluded that NEI 94-01, Revision2, describes an acceptable approach for implementing the optional performance-basedrequirements of Option B, and found that NEI 94-01, Revision 2, is acceptable for referencing bylicensees proposing to amend their TS in regard to containment leakage rate testing, subject to thelimitations and conditions noted in Section 4.0 of the SE. IPEC is not applying for the extendedType C performance based testing beyond 60 months but will be adopting the testing criteriaANSI/ANS 56.8 -2002 rather than the criteria of ANSI/ANS 56.8 -1994.The proposed extension of the interval for the primary containment ILRT, which is currentlyrequired to be performed at ten year intervals, to 15 years from the last ILRT would revise the nextscheduled ILRT to March 2021 as opposed to the ILRT currently scheduled for March 2016. Thisis approximately 15 years since the last ILRT which was completed in April 2006.The currently proposed change would allow successive ILRTs to be performed at 15-year intervals(assuming acceptable performance history). The performance of fewer ILRTs would result insignificant savings in radiation exposure to personnel, cost, and critical path time during futurerefueling outages.4.0 Technical EvaluationAs required by 10 CFR 50.54(o), the IP2 containment is subject to the requirements set forth in 10CFR 50, Appendix J. Option B of Appendix J which requires that test intervals for Type A, Type B,and Type C testing be determined by using a performance-based approach. Currently, the 10CFR 50 Appendix J Testing Program Plan is based on RG 1.163, which endorses NEI 94-01,Revision 0. This LAR proposes to revise the 10 CFR 50, Appendix J Testing Program Plan byimplementing the guidance in NEI 94-01, Revision 2A but will not extend the Type B and Cleakage beyond 60 months. Testing will be performed in accordance with ANSI/ANS 56.8 -2002.

NL-14-128Docket No. 50-247Attachment 1Page 4 of 194.1 Limitations and ConditionsIn the June 25, 2008 NRC SE, the NRC concluded that NEI 94-01, Revision 2, describes anacceptable approach for implementing the optional performance-based requirements of Option B,and found that NEI 94-01, Revision 2, is acceptable for referencing by licensees proposing toamend their TS in regard to containment leakage rate testing, subject to the limitations andconditions noted in Section 4.0 of the SE.The following Table 4.1 -1 lists the SE Section 4.1 Limitations and Conditions as well ascompliance with each of the six limitations and conditions.Table 4.1-1Limitations and Conditions (Section IP2 Compliance4.1 of Safety Evaluation Dated June,25,2008)For calculating the Type A leakage rate, Implementation of NEI 94-01 Rev 2A willthe licensee should use the definition in require use of the definition of "performancethe NEI TR 94-01, Revision 2, in lieu of leakage rate" defined in Section 5.0 forthat in ANSI/ANS-56.8-2002. (Refer to SE calculating the Type A leakage rate whenSection 3.1.1.1). performing Type A tests.The licensee submits a schedule of NEI-94-01 Rev 2A, Section 9.2.3.2 requires acontainment inspections to be performed general visual examination prior to each Typeprior to and between Type A tests. (Refer A test and at least 3 other outages before theto SE Section 3.1.1.3). ILRT. This should be scheduled inconjunction with or coordinated withexaminations required by ASME Code,Section Xl, Subsections IWE and IWL. Aschedule of containment inspections isprovided in Section 4.4The licensee addresses the areas of the A general visual examination of accessiblecontainment structure potentially interior and exterior surfaces is conducted persubjected to degradation. (Refer to SE the Containment Inservice Inspection PlanSection 3.1.3). which implements the requirements of ASME,Section Xl, Subsections IWE and IWL. IP2will explore / consider inaccessibledegradation-susceptible areas that can beinspected using viable, commercially availableNDE methods.The licensee addresses any tests and The design change process will address anyinspections performed following major testing and inspection requirements followingmodifications to the containment future major modifications to the containmentstructure, as applicable. (Refer to SE structure. This process provides a disciplinedSection 3.1.4). approach for determining the program andsystem interfaces associated with designchange. This process evaluates requirementspertaining to the ASME Containment In-Service Inspection Program, ASME AppendixJ (Primary Containment Leak Rate Testing)

NL-14-128Docket No. 50-247Attachment 1Page 5 of 19Table 4.1-1Limitations and Conditions (Section IP2 Compliance4.1 of Safety Evaluation Dated June,25,2008)Program, and ASME Section Xl.The normal Type A test interval should be IP2 is adopting, consistent with Section 9.2.2less than 15 years. If a licensee has to of NEI 94-01 Rev 2A, a Type A test intervalutilize the provision of Section 9.1 of NEI defined as the time period from the completionTR 94-01, Revision 2, related to extending of a Type A test to the start of the next test.the ILRT interval beyond 15 years, the This definition will be used for scheduling andlicensee must demonstrate to the NRC planning of the next Type A test to the monthstaff that it is an unforeseen emergent and year (see RIS 2008-27).condition. (Refer to SE Section 3.1.1.2).For plants licensed under 10 CFR Part 52, Not applicable to IP2.applications requesting a permanentextension of the ILRT surveillance intervalto 15 years should be deferred until afterthe construction and testing ofcontainments for that design have beencompleted and applicants have confirmedthe applicability of NEI TR 94-01, Revision2, and EPRI Report No. 1009325,Revision 2, including the use of pastcontainment ILRT data.4.2 Existing ExceptionsThe provisions of RG 1.163 have been incorporated into NEI 94-01 Revision 2A so if there hadbeen an exception to RG 1.63 it would remain unchanged.4.3 Previous Test results4.3.1 ILRT Test ResultsPast IP2 ILRT results have confirmed that the containment is acceptable with respect to the designcriterion of 0.1% leakage of containment air weight at the design basis loss of coolant accidentpressure (La). Since the last two Type A "as found" tests for IP2 had "as found" test results of lessthan 1.01La, a test frequency of 15 years in accordance with NEI 94-01 Revision 2A would beacceptable. The last two tests were:1. The last ILRT in April 2006 had a measured containment leak rate (Ltm) at the testpressure of 60.5 psia was 0.0636 % containment air weight / day with a 95% confidencelevel.2. The prior ILRT in June 1991 had a measured containment leak rate (Ltm) at the testpressure of 61.7 psia was 0.0478 % containment air weight / day with a 95% confidencelevel.For background, the prior three Type A tests had the following results:

NL-14-128Docket No. 50-247Attachment 1Page 6 of 19Date As found Leakage (% Test Pressure (psia)Containment weight perday)December, 1987 0.0342 62.9September, 1984 0.0320 65.6August, 1979 0.0260 62.74.3.2 Type B and C testingThe IP2 Appendix J, Type B and Type C testing program requires testing of the componentsrequired by 10 CFR 50, Appendix J, Option B. Technical Specification Amendment 174, datedJune 17, 1997, approved the adoption of 10 CFR 50, Appendix J, Option B performance basedtesting requirements for containment leakage testing. The minimum pathway combined Type Band Type C leakage from the March 2006 outage, when the last Type A test was performed, isprovided below. The subsequent combined as found Type B and Type C test values during eachsuccessive outage since the last Type A test are also provided below. The data is provided inpercentage of leakage allowed (0.6La).Table 4.3-1Date As-Found La (ccm) Percent ((As- Percent ((As-Leakage Found/La) xl00) Found/.6La))xl 00)(sccm)April 46,105.04 215490 0.214 0.3572006April 54,659.95 215490 0.254 0.4232008April 28,880.44 215490 0.134 0.2232010April 47,304.18 215490 0.220 0.3662012March 79,176.85 215490 0.367 0.6122014 _I IBased on the results the largest as found leakage and the as left conditions are within theacceptance criterion associated with the 15 year ILRT.Table 4.3-2 provides a listing of the containment penetrations subject to Type B and C testing, thetest frequency, the last test date and the next test date, and the as left leakage. Notes are providedfor test failures.

NL-14-128Docket No. 50-247Attachment 1Page 7 of 19Table 4.3-2Penetration Description Type Test Frequency Last Test date Next test date "as -Left"(months) Leakage(cc/min)Penetration UU B 30 3/2/14 3/16 0.00Penetration W B 30 3/2/14 3/16 0.00Fuel Transfer Tube B 30 3/16/14 3/16 38.25Equipment Hatch Seal B 30 3/15/14 3/16 93.0080ALOK Personnel Airlock -80 foot B 30 6/12/14 12/16 6818.6095ALOK Personnel Airlock -95 foot B 30 6/6/14 12/16 14481.00WCCPP Zone 2 -Racks 10, 11 B 36 4/12/13 4/12/15 12744.00WCCPP Zone 2 -Racks 12,13 B 36 4/12/13 4/12/15 2265.60Y Pressurizer relief tank N2 supply tank C 30 2/28/14 3/16 1387.50RCS -Valve RC-518Y Pressurizer relief tank N2 supply tank C 60 3/10/14 3/18 3.50RCS -Valve RC-3418, 3419 and 4136GG Containment spray headers -Valve C 60 3/12/14 3/18 1570.25867A,878AP Containment spray headers -Valve SI- C 60 3/6/14 3/18 0867BRR Accumulator N2 supply -Valve 863- C 60 3/16/12 3/16 199.00RR Accumulator N2 supply -Valve 4312 C 60 3/16/12 3/16 6.00V Primary system vent and N2 supply -C 60 3/14/14, 3/18 31.00Valve WD-3416, 3417, 5459V Primary system vent and N2 supply- C 30 3/14/14 3/16 21000Valve WD-1616RR Containment Air Sample In (Rad) -C 60 3/5/13 3/18 32.50Valves PCV-1234, PCV-1235RR Containment Air Sample Pot (Rad) -C 60 3/5/13 3/18 2.80Valves PCV-1236, PCV-1237R Air Ejector Discharge to Containment -C 30 3/3/14 3/16 271.50Valve CA-1229R Air Ejector Discharge to Containment -C 30 3/3/14 3/16 135.75 NL-14-128Docket No. 50-247Attachment 1Page 8 of 19Table 4.3-2Penetration Description Type Test Frequency Last Test date Next test date "as -Left"(months) Leakage(cc/min)Valve CA-1230EE Vent Purge Supply Duct -Valve VS- C 30 3/12/14 3/16 6380.001170 and VS-1171FF Vent Purge Exhaust Duct -Valve VS- C 30 3/12/14 3/16 9482.501172 and VS-1173PP Cont Pressure Relief Vent -Valves VS- C 30 3/12/14 3/16 300.001190, VS-1191PP Cont Pressure Relief Vent -Valve VS- C 30 3/12/14 3/16 294.001192TT Post Accident Sample system supply C 60 3/12/14 3/18 0.00lines -Valve SP-5018 and SP-5019LL Post Accident Sample system supply C 60 3/12/14 3/18 3.00lines -Valve SP-5020 and SP-5021R Post Accident Sample system return C 60 2/28/14 3/18 0.00lines -Valve SP-5022 and SP-50230 Post Accident Sample system return C 60 2/28/14 3/18 0.00lines -Valve SP-5024 and SP-5025Y Instrument air (post accident vent C 60 3/3/14 3/18 8.50supply) -Valve IA-39Y Instrument air (post accident vent C 30 3/29/11 3/16 24.25supply) -Valve IA-1228LL Post Accident Vent Exhaust Valves E-2 C 60 2/26/14 3/18 0.00and E-1, E-3, E-5Personnel air lock -Outer Door Valve C 60 2/28/13 3/18 57.0085APersonnel air lock -Outer Door Valve C 60 2/28/13 3/18 250.1095APersonnel air lock -Inner Door Valve C 60 2/28/13 3/18 59.5085BPersonnel air lock -Inner Door Valve C 60 2/28/13 3/18 0.3595B I IIII_ I NL-14-128Docket No. 50-247Attachment 1Page 9 of 19Table 4.3-2Penetration Description Type Test Frequency Last Test date Next test date "as -Left"(months) Leakage(cc/min)Personnel air lock- Inner Door Valve C 60 2/28/13 3/18 37.5085CPersonnel air lock -Inner Door Valve C 60 2/28/13 3/18 0.0095CPersonnel air lock- Inner Door Valve C 60 2/28/13 3/18 47.2585DPersonnel air lock -Inner Door Valve C 60 2/28/13 3/18 1.8095DPneumatic Indicator Lines (SG level-2, C 60 3/14/14 3/18 5.00pressurizer level-i, pressurizerpressure-I) -Valve IIP-500 and IIP-501Pneumatic Indicator Lines (SG level-2, C 30 3/14/14 3/18 590.00pressurizer level-i, pressurizerpressure-I) -Valve IIP-502 and IIP-503Pneumatic Indicator Lines (SG level-2, C 60 3/14/14 3/18 7.00pressurizer level-i, pressurizerpressure-I) -Valve IIP-504 and IIP-505Pneumatic Indicator Lines (SG level-2, C 60 3/14/14 3/18 16.00pressurizer level-i, pressurizerpressure-1) -Valve IIP-506 and IIP-507 NL-14-128Docket No. 50-247Attachment 1Page 10 of 194.4 Code InspectionsPrior to each Type A test a general visual examination is required of accessible interior andexterior surfaces of the containment for structural issues that may affect the performance of theType A test. This inspection will be performed as part of the Containment Inservice Inspection (ISI)Plan to implement the requirements of ASME, Section Xl, Subsection IWE and IWL (the applicablecode edition and addenda for the fourth 10 year interval is ASME Section Xl, 2001 Editionincluding the 2002 and 2003 Addenda in paragraph (b)(2)).The examination performed in accordance with the ISI program to meet Subsections IWE and IWLsatisfies the general visual examination requirements specified in Option B. The identification andevaluation of inaccessible areas are addressed in accordance with the requirements of 10 CFR50.55a(b)(2)(ix). Each ten year ISI interval is divided into three approximately equal inspectionperiods. A minimum of one inspection required by the IWE inspection program is performed duringeach inspection period of the ISI period to meet the program requirements. IWL visualexaminations of accessible concrete containment surfaces are to be completed once every 5 yearswithin the limitations specified in IWL-2410(b), (c), and (d) resulting in at least two IWLexaminations being performed during a 15 year type A and typically scheduled in two of the threeinspection periods of a 10 year ISI interval. Therefore, the frequency of the examinationsperformed in accordance with the IWE / IWL program will satisfy the requirements of NEI 94-01Revision 2A, Section 9.2.3.2, to perform a general visual examination before the Type A test duringat least three other outages before the next Type A test if the interval is extended to 15 years. Thelast ILRT was performed April 2006 and the next 15 year interval will end 12 months after 2R24scheduled for the spring of 2020. The following Tables illustrates the current and plannedinspection intervals for the IP2 first and second IWE inspection intervals:Table 4.4-1IWE InspectionsInspection Inspection Period Start Period End Refuel RefuelInterval Period Date Date Outage Month/YearSeptember September 2R13 Spring 19971 1 9,1996 9, 2001 2R14 Spring 2001September Jan 9, 2005 Spring 20021 2 9, 2001 2R151 3 Jan 10, 2005 Feb 28,2007* 2R16 Spring 20042R17 Spring 20062R18 Spring 20082 1 March 1, 2007 May 31, 2010 2R19 Spring 20102 2 June 1, 2010 May 31,2013 2R20 Spring 20122 3 June 1,2013 May 31, 2016 2R21 Spring 20142R22 Spring 2016* Based upon this extended First Period that ended on September 9, 2001, the First 10-YrInterval for IP2 Containment ISI was originally scheduled to end on May 9, 2010, but wasshortened to align with the Third ISI Interval.

NL-14-128Docket No. 50-247Attachment 1Page 11 of 19The IWL inspections are performed per the following schedule:Table 4.4.2IWL InspectionsInspection Interval Inspection Period IWL Inspection Dates1 1 June 20001 2 June 20051 3 June 20102 1 June 20152 2 June 2020For IP2 the First Interval CII Program Plan was originally effective from September 9, 1996,through and including May 9, 2010. This time period has been shortened to end on February 28,2007. IWE Containment inservice examinations scheduled for the first 40-month period werecompleted during the Third Period of the Third ISI Inspection Interval. These examinations nowserve the same purpose as pre-service baseline examinations. The required IWL inserviceexaminations were also completed and re-inspections are scheduled at 5 year frequency.The Second Ten-Year Interval for IWE Containment ISI inspections at IP2 will commence onMarch 1, 2007 coincident with the start of the Fourth 10-Year ISI Program Interval. Therefore, boththe ISI and the CII IWE & IWL Program Plans will be aligned with the Fourth Interval ISI Programschedule and ASME Code requirements.The following information provides the IP2 IWE examination results of the containment metal linercompleted during refuel outages 2R18 (2008), 2R20 (2012) and 2R21 (2014) and the IWLexamination results for the containment concrete visual inspections completed in 2005 and 2010(these are not always completed in an outage). The next IWE examination is scheduled for 2R23(2018) prior to the proposed date for the next ILRT. The next IWL examination is scheduled for2016 and the inspection will also be scheduled prior to the proposed date for the next ILRT 2R24(2020). Corrective Actions identified by these inspections are provided with the discussions. Thereare no primary containment surface areas that require augmented examination in accordance withASME Section XI, IWE-1240.4.4.1 IWE ExaminationsIP2 IWE containment inspection for the current fourth ISI interval was performed on 2008 -2R18outage, 2012 -2R20 outage and 2014- 2R21 outage.Refueling Outage 2R18 (2008) Containment Inservice InspectionExaminations were performed for the Containment Surfaces (Containment Vessel AccessibleSurface Areas, Bolted connections, Wall and Dome Liner and Moisture Barriers) during 2R18 in2008. There were some deficiencies noted such as general surface corrosion, minor coatingpeeling/flaking, blistered paint, loose stainless steel insulation panels and buckling stainless steelinsulation panels (VC liner inaccessible) at columns 10 and 11 elevation 68'.The general surface corrosion, minor coating peeling/flaking and blistered paint were previouslyidentified and evaluated. These conditions were a repeat of previous inspections and were minorwith no change and therefore acceptable.

NL-14-128Docket No. 50-247Attachment 1Page 12 of 19The condition of the buckling locations and looseness on the VC liner plate insulation wasdocumented in the Corrective Action Program as Condition Report CR-IP2-2008-01892. CivilEngineering performed an inspection of the stainless steel insulation jacket and has determinedthat all but 2 of the insulation jacket issues are acceptable. The two areas not acceptable wererepaired during 2R1 8 outage.Refueling Outage 2R20 (2012) Containment Inservice InspectionExaminations were performed for the Containment Surfaces (Containment Vessel AccessibleSurface Areas, Bolted connections, Wall and Dome Liner and Moisture Barriers) during 2R20 in2012. Most of the findings such as surface corrosion and minor coating flaking and peeling were arepeat of previous inspections and were minor with no change and therefore acceptable. Therewere also some deficiencies noted on the Electrical penetration #69 of the Containment Buildingpenetrations; there was observed water seeping adjacent to penetration #69. This condition wasdocumented in IP2 Corrective Action Program under Condition Report CR-1P2-2012-01760. CivilDesign engineering walked down the penetration and the water seepage is from areas wherecrack/delimitation repairs where performed back in 2000. The water seepage observed has noadverse effect on the penetration as it is not emanating from the penetration sleeve. The sealaround the penetration is intact and the inside of the penetration itself is dry. This penetration wasalso looked at from the inside of the VC during the Maintenance Rule Inspection and no anomalieswere observed.All of the conditions noted during this inspection did not result in any structural degradation thatadversely affects the ability of the containment to perform its design function of maintainingintegrity during accident conditions.Refueling Outage 2R21 (2014) Containment Inservice InspectionExaminations were performed for the Containment Surfaces (Containment Vessel AccessibleSurface Areas, Bolted connections, Wall and Dome Liner and Moisture Barriers) during 2R21 in2014. Most of the findings were a repeat of previous inspections and were minor with no changeand therefore acceptable. All NDE examination reports were accepted during the 2014containment inspection therefore no condition reports were generated.4.4.2 IWL ExaminationsThe inspections are general visual inspections performed in accordance with the requirements ofthe ASME Boiler and Pressure Vessel Code, 1998 Edition,Section XI. Division 1, SubsectionIWL as required and modified by NRC, Code of Federal Regulation, Title 10, Part 50,Section 55a, "Codes and Standards,"(10 CFR 50.55a -1999). When needed, opticalenhancement equipment with zoom capabilities are used as visual aids during the inspections.All of the inspections are performed under the direction of the IWL Responsible Engineer(RE). The RE is the Civil/Structural Design Engineering Supervisor at IPEC and a NewYork State Registered Professional Engineer in accordance with the IWL Procedure. TheResponsible Engineer has knowledge of the Design and Construction Codes as well as othercriterion used in IP2's Containment. Degreed engineers perform the inspections under thedirection of the RE and are knowledgeable and trained in the design, evaluation andperformance requirements of structures and qualified to perform visual examination eitherdirectly or remotely, with adequate illumination, to detect evidence of degradation.

NL-14-128Docket No. 50-247Attachment 1Page 13 of 19The second period of the first interval of the IP2 IWL containment inspection was performed in thespring of 2005 and documented in IP-RPT-06-00019. Visual examinations were performed of allaccessible areas of the containment building exterior concrete including areas visible from insideother surrounding buildings. The concrete exhibited signs of normal weathering that are to beexpected for the time period that it has been in service. These indications include minor crackingto due pressurization, and minor areas of spalling with exposed rebar and cadwelds. The spallingat the cadwelds appears to be due to lack of concrete cover as a result of the cadwelds havingtwice the diameter as the rebar. There were also some locations of efflorescence which weredetermined to be unchanged since the previous inspection and thus deemed inactive. Severalareas of rust bleeding were identified but easily attributed to the lightning arrestors and the ductwork and have no impact on the structural capacity of the containment building. All together therewere 91 recordable indications identified during the inspection however all of them have beenevaluated and are not structural concerns. None of the indications reduce the structural capacityor ability of the containment structure to perform its safety function. Based on condition ofinspected areas it was not deemed necessary to inspect non-accessible areas. No conditionreports or work orders were required as a result of the inspection.The third period of the first interval of the IP2 IWL containment inspection was performed in thespring of 2010 and documented in IP-RPT-10-00027. Visual examinations were performed of allaccessible areas of the containment building exterior concrete including areas visible from insideother surrounding buildings. The concrete exhibited signs of normal weathering that are to beexpected for the time period that it has been in service. These indications include minor crackingto due pressurization, and minor areas of spalling with exposed rebar and cadwelds. The spallingat the cadwelds appears to be due to lack of concrete cover as a result of the cadwelds havingtwice the diameter as the rebar. There were also some locations of efflorescence which weredetermined to be unchanged since the previous inspection and thus deemed inactive. Severalareas of rust bleeding were identified but easily attributed to the lightning arrestors and the ductwork and have no impact on the structural capacity of the containment building. All together therewere 125 recordable indications identified during the inspection which increased from the 91identified in the previous inspection. This is partially attributed to the ILRT performed in 2006which caused several of the previous identified areas of potential future spalling to indeed spall. Inthe fall of 2009 several of the previously identified areas were cleaned and a coating was appliedto protect the exposed steel from future corrosion. All of the recordable indications identifiedduring the inspection have been evaluated and are not structural concerns. None of theindications reduce the structural capacity or ability of the containment structure to perform its safetyfunction. Based on condition of inspected areas it was not deemed necessary to inspect non-accessible areas. No condition reports or work orders were required as a result of the inspection.4.5 Confirmatory Analysis4.5.1 MethodologyAn evaluation has been performed to assess the risk impact of extending the IP2 ILRT interval fromthe current ten years to 15 years. This plant-specific risk assessment followed the guidance in NEI94-01, Revision 2A, the methodology outlined in EPRI TR-1 04285, August 1994 and TR-1 009325,Revision 2A, and the NRC regulatory guidance outlined in RG 1.174 on the use of Probabilistic RiskAssessment (PRA) findings and risk insights in support of a request to change the licensing basis ofthe plant. In addition, the methodology used for Calvert Cliffs Nuclear Power Plant to estimate thelikelihood and risk implication of corrosion-induced leakage of steel containment liners goingundetected during the extended ILRT interval was also used for sensitivity analysis.

NL-14-128Docket No. 50-247Attachment 1Page 14 of 19In their June 25, 2008, SE, the NRC concluded that a 15 year extension to the Type A ILRT intervalwas acceptable and that the methodology in EPRI TR-1009325, Revision 2, is acceptable forreferencing in a proposal to amend TS to extend the ILRT surveillance interval to 15 years. Thisapproval was subject to the limitations and conditions noted in Section 4.0 of the SE. The followingTable 4.5-1 lists the SE Section 4.2 Limitations and Conditions and a description of how the IP2analysis complies with those four limitations and conditionsTable 4.5 -1Limitations and Conditions of Risk IP2 ComplianceAssessmentThe licensee submits documentation The technical adequacy of the IP2 PRA andindicating that the technical adequacy of their consistency with the RG 1.200 requirementsPRA is consistent with the requirements of relevant to the ILRT extension are discussed inRG 1.200 relevant to the ILRT extension Section 4.5.2 and detailed in Appendix A ofapplication. Attachment 3.The licensee submits documentation The IP2 risk evaluation is summarized inindicating that the estimated risk increase Section 4.5.3 and described in detail inassociated with permanently extending the Attachment 3. The results of thatILRT surveillance interval to 15 years is small, evaluation demonstrate that the estimatedand consistent with the clarification provided risk increase is small and consistent within Section 3.2.4.5 of this SE. Specifically, a the criteria discussed in the SE.small increase in population dose should bedefined as an increase in population dose ofless than or equal to either 1.0 person-remper year or 1 percent of the total populationdose, whichever is less restrictive. In addition,a small increase in CCFP should be definedas a value marginally greater than thataccepted in previous one-time 15-year ILRTextension requests. This would require thatthe increase in CCFP be less than or equal to1.5 percentage point. While acceptable forthis application, the NRC staff is notendorsing these threshold values for otherapplications. Consistent with this limitationand condition, EPRI Report No. 1009325 willbe revised in the "-A" version of the report, tochange the population dose acceptanceguidelines and the CCFP guidelines.The methodology in EPRI Report No. The IP2 analysis used a pre-existing containment1009325, Revision 2, is acceptable except for leak rate of 1 0OLa to calculate the increase inthe calculation of the increase in expected population dose for the large leak rate accidentpopulation dose (per year of reactor case (EPRI Class 3b) .(Attachment 3, Sectionoperation). In order to make the methodology 1.3).acceptable, the average leak rate for the pre-existing containment large leak rate accidentcase (accident case 3b) used by thelicensees shall be 100 La instead of 35 La.

NL-14-128Docket No. 50-247Attachment 1Page 15 of 19Table 4.5 -1Limitations and Conditions of Risk IP2 ComplianceAssessmentA LAR is required in instances where Containment overpressure is not relied upon forcontainment over-pressure is relied upon for ECCS performance (Attachment 3, Section 5.8).ECCS performance.4.5.2 PRA QualityThe risk assessment performed for the IP2 ILRT extension request is based on the current Level 1and Level 2 PRA model of record, which was released in November 2011. Information developedfor the license renewal effort to support the Level 2 release categories is also used in this analysissupplemented by additional calculations to more appropriately represent the intact containmentcase in the ILRT extension risk assessment. A discussion of the Entergy model update process,the peer review performed on the IP2 model, the results of that peer review and the potentialimpact of peer review findings on the ILRT extension risk assessment are provided in Attachment3, Section A.2.It should be noted that, while the analysis presented in Attachment 3 was performed for both IP2and IP3, this submittal only addresses a LAR for IP2. The IP2 information presented in Attachment3 is therefore informational only and not part of the basis for the current LAR.4.5.3 Summary of Plant-Specific Risk Assessment ResultsThe findings of the IP2 risk assessment confirm the general findings of previous studies that therisk impact associated with extending the ILRT interval to one in 15 years is small. The IP2 plant-specific results for extending the ILRT interval to 15 years, taken from Attachment 3, Section 7.0,Conclusions, are summarized below.1. Reg. Guide 1.174 provides guidance for determining the risk impact of plant-specific changesto the licensing basis. Reg. Guide 1.174 defines "very small" changes in risk as resulting inincreases of CDF below 1.OE-06/yr and increases in LERF below 1.OE-07/yr. "Small" changesin risk are defined as increases in CDF below 1.0E-05/yr and increases in LERF below 1.OE-06/yr. Since the ILRT extension was demonstrated to have no impact on CDF for IP2, therelevant criterion is LERF. The increase in internal events LERF resulting from a change in theType A ILRT test interval for the base case with corrosion included for IP2 is estimated at9.84E-08 /yr (see Attachment 3, Table 5.6-1A), which is within the small change region of theacceptance guidelines in Reg. Guide 1.174. In using the EPRI Expert Elicitation methodology,the change is estimated as 1.05E-08 /yr (see Attachment 3, Table 6.2-2A), which is within thevery small change region of the acceptance guidelines in Reg. Guide 1.174.2. The change in dose risk for changing the Type A test frequency from three-per-ten years toonce-per-fifteen-years, measured as an increase to the total integrated dose risk for all internalevents accident sequences is 0.584 person-rem/yr (0.62%) using the EPRI guidance with thebase case corrosion case (Attachment 3, Table 5.6-1A). The change in dose risk drops to0.111 person-rem/yr when using the EPRI Expert Elicitation methodology (Attachment 3, Table6.2-2A).

NL-14-128Docket No. 50-247Attachment 1Page 16 of 193. The increase in the conditional containment failure frequency from the three in ten year intervalto one in fifteen years including corrosion effects using the EPRI guidance (see Section 5.5) is0.84% for IP2. This value drops to less that 0.10% for IP2 using the EPRI Expert Elicitationmethodology (see Attachment 3 Table 6.2-2A). This is below the acceptance criteria of lessthan 1.5% defined Attachment 3 in Section 1.3.4. To determine the potential impact from external events, a bounding assessment from the riskassociated with external events utilizing information from the IP2 IPEEEs similar to theapproach used in the License Renewal SAMA analysis. As shown in Attachment 3 Table 5.7-2A the total increase in LERF for IP2 due to internal events and the bounding external eventsassessment is 5.20E-07/yr. This value is in Region II of the Reg. Guide 1.174 acceptanceguidelines.5. As shown in Attachment 3, Table 5.7-4, the same bounding analysis indicates that the totalLERF from both internal and external risks is 6.78E-06/yr for IP2, which is less than the Reg.Guide 1.174 limit of 1.OE-05/yr given that the ALERF is in Region II (small change in risk).6. Finally, since the external events assessment led to exceeding one of the two alternativeacceptance criteria (i.e. greater than 1.0 person-rem/yr, an alternative detailed boundingexternal events assessment was also performed to demonstrate that the alternate 1.0%person-rem/yr criterion and the other acceptance criteria could still be met. In this case, asshown in Attachment 3, Table 5.7-7 for IP2, the total change in LERF from both internal andexternal events was 5.52E-7/yr, the change in person-rem/yr was 3.28/yr representing 0.59%of the total, and the change in the CCFP was 0.89%. All of these calculated changes meet theacceptance criteria. As shown in Attachment 3, Table 5.7-8, this assessment indicates that thetotal LERF from both internal and external risks is 2.65E-06/yr for IP2, which is less than theReg. Guide 1.174 limit of 1.OE-05/yr given that the ALERF is in Region II (small change in risk).7. Including age-adjusted steel liner corrosion effects in the ILRT assessment was demonstratedto be a small contributor to the impact of extending the ILRT interval for IP2.Therefore, increasing the ILRT interval on a permanent basis to a one-in-fifteen yearfrequency is not considered to be risk significant. Details of the IP2 risk assessment are containedin Attachment 3.4.6 ConclusionNEI 94-01, Revision 2A, describes an NRC-accepted approach for implementing theperformance-based requirements of 10 CFR 50, Appendix J, Option B. It incorporates theregulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals to15 years. NEI 94-01, Revision 2A delineates a performance-based approach for determiningType A, Type B, and Type C containment leakage rate surveillance test frequencies. IP2 isproposing to adopt the guidance of NEI 94-01, Revision 2A for the 10 CFR 50, Appendix J, testingprogram plan and the ANSI/ANS 56.8 -2002 standard for Type A, B and C tests..Based on the previous ILRT tests conducted at IP2, supplemented by risk analysis studies,including the IP2 risk analysis provided in Attachment 3, it may be concluded thatextension of the containment ILRT interval from ten to 15 years represents minimal riskperformed in accordance with Option B and inspected per the guidance NEI-94-01 Revision 2A.

NL-14-128Docket No. 50-247Attachment 1Page 17 of 195.0 REGULATORY ANALYSIS5.1 No Significant Hazards ConsiderationEntergy has evaluated the safety significance of the proposed change to the IP2 TS which reviseIP2 TS 3.5.15, "Containment Leakage Rate Testing Program," to allow a permanent extension tothe frequency of Type A testing based upon performance criteria. The proposed changes havebeen evaluated according to the criteria of 10 CFR 50.92, "Issuance of Amendment". Entergy hasdetermined that the subject changes do not involve a Significant Hazards Consideration, asdiscussed below1. Does the proposed amendment involve a significant increase in the probabilityor consequences of an accident previously evaluated?Response: No.The proposed amendment involves changes to the IP2 containment leakage rate testingprogram. The proposed amendment does not involve a physical change to the plant or achange in the manner in which the plant is operated or controlled. The primarycontainment function is to provide an essentially leak tight barrier against the uncontrolledrelease of radioactivity to the environment for postulated accidents. As such, thecontainment itself and the testing requirements to periodically demonstrate the integrity ofthe containment exist to ensure the plant's ability to mitigate the consequences of anaccident do not involve any accident precursors or initiators. Therefore, the probability ofoccurrence of an accident previously evaluated is not significantly increased bythe proposed amendment.The proposed amendment adopts the NRC accepted guidelines of NEI 94-01, Revision2A, for development of the IP2 performance-based testing program for the Type A testing.Implementation of these guidelines continues to provide adequate assurance that duringdesign basis accidents, the primary containment and its components would limit leakagerates to less than the values assumed in the plant safety analyses. The potentialconsequences of extending the ILRT interval to 15 years have been evaluated byanalyzing the resulting changes in risk. The increase in risk in terms of person-rem peryear within 50 miles resulting from design basis accidents was estimated to be acceptablysmall and determined to be within the guidelines published in RG 1.174. Additionally, theproposed change maintains defense-in-depth by preserving a reasonable balance amongprevention of core damage, prevention of containment failure, and consequencemitigation. Entergy has determined that the increase in conditional containment failureprobability due to the proposed change would be very small. Therefore, it is concludedthat the proposed amendment does not significantly increase the consequences of anaccident previously evaluated.Therefore, the proposed change does not involve a significant increase in theprobability or consequences of an accident previously evaluated.

NL-14-128Docket No. 50-247Attachment 1Page 18 of 192. Does the proposed amendment create the possibility of a new or differentkind of accident from any accident previously evaluated?Response: No.The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 2A,for the development of the IP2 performance-based leakage testing program, andestablishes a 15-year interval for the performance of the containment ILRT. Thecontainment and the testing requirements to periodically demonstrate the integrity of thecontainment exist to ensure the plant's ability to mitigate the consequences of an accidentdo not involve any accident precursors or initiators. The proposed change does not involvea physical change to the plant (i.e., no new or different type of equipment will be installed)or a change to the manner in which the plant is operated or controlled.Therefore, the proposed change does not create the possibility of a new or different kindof accident from any previously evaluated.3. Does the proposed amendment involve a significant reduction in a margin ofsafety?Response: No.The proposed amendment adopts the NRC-accepted guidelines of NEI 94-01, Revision 2A,for the development of the IP2 performance-based leakage testing program, and establishesa 15-year interval for the performance of the containment ILRT. This amendment does notalter the manner in which safety limits, limiting safety system setpoints, or limiting conditionsfor operation are determined. The specific requirements and conditions of the containmentleakage rate testing program, as defined in the TS, ensure that the degree of primarycontainment structural integrity and leak-tightness that is considered in the plant's safetyanalysis is maintained. The overall containment leakage rate limit specified by the TS ismaintained, and the Type A containment leakage tests would be performed at the frequenciesestablished in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 2A withno change to the 60 month frequencies of Type B, and Type C tests.Containment inspections performed in accordance with other plant programs serve to providea high degree of assurance that the containment would not degrade in a manner that is notdetectable by an ILRT. A risk assessment using the current IP2 PSA model concluded thatextending the ILRT test interval from ten years to 15 years results in a very small change to therisk profile.Therefore, the proposed change does not involve a significant reduction in a margin ofsafety.Based on the above, Entergy concludes that the proposed amendment to the Indian Point 2Technical Specifications presents no significant hazards consideration under the standards setforth in 10 CFR 50.92(c), and accordingly, a finding of 'no significant hazards consideration' isjustified.

NL-14-128Docket No. 50-247Attachment 1Page 19 of 195.2 Applicable Regulatory Requirements / CriteriaThe NRC Order of February 11, 1980 required an evaluation of the degree of compliance with theGDC at the time. This section discusses continued compliance with certain of those criteria.The plant will continue to meet Criterion 1 of 10 CFR 50.36 which says "Structures, systems andcomponents important to safety shall be designed, fabricated, erected, and tested to qualitystandards commensurate with the importance of the safety functions to be performed. Wheregenerally recognized codes and standards are used, they shall be identified and evaluated todetermine their applicability, adequacy, and sufficiency and shall be supplemented or modified asnecessary to assure a quality product in keeping with the required safety function. A qualityassurance program shall be established and implemented in order to provide adequate assurancethat these structures, systems and components will satisfactorily perform their safety functions.Appropriate records of the design, fabrication, erection, and testing of structures, systems andcomponents important to safety shall be maintained by or under the control of the nuclear powerplant licensee throughout the life of the unit' and Criterion 3 which says "Structures, systems, andcomponents important to safety shall be designed to withstand the effects of natural phenomenasuch as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capabilityto perform their safety functions. The design bases for these structures, systems and componentsshall reflect: (1) appropriate consideration of the most severe of the natural phenomena that havebeen historically reported for the site and surrounding area, with sufficient margin for the limitedaccuracy, quantity, and period of time in which the historical data have been accumulated, (2)appropriate combinations of the effects of normal and accident conditions with the effects of thenatural phenomena and (3) the importance of the safety functions to be performed."The extension of the duration of the ILRT for the containment will not affect the design, fabrication,or construction of the containment structure and the design will continue to account for the effectsof natural phenomena. The ILRT of the containment will continue to be done in accordance with10 CFR 50 Appendix J using 10 CFR 50 Appendix B quality standards. The frequency of the ILRTis being changed in accordance with standards reviewed and approved as compliant withAppendix J. Therefore there will be no instances where the applicable regulatory criteria are notmet.5.3 Environmental ConsiderationsThe proposed changes to the IP2 TS do not involve (i) a significant hazards consideration, (ii) asignificant change in the types or significant increase in the amounts of any effluent that may bereleased offsite, or (iii) a significant increase in individual or cumulative occupational radiationexposure. Accordingly, the proposed amendment meets the eligibility criterion for categoricalexclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), noenvironmental impact statement or environmental assessment need be prepared in connectionwith the proposed amendment.PRECEDENCEThis request is similar in nature to the license amendment authorized by the NRC on April 22,2012 for the Palisades Nuclear Plant (TAC No. ME5997, Accession Number ML1 20740081).

ATTACHMENT 2 TO NL-14-128MARKED UP TECHNICAL SPECIFICATIONS PAGES FOR PROPOSEDCHANGES REGARDING 15 YEAR CONTAINMENT ILRTChanges indicated by lineout for deletion and Bold/Italics for additionsUnit 2 Affected Pages:5.5-14ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NO. 2DOCKET NO. 50-247 Programs and Manuals5.55.5 Programs and Manuals5.5.13 Safety Function Determination Program (SFDP) (continued)The SFDP identifies where a loss of safety function exists. If a loss of safetyfunction is determined to exist by this program, the appropriate Conditions andRequired Actions of the LCO in which the loss of safety function exists are requiredto be entered. When a loss of safety function is caused by the inoperability of asingle Technical Specification support system, the appropriate Conditions andRequired Actions to enter are those of the support system.5.5.14 Containment Leakage Rate Testing Programa. A program shall establish the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordance withNEI 94-01, Revision 2A, "Industry Guidelines for ImplementingPerformance-Based Option of 10 CFR Part 50, Appendix J," October2008.the guidlinS containd in r .egulator; Guide 1.163, "P,,f.rman...Based Cont-ainmont Leak Toct Program," dated Soptombor, 1995.b. The calculated peak containment internal pressure for the design basis loss ofcoolant accident, Pa, is assumed to be the containment design pressure of47 psig.c. The maximum allowable containment leakage rate, La, at P,, and 271 OF shallbe 0.1% of containment steam air weight per day.d. Leakage rate acceptance criteria:1. Containment leakage rate acceptance criterion is 1.0 La. During the firstunit startup following testing in accordance with this program, theleakage rate acceptance criteria are < 0.60 La for the Type B and C testsand 0.75 La for Type A tests.2. Air lock testing acceptance criteria shall be established to ensure thatlimits for Type B and C testing in Technical Specification 5.5.14.d.1 aremet.(continued)INDIAN POINT 25.5- 14Amendment No. 262 ATTACHMENT 3 TO NL-14-128RISK IMPACT OF EXTENDING THE ILRT INTERVAL ASSOCIATEDWITH THE PROPOSED TECHNICAL SPECIFICATION CHANGESENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NO. 2DOCKET NO. 50-247

-allRISK ASSESSMENT FOR INDIAN POINTREGARDING THE ILRT (TYPE A)PERMANENT EXTENSION REQUESTPrepared for:0U-EntergyEntergy Services, Inc.1340 Echelon Parkway, M-ECH-492Jackson, MS 39213October 2013glneerlpg and Research, Znc.158 West Gay StreetSuite 400West Chester, PA 19380(610) 431-8260 RISK ASSESSMENT FOR INDIAN POINT REGARDING THEILRT (TYPE A) PERMANENT EXTENSION REQUESTRevision 0Prepared for:IEntergEntergy Services, Inc.1340 Echelon Parkway, M-ECH-492Jackson, MS 39213Prepared by:158 West Gay Street, Suite 400West Chester, PA 19380(610) 431-8260Document No. 0247-13-0002-4722Prepared by:Reviewed by:Approved by:Donald E. VanoverDonald E. MacLeodJeff R. GaborDate: 016 /201 3Date: //// --) 61-3Date:

Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE OF CONTENTSSection Page1.0 PURPO SE O F A NA LYSIS ................................................................................ 1-11 .1 P U R PO S E ......................................................................................... 1-11.2 BA C K G R O U N D .................................................................................. 1-11.3 ACCEPTANCE CRITERIA ........................................ 1-22 .0 M ET H O D O LO G Y .......................................................................................... 2-13 .0 G R O U N D R U LES .......................................................................................... 3-14 .0 IN P U T S ...................................................................................................... 4 -14.1 GENERAL RESOURCES AVAILABLE ....................................................... 4-14.2 PLANT-SPECIFIC INPUTS .................................................................... 4-64.3 IMPACT OF EXTENSION ON DETECTION OF COMPONENT FAILURESTHAT LEAD TO LEAKAGE (SMALL AND LARGE) ...................................... 4-134.4 IMPACT OF EXTENSION ON DETECTION OF STEEL LINER CORROSIONTHAT LEADS TO LEAKAGE ................................................................. 4-155 .0 R E S U LT S ................................................................................................... 5 -15.1 STEP 1 -QUANTIFY THE BASE-LINE RISK IN TERMS OF FREQUENCYPER REA CTO R YEA R ........................................................................... 5-25.2 STEP 2 -DEVELOP PLANT-SPECIFIC PERSON-REM DOSE (POPULATIONDOSE) PER REACTOR YEAR ................................................................. 5-65.3 STEP 3 -EVALUATE RISK IMPACT OF EXTENDING TYPE A TESTINTERVAL FROM 10-TO-15 YEARS ...................................................... 5-135.4 STEP 4 -DETERMINE THE CHANGE IN RISK IN TERMS OF LARGEEARLY RELEASE FREQUENCY ............................................................. 5-225.5 STEP 5 -DETERMINE THE IMPACT ON THE CONDITIONALCONTAINMENT FAILURE PROBABILITY ................................................ 5-225.6 SUMMARY OF INTERNAL EVENTS RESULTS .......................................... 5-235.7 EXTERNAL EVENTS CONTRIBUTION .................................................... 5-265.7.1 Indian Point 2 External Events Discussion ............................... 5-265.7.2 Indian Point 3 External Events Discussion ............................... 5-295.7.3 Additional Seism ic Risk Discussion ......................................... 5-315.7.4 External Events Impact Sum mary .......................................... 5-315.7.5 External Events Impact on ILRT Extension Assessment ............. 5-325.7.6 Alternative Approach for External Events Impact on ILRT ExtensionA ssessm ent ......................................................................... 5-365.8 CONTAINMENT OVERPRESSURE IMPACTS ON CDF ................................ 5-476 .0 S EN S IT IV IT IES ........................................................................................... 6-16.1 SENSITIVITY TO CORROSION IMPACT ASSUMPTIONS ............................ 6-16.2 EPRI EXPERT ELICITATION SENSITIVITY .............................................. 6-47 .0 C O N C LU S IO N S ........................................................................................... 7-18 .0 R E FE R E N C ES .............................................................................................. 8 -1APPENDIX A PRA TECHNICAL ADEQUACYP0247130002-4722 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyList of TablesTable 4.1-1 EPRI/NEI Containment Failure Classifications ........................................... 4-4Table 4.2-1 Level 2 Release Category Frequencies for IP2 and IP3 ............................... 4-7Table 4.2-2 Release Category Definitions from the License Renewal Effort ..................... 4-8Table 4.2-3 Population Dose per License Renewal Release Category for IP2 and IP3 ....... 4-8Table 4.2-4 Population Dose for Intact Containment Cases for IP2 and IP3 .................... 4-9Table 4.2-5 Weighted Average Population Dose for Intact Containment Case for IP2a n d IP 3 ............................................................................................. 4 -1 0Table 4.2-6a IP2 Population Dose and Population Dose Risk Organized by EPRIRelease C ategory ................................................................................ 4-11Table 4.2-6b IP3 Population Dose and Population Dose Risk Organized by EPRIRelease C ategory ................................................................................ 4-12Table 4.4-1 Steel Liner Corrosion Base Case ........................................................... 4-17Table 5.0-1 A ccident C lasses .................................................................................. 5-1Table 5.1-1 Radionuclide Release Frequencies As A Function Of Accident Class (IP2and IP3 Base C ase) ............................................................................... 5-6Table 5.2-1 IP2 and IP3 Population Dose for Population Within 50 Miles ....................... 5-8Table 5.2-2a IP2 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 3 in 10 Year ILRT Frequency .............................................. 5-9Table 5.2-2b IP3 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 3 in 10 Year ILRT Frequency ............................................ 5-11Table 5.3-1a IP2 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 1 in 10 Year ILRT Frequency ............................................ 5-14Table 5.3-1b IP3 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 1 in 10 Year ILRT Frequency ............................................ 5-16Table 5.3-2a IP2 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 1 in 15 Year ILRT Frequency ............................................ 5-18Table 5.3-2b IP3 Annual Dose As A Function Of Accident Class; Characteristic OfConditions For 1 in 15 Year ILRT Frequency ............................................ 5-20Table 5.5-1 IP2 and IP3 ILRT Conditional Containment Failure Probabilities ................. 5-23Table 5.6-1a IP2 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (IncludingAge Adjusted Steel Liner Corrosion Likelihood) ........................................ 5-24Table 5.6-1b IP3 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (IncludingP0247130002-4722ii Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyAge Adjusted Steel Liner Corrosion Likelihood) ........................................ 5-25Table 5.6-2 IP2 and IP3 ILRT Extension Comparison to Acceptance Criteria ................. 5-26Table 5.7-1 External Events Contributor Summary [20] ........................................... 5-32Table 5.7-2a IP2 3b (LERF/YR) as a Function of ILRT Frequency for Internal andExternal Events (Including Age Adjusted Steel Liner Corrosion Likelihood) .. 5-33Table 5.7-2b IP3 3b (LERF/YR) as a Function of ILRT Frequency for Internal andExternal Events (Including Age Adjusted Steel Liner Corrosion Likelihood) .. 5-33Table 5.7-3 Comparison to Acceptance Criteria Including External EventsContribution for IP2 and IP3 ................................................................. 5-35Table 5.7-4 Impact of 15-yr ILRT Extension on LERF for IP2 and IP3 .......................... 5-36Table 5.7-5a Population Dose Risk As A Function Of Accident Class (IP2 AlternativeExternal Events Base Case) .................................................................. 5-41Table 5.7-5b Population Dose Risk As A Function Of Accident Class (IP3 AlternativeExternal Events Base Case) .................................................................. 5-42Table 5.7-6a Population Dose Risk As a Function of Accident Class (IP2 AlternativeExternal Events Evaluation Characteristic of Conditions For 1 in 15 YearILRT Frequency) ................................................................................. 5-4 3Table 5.7-6b Population Dose Risk As A Function Of Accident Class (IP3 AlternativeExternal Events Evaluation Characteristic of Conditions For 1 in 15 YearILRT Frequency) ................................................................................. 5-44Table 5.7-7 Comparison to Acceptance Criteria Including Alternative External EventsEvaluation Contribution for IP2 and IP3 .................................................. 5-45Table 5.7-8 Impact of 15-yr ILRT Extension on LERF for IP2 and IP3 .......................... 5-46Table 6.1-1a Steel Liner Corrosion Sensitivity Cases for IP2 ........................................ 6-1Table 6.1-1b Steel Liner Corrosion Sensitivity Cases for IP3 ........................................ 6-3Table 6.2-1 EPRI Expert Elicitation Results ................................................................ 6-4Table 6.2-2a IP2 ILRT Cases: 3 in 10 (Base Case), 1 in 10, and 1 in 15 Yr intervals(Based on EPRI Expert Elicitation Leakage Probabilities) ............................. 6-6Table 6.2-2b IP3 ILRT Cases: 3 in 10 (Base Case), 1 in 10, and 1 in 15 Yr intervals(Based on EPRI Expert Elicitation Leakage Probabilities) ............................. 6-7Table A.2-1 Summary of Industry Peer Review Findings for the IP2 Internal EventsPRA M odel U pdate ................................................................................. A -7Table A.2-2 Summary of Industry Peer Review Findings for the IP3 Internal EventsPRA M odel Update ............................................................................ A-18P0247130002-4722iii Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy1.0 PURPOSE OF ANALYSIS1.1 PURPOSEThe purpose of this analysis is to provide an assessment of the risk associated withimplementing a permanent extension of the Indian Point Units 2 and 3 (IP2 and IP3)containment Type A integrated leak rate test (ILRT) interval from ten years to fifteen years.The risk assessment follows the guidelines from NEI 94-01 [1], the methodology outlined inEPRI TR-104285 [2], the EPRI Risk Impact Assessment of Extended Integrated Leak RateTesting Intervals [3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment(PRA) findings and risk insights in support of a request for a plant's licensing basis as outlinedin Regulatory Guide (RG) 1.174 [4], and the methodology used for Calvert Cliffs to estimatethe likelihood and risk implications of corrosion-induced leakage of steel liners goingundetected during the extended test interval [5]. The format of this document is consistentwith the intent of the Risk Impact Assessment Template for evaluating extended integratedleak rate testing intervals provided in the October 2008 EPRI final report [3].1.2 BACKGROUNDRevisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the IntegratedLeak Rate Test (ILRT) Type A surveillance testing requirements from three-in-ten years to atleast once per ten years. The revised Type A frequency is based on an acceptableperformance history defined as two consecutive periodic Type A tests at least 24 months apartin which the calculated performance leakage was less than the normal containment leakage of1.OLa (allowable leakage).The basis for a 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, andwas established in 1995 during development of the performance-based Option B to Appendix J.Section 11.0 of NEI 94-01 states that NUREG-1493 [6], "Performance-Based ContainmentLeak Test Program," provides the technical basis to support rulemaking to revise leakage ratetesting requirements contained in Option B to Appendix J. The basis consisted of qualitativeand quantitative assessments of the risk impact (in terms of increased public dose) associatedwith a range of extended leakage rate test intervals. To supplement the NRC's rulemakingbasis, NEI undertook a similar study. The results of that study are documented in ElectricPower Research Institute (EPRI) Research Project Report TR-104285 [2].The NRC report on performance-based leak testing, NUREG-1493, analyzed the effects ofcontainment leakage on the health and safety of the public and the benefits realized from thecontainment leak rate testing. In that analysis, it was determined for a representative PWRP0247130002-47221-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacyplant (i.e., Surry) that containment isolation failures contribute less than 0.1 percent to thelatent risks from reactor accidents. Because ILRTs represent substantial resourceexpenditures, it is desirable to show that extending the ILRT interval will not lead to asubstantial increase in risk from containment isolation failures to support a reduction in thetest frequency for IP2 and IP3.Earlier ILRT frequency extension submittals have used the EPRI TR-104285 [2] methodologyto perform the risk assessment. In October 2008, EPRI 1018243 [3] was issued to develop ageneric methodology for the risk impact assessment for ILRT interval extensions to 15 yearsusing current performance data and risk informed guidance, primarily NRC Regulatory Guide1.174 [4]. This more recent EPRI document considers the change in population dose, largeearly release frequency (LERF), and containment conditional failure probability (CCFP),whereas EPRI TR-104285 considered only the change in risk based on the change in populationdose. This ILRT interval extension risk assessment for IP2 and IP3 employs the EPRI 1018243methodology, with the affected System, Structure, or Component (SSC) being the primarycontainment boundary.1.3 ACCEPTANCE CRITERIAThe acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanentextension of the Type A test interval beyond that established during the Option B rulemakingof Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines asincreases in core damage frequency (CDF) less than 1.OE-06 per reactor year and increases inlarge early release frequency (LERF) less than 1.OE-07 per reactor year. Note that a separatediscussion in Section 5.8 confirms that the CDF is not impacted by the proposed change for IP2and IP3. Therefore, since the Type A test does not impact CDF for IP2 and IP3, the relevantcriterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 1.OE-06per reactor year, provided that the total LERF from all contributors (including external events)can be reasonably shown to be less than 1.OE-05 per reactor year. RG 1.174 discussesdefense-in-depth and encourages the use of risk analysis techniques to help ensure and showthat key principles, such as the defense-in-depth philosophy, are met. Therefore, the increasein the conditional containment failure probability (CCFP) is also calculated to help ensure thatthe defense-in-depth philosophy is maintained.With regard to population dose, examinations of NUREG-1493 and Safety Evaluation Reports(SERs) for one-time interval extension (summarized in Appendix G of [3]) indicate a range ofincremental increases in population dose1 that have been accepted by the NRC. The range of1 The one-time extensions assumed a large leak (EPRI class 3b) magnitude of 35La, whereas thisP0247130002-47221-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacyincremental population dose increases is from _<0.01 to 0.2 person-rem/yr and 0.002 to 0.46%of the total accident dose. The total doses for the spectrum of all accidents (Figure 7-2 ofNUREG-1493) result in health effects that are at least two orders of magnitude less than theNRC Safety Goal Risk. Given these perspectives, the NRC SER on this issue [7] defines a smallincrease in population dose as an increase of 5 1.0 person-rem per year, or 51 0% of the totalpopulation dose, whichever is less restrictive for the risk impact assessment of the extendedILRT intervals. This definition has been adopted by the IP2/IP3 analysis.The acceptance criteria are summarized below.1. The estimated risk increase associated with permanently extending the ILRTsurveillance interval to 15 years must be demonstrated to be small. (Note thatRegulatory Guide 1.174 defines very small changes in risk as increases in CDFless than 1.OE-6 per reactor year and increases in LERF less than 1.OE-7 perreactor year. Since the type A ILRT test is not expected to impact CDF forIndian Point, the relevant risk metric is the change in LERF. Regulatory Guide1.174 also defines small risk increase as a change in LERF of less than 1.OE-6reactor year.) Therefore, a small change in risk for this application is definedas a LERF increase of less than 1.OE-6.2. Per the NRC SE, a small increase in population dose is also defined as anincrease in population dose of less than or equal to either 1.0 person-rem peryear or 1 percent of the total population dose, whichever is less restrictive.3. In addition, the SE notes that a small increase in Conditional ContainmentFailure Probability (CCFP) should be defined as a value marginally greater thanthat accepted in previous one-time 15-year ILRT extension requests (typicallyabout 1% or less, with the largest increase being 1.2%). This would requirethat the increase in CCFP be less than or equal to 1.5 percentage points.analysis uses lOOLa.P0247130002-47221-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy2.0 METHODOLOGYA simplified bounding analysis approach consistent with the EPRI methodology is used forevaluating the change in risk associated with increasing the test interval to fifteen years [3].The analysis uses results from a Level 2 analysis of core damage scenarios from the currentIP2 and IP3 PRA analyses of record and the subsequent containment responses to establishthe various fission product release categories including the release size.The six general steps of this assessment are as follows:1. Quantify the baseline risk in terms of the frequency of events (per reactor year) foreach of the eight containment release scenario types identified in the EPRI report [3].2. Develop plant-specific population dose rates (person-rem per reactor year) for each ofthe eight containment release scenario types from plant specific consequence analyses.3. Evaluate the risk impact (i.e., the change in containment release scenario typefrequency and population dose) of extending the ILRT interval to fifteen years.4. Determine the change in risk in terms of Large Early Release Frequency (LERF) inaccordance with RG 1.174 and compare this change with the acceptance guidelines ofRG 1.174 [4].5. Determine the impact on the Conditional Containment Failure Probability (CCFP)6. Evaluate the sensitivity of the results to assumptions in the liner corrosion analysis andto variations in the fractional contributions of large isolation failures (due to linerbreach) to LERF.Furthermore," Consistent with the previous industry containment leak risk assessments, the IP2and IP3 assessment uses population dose as one of the risk measures. The otherrisk measures used in the IP2 and IP3 assessment are the conditional containmentfailure probability (CCFP) for defense-in-depth considerations, and change in LERF todemonstrate that the acceptance guidelines from RG 1.174 are met." This evaluation for IP2 and IP3 uses ground rules and methods to calculate changesin the above risk metrics that are consistent with those outlined in the current EPRImethodology [3].P0247130002-47222-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy3.0 GROUND RULESThe following ground rules are used in the analysis:" The IP2 and IP3 Level 1 and Level 2 internal events PRA models providerepresentative core damage frequency and release category frequency distributionsto be utilized in this analysis." It is appropriate to use the IP2 and IP3 internal events PRA model as a gauge toeffectively describe the risk change attributable to the ILRT extension. It isreasonable to assume that the impact from the ILRT extension (with respect topercent increases in population dose) will not substantially differ if external eventswere to be included in the calculations; however, external events have beenaccounted for in the analysis based on the available information from the IP2 and IP3IPEEEs [8, 9] as reported and used in the IP2 and IP3 SAMA analysis performed aspart of the License Renewal efforts as described in Section 5.7." Dose results for the containment failures modeled in the PRA can be characterized byinformation that was prepared to support the SAMA analysis as part of the LicenseRenewal effort [10]. This information is supplemented with revised calculations [11]for the base case containment intact scenarios which are critical for use in the ILRTextension assessment.* Accident classes describing radionuclide release end states and their definitions areconsistent with the EPRI methodology [3] and are summarized in Section 4.2." The representative containment leakage for Class 1 sequences is 1La. Class 3accounts for increased leakage due to Type A inspection failures." The representative containment leakage for Class 3a is 10 La and for Class 3bsequences is 10OLa, based on the recommendations in the latest EPRI report [3] andas recommended in the NRC SE on this topic [7]. It should be noted that this ismore conservative than the earlier previous industry ILRT extension requests, whichutilized 35La for the Class 3b sequences." Based on the EPRI methodology and the NRC SE, the Class 3b sequences arecategorized as LERF and the increase in Class 3b sequences is used as a surrogatefor the ALERF metric." The impact on population doses from containment bypass scenarios is not altered bythe proposed ILRT extension, but is accounted for in the EPRI methodology as aseparate entry for comparison purposes. Since the containment bypass contributionto population dose is fixed, no changes on the conclusions from this analysis willresult from this separate categorization." The reduction in ILRT frequency does not impact the reliability of containmentisolation valves to close in response to a containment isolation signal.* The use of the estimated 2035 population data from the MACCS2 off-siteconsequence runs [10, 11] is appropriate for this analysis. This assumption isconsistent with that made in the SAMA analysis.* An evaluation of the risk impact of the ILRT on shutdown risk is addressed using thegeneric results from EPRI TR-105189 [12].P0247130002-47223-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy4.0 INPUTSThis section summarizes the general resources available as input (Section 4.1) and the plantspecific resources required (Section 4.2).4.1 GENERAL RESOURCES AVAILABLEVarious industry studies on containment leakage risk assessment are briefly summarized here:1. NUREG/CR-3539 [13]2. NUREG/CR-4220 [14]3. NUREG-1273 [15]4. NUREG/CR-4330 [16]5. EPRI TR-105189 [12]6. NUREG-1493 [6]7. EPRI TR-104285 [2]8. Calvert Cliffs liner corrosion analysis [5]9. EPRI 1018243 [3]10. NRC Final Safety Evaluation [7]The first study is applicable because it provides one basis for the threshold that could be usedin the Level 2 PRA for the size of containment leakage that is considered significant and to beincluded in the model. The second study is applicable because it provides a basis of theprobability for significant pre-existing containment leakage at the time of a core damageaccident. The third study is applicable because it is a subsequent study to NUREG/CR-4220that undertook a more extensive evaluation of the same database. The fourth study providesan assessment of the impact of different containment leakage rates on plant risk. The fifthstudy provides an assessment of the impact on shutdown risk from ILRT test intervalextension. The sixth study is the NRC's cost-benefit analysis of various alternative approachesregarding extending the test intervals and increasing the allowable leakage rates forcontainment integrated and local leak rate tests. The seventh study is an EPRI study of theimpact of extending ILRT and LLRT test intervals on at-power public risk. The eighth studyaddresses the impact of age-related degradation of the containment liners on ILRT evaluations.EPRI 1018243 complements the previous EPRI report and provides the results of an expertelicitation process to determine the relationship between pre-existing containment leakageprobability and magnitude. Finally, the NRC Safety Evaluation (SE) documents the acceptanceby the NRC of the proposed methodology with a few exceptions. These exceptions (associatedwith the ILRT Type A tests) were addressed in the Revision 2-A of NEI 94-01 and the finalversion of the updated EPRI report [3], which was used for this application.P0247130002-47224-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyNUREG/CR-3539 [131Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leakrates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 [31] asthe basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rateson LWR accident risks is relatively small.NUREG/CR-4220 [141NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985.The study reviewed over two thousand LERs, ILRT reports and other related records tocalculate the unavailability of containment due to leakage. It assessed the "large" containmentleak probability to be in the range of 1E-3 to 1E-2, with 5E-3 identified as the point estimatebased on 4 events in 740 reactor years and conservatively assuming a one-year duration foreach event.NUREG-1273 r151A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of theNUREG/CR-4220 database. This assessment noted that about one-third of the reported eventswere leakages that were immediately detected and corrected. In addition, this study notedthat local leak rate tests can detect "essentially all potential degradations" of the containmentisolation system.NUREG/CR-4330 [161NUREG/CR-4330 is a study that examined the risk impacts associated with increasing theallowable containment leakage rates. The details of this report have no direct impact on themodeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakagerate and the ILRT test interval extension study focuses on the frequency of testing intervals.However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 andother similar containment leakage risk studies:"...the effect of containment leakage on overall accident risk is small since risk isdominated by accident sequences that result in failure or bypass ofcontainment."P0247130002-47224-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyEPRI TR-105189 r121The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessmentbecause this EPRI study provides insight regarding the impact of containment testing onshutdown risk. This study performed a quantitative evaluation (using the EPRI ORAMsoftware) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT andLLRT test intervals on shutdown risk.The result of the study concluded that a small but measurable safety benefit (shutdown CDFreduced by 1.OE-8/yr to 1.0E-7/yr) is realized from extending the test intervals from 3 per 10years to 1 per 10 years.NUREG-1493 [6]NUREG-1493 is the NRC's cost-benefit analysis for proposed alternatives to reducecontainment leakage testing frequencies and/or relax allowable leakage rates. The NRCconclusions are consistent with other similar containment leakage risk studies:" Reduction in ILRT frequency from 3 per 10 years to 1 per 20 years results in an"imperceptible" increase in risk." Given the insensitivity of risk to the containment leak rate and the small fraction ofleak paths detected solely by Type A testing, increasing the interval betweenintegrated leak rate tests is possible with minimal impact on public risk.EPRI TR-104285 r2lExtending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study),the EPRI TR-104285 study is a quantitative evaluation of the impact of extending IntegratedLeak Rate Test (ILRT) and (Local Leak Rate Test) LLRT test intervals on at-power public risk.This study combined IPE Level 2 models with NUREG-1150 [17] Level 3 population dosemodels to perform the analysis. The study also used the approach of NUREG-1493 [6] incalculating the increase in pre-existing leakage probability due to extending the ILRT and LLRTtest intervals.EPRI TR-104285 used a simplified Containment Event Tree to subdivide representative coredamage sequences into eight categories of containment response to a core damage accident:1. Containment intact and isolated2. Containment isolation failures due to support system or active failures3. Type A (ILRT) related containment isolation failures4. Type B (LLRT) related containment isolation failures5. Type C (LLRT) related containment isolation failuresP0247130002-47224-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy6. Other penetration related containment isolation failures7. Containment failure due to core damage accident phenomena8. Containment bypassConsistent with the other containment leakage risk assessment studies, this study concluded:"These study results show that the proposed CLRT [containment leak ratetests] frequency changes would have a minimal safety impact. The change inrisk determined by the analyses is small in both absolute and relative terms..."Release Category DefinitionsTable 4.1-1 defines the accident classes used in the ILRT extension evaluation, which isconsistent with the EPRI methodology [3]. These containment failure classifications are usedin this analysis to determine the risk impact of extending the Containment Type A test intervalas described in Section 5 of this report.TABLE 4.1-1EPRI/NEI CONTAINMENT FAILURE CLASSIFICATIONSCLASS] DESCRIPTION1 Containment remains intact including accident sequences that do not lead tocontainment failure in the long term. The release of fission products (andattendant consequences) is determined by the maximum allowable leakagerate values La, under Appendix J for that plant2 Containment isolation failures (as reported in the IPEs) include those accidentsin which there is a failure to isolate the containment.3 Independent (or random) isolation failures include those accidents in which thepre-existing isolation failure to seal (i.e., provide a leak-tight containment) isnot dependent on the sequence in progress.4 Independent (or random) isolation failures include those accidents in which thepre-existing isolation failure to seal is not dependent on the sequence inprogress. This class is similar to Class 3 isolation failures, but is applicable tosequences involving Type B tests and their potential failures. These are theType B-tested components that have isolated but exhibit excessive leakage.5 Independent (or random) isolation failures include those accidents in which thepre-existing isolation failure to seal is not dependent on the sequence inprogress. This class is similar to Class 4 isolation failures, but is applicable tosequences involving Type C tests and their potential failures.6 Containment isolation failures include those leak paths covered in the planttest and maintenance requirements or verified per in service inspection andtesting (ISI/IST) program.7 Accidents involving containment failure induced by severe accidentphenomena. Changes in Appendix J testing requirements do not impact theseaccidents.P0247130002-47224-4 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.1-1EPRI/NEI CONTAINMENT FAILURE CLASSIFICATIONSCLASS DESCRIPTION8 Accidents in which the containment is bypassed (either as an initial conditionor induced by phenomena) are included in Class 8. Changes in Appendix Jtesting requirements do not impact these accidents.Calvert Cliffs Liner Corrosion Analysis [51This submittal to the NRC describes a method for determining the change in likelihood, due toextending the ILRT, of detecting liner corrosion, and the corresponding change in risk. Themethodology was developed for Calvert Cliffs in response to a request for additionalinformation regarding how the potential leakage due to age-related degradation mechanismswas factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffsanalysis was performed for a concrete cylinder and dome and a concrete basemat, each with asteel liner. IP2 and IP3 have a similar type of containment.EPRI 1018243 [31This report presents a risk impact assessment for extending integrated leak rate test (ILRT)surveillance intervals to 15 years. This risk impact assessment complements the previousEPRI report, TR-104285, Risk Impact Assessment of Revised Containment Leak Rate TestingIntervals. The earlier report considered changes to local leak rate testing intervals as well aschanges to ILRT testing intervals. The original risk impact assessment considers the change inrisk based on population dose, whereas the revision considers dose as well as large earlyrelease frequency (LERF) and conditional containment failure probability (CCFP). This reportdeals with changes to ILRT testing intervals and is intended to provide bases for supportingchanges to industry and regulatory guidance on ILRT surveillance intervals.The risk impact assessment using the Jeffrey's Non-Informative Prior statistical method isfurther supplemented with a sensitivity case using expert elicitation performed to addressconservatisms. The expert elicitation is used to determine the relationship between pre-existing containment leakage probability and magnitude. The results of the expert elicitationprocess from this report are used as a separate sensitivity investigation for the IP2 and IP3analysis presented here in Section 6.2.P0247130002-47224-5 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyNRC Safety Evaluation Report r7]This SE documents the NRC staff's evaluation and acceptance of NEI TR 94-01, Revision 2, andEPRI Report No. 1009325, Revision 2, subject to the limitations and conditions identified in theSE and summarized in Section 4.0 of the SE. These limitations (associated with the ILRT TypeA tests) were addressed in the Revision 2-A of NEI 94-01 which are also included in Revision3-A of NEI 94-01 [1] and the final version of the updated EPRI report [3]. Additionally, the SEclearly defined the acceptance criteria to be used in future Type A ILRT extension riskassessments as delineated previously in the end of Section 1.3.4.2 PLANT-SPECIFIC INPUTSThe IP2 and IP3 specific information used to perform this ILRT interval extension riskassessment includes the following:* Level 1 and Level 2 PRA model quantification results [18, 19]* Population dose within a 50-mile radius for various release categories [10, 11]IP2 and IP3 Internal Events Core Damage FrequenciesThe current IP2 and IP3 Internal Events PRA analyses of record are based on an event tree /linked fault tree model characteristic of the as-built, as-operated plant. Based on the resultsfound in Tables J1.6-2 of Reference [18] and Reference [19], the internal events Level 1 PRAcore damage frequency (CDF) is 1.17E-05/yr for IP2 and 1.48E-05/yr for IP3.IP2 and IP3 Internal Events Release Category FrequenciesThe Level 2 release category frequencies were developed from the contributions to CDF forthose analyzed containment failure modes that were documented in Tables J1.6-2 and TablesJ1.7-4 for IP2 and IP3 of Reference [18] and Reference [19], respectively. Table 4.2-1summarizes the pertinent IP2 and IP3 results in terms of end-states where a representativerelease category is assigned for each end-state. The total Large Early Release Frequency(LERF) in Table 4.2-1 is 1.16E-06/yr for IP2 and 1.25E-06/yr for IP3. The individual releasecategory frequencies are utilized here to provide the necessary delineation for the ILRT riskassessment with the corresponding EPRI class for each release category. A discussion of theavailable population dose information for various release categories follows this table.P0247130002-47224-6 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.2-1LEVEL 2 RELEASE CATEGORY FREQUENCIES FOR IP2 AND IP3RELEASE CATEGORY DESCRIPTION INDIAN POINT 2 INDIAN POINT 3(FREQUENCY/YR) (FREQUENCY/YR)No Containment Failure 7.86E-06 1.13E-05Late Release 2.71E-06 2.17E-06Low to Moderate Early Release 4.66E-09 1.17E-07High Early Release (LERF) 1.16E-06 1.25E-06LERF: Containment Bypass (SGTRInitiating Events) 9.58E-07 9.19E-07LERF: Containment Bypass (ISLOCA) 2.77E-08 1.93E-07LERF: Containment Bypass (InducedSGTR events) 8.72E-08 5.78E-08LERF: Containment Isolation Failure 1.11E-08 3.99E-09LERF: Energetic Containment Failures 6.90E-08 7.14E-08Total: 1.17E-05 1.48E-05IP2 and IP3 Population Dose InformationIn the License Renewal analysis for IP2 and IP3 [20], the release categories considered themagnitude of the radionuclide release, e.g., concentration of cesium iodide (CsI), and the timeof the release. Table 4.2-2 shows how the different release categories were organized for thelicense renewal effort. While that breakdown was appropriate for that submittal, thebreakdown in Table 4.2-1 is sufficient for this ILRT extension risk assessment.P0247130002-47224-7 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.2-2RELEASE CATEGORY DEFINITIONS FROM THE LICENSE RENEWAL EFFORTRELEASE SEVERITY SOURCE TERMRELEASE TIMING RELEASE FRACTIONCLASSIFICATION TIME OF RELEASE CLASSIFICATION PERCENT CSI INCATEGORY (NOBLE GASES OR CATEGORY RELEASECSI)Late (L) > 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> High (H) > 10Moderate (M) 1 to 10Early (E) < 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Low (L) 0.1 to 1Low-Low (LL) 0.01 to 0.1No Containment < 0.01 (Little to NoFailure (NCF) Release)The population dose results from latest relevant License Renewal submittal [10] form the basisof the initial ILRT assessment using the latest available release category frequency informationas described above. The results for IP2 are taken from Table 5 of Reference [10] and theresults for IP3 are taken from Table 6 of Reference [10]. Those population dose results arereproduced in Table 4.2-3 converted to the corresponding values in person-rem (i.e., 100 *person-sv) used for this analysis.TABLE 4.2-3POPULATION DOSE PER LICENSE RENEWAL RELEASE CATEGORY FOR IP2 AND IP3RELEASE CATEGORY DESCRIPTION INDIAN POINT 2 INDIAN POINT 3(PERSON-REM) (PERSON-REM)No Containment Failure (NCF) 4.75E+03 8.04E+03Early High 6.51E+07 5.08E+07Early Medium 1.94E+07 2.OOE+07Early Low 7.93E+06 5.21E+06Late High 1.63E+07 1.63E+07Late Medium 6.87E+06 6.85E+06Late Low 1.61E+06 1.61E+06Late Low-Low 1.38E+06 1.38E+06P0247130002-47224-8 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacySince the ILRT methodology is based on multipliers to a bounding case which is representativeof an allowable leakage of 1.OLa, the NCF case from the License Renewal effort, whichrepresents a best estimate release, could not be used. As a result, additional analyses wererequired for the ILRT assessment to be consistent with the methodology employed. Table4.2-4 shows the results of four different potential case runs to provide a representative 1.0Larelease [11]. Note that for the containment intact case, given the similarities between IP2 andIP3, the results are assumed to be applicable to both units. These case results arerepresentative of the 1.OLa release as required by the ILRT methodology.TABLE 4.2-4POPULATION DOSE FOR INTACT CONTAINMENT CASES FOR IP2 AND IP3RELEASE CATEGORY DESCRIPTION INDIAN POINT 2 INDIAN POINT 3(PERSON-REM) (PERSON-REM)Intact Scenario #1 (Vessel Breach Occurs,Containment Fan Coolers Available) 8.28E+04 8.28E+04Intact Scenario #2 (Vessel Breach Occurs,Containment Sprays Available) 1.59E+04 1.59E+04Intact Scenario #3 (Vessel Breach Occurs,Fan Coolers and Sprays Available) 1.32E+04 1.32E+04Intact Scenario #4 (No Vessel Breach,Containment Fan Coolers Available) 2.94E+04 2.94E+04Based on a review of cutsets associated with the intact containment end state, anapportionment of the intact containment associated release categories was made. First, it wasnoted that containment sprays were not failed in more than 99% of the intact containmentcases for both IP2 and IP3, but their use could only be definitively declared in Medium andLarge LOCA scenarios or when vessel breach occurs (i.e., other cases with fan coolers availableand no vessel breach are unlikely to reach the automatic containment spray initiation set pointof 24 psig for IP2 and 22 psig for IP3). For IP2 about 68% of the intact containment casesalso involved no vessel breach, and for IP3 about 63% of the intact containment casesinvolved no vessel breach. For IP2 and IP3, the medium and large LOCA contribution to theintact containment case was about 10%. Therefore, it was conservatively assumed that just10% of the intact containment cases could be represented by a case with containment spraysavailable (i.e., intact scenario #2 from Table 4.2-4). Of the remaining 90%, based on theP0247130002-47224-9 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacycontribution from no vessel breach scenarios noted above, it was assumed that about 60% ofthe cases involved scenarios with no vessel failure and about 30% involved scenarios wherevessel failure occurred for both IP2 and IP3. Intact scenario #4 from Table 4.2-4 is then usedas a representative case for the no vessel failure scenarios, and intact scenario #1 is thenconservatively used as a representative case for the remaining vessel failure scenarios.Although sprays are likely available in those scenarios, the SAMG procedures may limit theiruse based on hydrogen detonation concerns. This leads to an overall weighted averagepopulation dose for the intact containment case as shown in Table 4.2-5. This weightedaverage population dose of 4.41E+04 person-rem is used in the remainder of the calculationsusing the ILRT methodology.TABLE 4.2-5WEIGHTED AVERAGE POPULATION DOSE FOR INTACT CONTAINMENT CASE FORIP2 AND IP3RELEASE CATEGORY DESCRIPTION PERCENT POPULATION DOSECONTRIBUTION (PERSON-REM)Intact Scenario #1 (Vessel Breach Occurs,Containment Fan Coolers Available) 30% 8.28E+04Intact Scenario #2 (Vessel Breach Occurs,Containment Sprays Available) 10% 1.59E+04Intact Scenario #3 (Vessel Breach Occurs,Fan Coolers and Sprays Available) N/A 1.32E+04Intact Scenario #4 (No Vessel Breach,Containment Fan Coolers Available) 60% 2.94E+040.3 * (8.28E+04) +Weighted Average 0.1 * (1.59E+04) +0.6 * (2.94E+04) 4.41E+04Population Dose Risk CalculationsThe next step is to take the frequency information from Table 4.2-1, assign each category tothe relevant EPRI release category class from Table 4.1-1, and then associate a representativepopulation dose from Table 4.2-3 or Table 4.2-5 for each release category. Table 4.2-6a liststhe population dose risk and average population dose organized by EPRI release category forIP2, including the delineation of early and late frequencies for Class 7, and a delineation ofSGTR and ISLOCA frequencies for Class 8. Note that the population dose risk (Column 4 ofTable 4.2-6a) was found by multiplying the release category frequency (Column 2 of TableP0247130002-47224-10 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy4.2-6a) by the associated population dose (Column 3 of Table 4.2-6a). The correspondinginformation for IP3 is shown in Table 4.2-6b. Note that only the applicable EPRI releasecategories at this point are shown in the tables (i.e., the Class 3 frequencies are derived laterand the Class 4, 5, and 6 frequencies are not utilized in the EPRI methodology for the ILRTextension risk assessment).IP2 POPULATIONTABLE 4.2-6ADOSE AND POPULATION DOSE RISK ORGANIZEDBY EPRI RELEASE CATEGORYEPRI RELEASE CATEGORY RELEASE ASSIGNED POPULATION DOSEAND DESCRIPTION FREQUENCY POPULATION RISK (PERSON-(1/YR) DOSE (PERSON- REM/YR)REM)1: Containment intact 7.86E-06 4.41E+04 3.47E-01[Weighted AverageFrom Table 4.2-5]2: Large containment 1.11E-08 6.51E+07 7.23E-01isolation failures [Early High FromTable 4.2-3]7-CFE: Phenomena-induced 4.66E-09 1.94E+07 9.04E-02containment failures [Early Medium From(Early-non LERF) Table 4.2-3]7-CFE: Phenomena-induced 6.90E-08 6.51E+07 4.49E+00containment failures [Early High From(Early LERF) Table 4.2-3]7-CFL: Phenomena- 2.71E-06 6.87E+06 1.86E+01induced containment [Late Medium Fromfailures (Late) Table 4.2-3](1)8-SGTR: Containment 1.05E-06 6.51E+07 6.80E+01bypass (SGTR) [Early High FromTable 4.2-3]8-ISLOCA: Containment 2.77E-08 6.51E+07 1.80E+00bypass (ISLOCA) [Early High FromI_ Table 4.2-3]Total: 1.17E-05 94.12) Although the current model does not distinguish between the different late release categories,the weighted average late release from the License Renewal was within 10% of the LateMedium population dose. The use of the Late Medium population dose for this releasecategory was therefore deemed appropriate for the ILRT assessment.P0247130002-47224-11 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.2-6BIP3 POPULATION DOSE AND POPULATION DOSE RISK ORGANIZEDBY EPRI RELEASE CATEGORYEPRI RELEASE CATEGORY RELEASE ASSIGNED POPULATION DOSEAND DESCRIPTION FREQUENCY POPULATION RISK (PERSON-(1/YR) DOSE (PERSON- REM/YR)REM)1: Containment intact 1.13E-05 4.41E+04 4.98E-01[Weighted AverageFrom Table 4.2-5]2: Large containment 3.99E-09 5.08E+07 2.03E-01isolation failures [Early High FromTable 4.2-3]7-CFE: Phenomena-induced 1.17E-07 2.OOE+07 2.34E+00containment failures [Early Medium From(Early-non LERF) Table 4.2-3]7-CFE: Phenomena-induced 7.14E-08 5.08E+07 3.63E+00containment failures [Early High From(Early LERF) Table 4.2-3]7-CFL: Phenomena-induced 2.17E-06 6.85E+06 1.49E+01containment failures [Late Medium From(Late) Table 4.2-3](1)8-SGTR: Containment 9.77E-07 5.08E+07 4.96E+01bypass (SGTR) [Early High FromI Table 4.2-3]8-ISLOCA: Containment 1.93E-07 5.08E+07 9.80E+00bypass (ISLOCA) [Early High FromTable 4.2-3]Total: 1.48E-05 80.96(1) Although the current model does not distinguish between the different late release categories,the weighted average late release from the License Renewal was within 10% of the LateMedium population dose. The use of the Late Medium population dose for this releasecategory was therefore deemed appropriate for the ILRT assessment.The frequencies for the severe accident classes defined in Table 4.1-1 are developed for IP2and IP3 based on the assignments shown above in Tables 4.2-6a and 4.2-6b. Then, thefrequencies for Classes 3a and 3b can be determined with that portion removed from Class 1.This step in the process is described in Section 4.3. Furthermore, adjustments are made tothe Class 3b as well as Class 1 frequencies to account for the impact of undetected corrosion ofthe steel liner per the methodology described in Section 4.4.P0247130002-47224-12 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy4.3 IMPACT OF EXTENSION ON DETECTION OF COMPONENT FAILURES THAT LEAD TOLEAKAGE (SMALL AND LARGE)The ILRT can detect a number of component failures such as liner breach and failure of somesealing surfaces, which can lead to leakage. The proposed ILRT test interval extension mayinfluence the conditional probability of detecting these types of failures. To ensure that thiseffect is properly accounted for, the EPRI Class 3 accident class as defined in Table 4.1-1 isdivided into two sub-classes representing small and large leakage failures. These subclassesare defined as Class 3a and Class 3b, respectively.The probability of the EPRI Class 3a failures may be determined, consistent with the latestEPRI guidance [3], as the mean failure estimated from the available data (i.e., 2 "small"failures that could only have been discovered by the ILRT in 217 tests leads to a2/217=0.0092 mean value). For Class 3b, consistent with latest available EPRI data, a non-informative prior distribution is assumed for no "large" failures in 217 tests (i.e., 0.5/(217+1)= 0.0023).The EPRI methodology contains information concerning the potential that the calculated deltaLERF values for several plants may fall above the "very small change" guidelines of the NRCregulatory guide 1.174. This information includes a discussion of conservatisms in thequantitative guidance for delta LERF. EPRI describes ways to demonstrate that, using plant-specific calculations, the delta LERF is smaller than that calculated by the simplified method.The methodology states:"The methodology employed for determining LERF (Class 3b frequency)involves conservatively multiplying the CDF by the failure probability for thisclass (3b) of accident. This was done for simplicity and to maintainconservatism. However, some plant-specific accident classes leading tocore damage are likely to include individual sequences that either mayalready (independently) cause a LERF or could never cause a LERF, and arethus not associated with a postulated large Type A containment leakagepath (LERF). These contributors can be removed from Class 3b in theevaluation of LERF by multiplying the Class 3b probability by only thatportion of CDF that may be impacted by type A leakage."The application of this additional guidance to the analysis for IP2 and IP3 (as detailed inSection 5) means that the Class 2, Class 7, and Class 8 LERF sequences are subtracted fromthe CDF that is applied to Class 3b. To be consistent, the same change is made to the Class3a CDF, even though these events are not considered LERF. Note that Class 2 events refer tosequences with a large pre-existing containment isolation failure that lead to LERF, a subset ofClass 7 events are LERF sequences due to an early containment failure from energeticphenomena, and Class 8 event are containment bypass events that contribute to LERF.P0247130002-47224-13 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyConsistent with the EPRI methodology [3], the change in the leak detection probability can beestimated by comparing the average time that a leak could exist without detection. Forexample, the average time that a leak could go undetected with a three-year test interval is1.5 years (3 yr / 2), and the average time that a leak could exist without detection for a ten-year interval is 5 years (10 yr / 2). This change would lead to a non-detection probability thatis a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRTtesting, given a 10-year vs. a 3-yr interval. Correspondingly, an extension of the ILRT intervalto fifteen years can be estimated to lead to about a factor of 5.0 (7.5/1.5) increase in the non-detection probability of a leak.IP2 and IP3 Past ILRT ResultsThe surveillance frequency for Type A testing in NEI 94-01 under option B criteria is at leastonce per ten years based on an acceptable performance history (i.e., two consecutive periodicType A tests at least 24 months apart) where the calculated performance leakage rate was lessthan 1.OLa, and in compliance with the performance factors in NEI 94-01, Section 11.3. Basedon the successful completion of two consecutive ILRTs at IP2 and IP3, the current ILRT intervalis once per ten years. Note that the probability of a pre-existing leakage due to extending theILRT interval is based on the industry-wide historical results as noted in the EPRI guidancedocument [3].EPRI MethodoloqyThis analysis uses the approach outlined in the EPRI Methodology [3]. The six steps of themethodology are:1. Quantify the baseline (three-year ILRT frequency) risk in terms of frequency perreactor year for the EPRI accident classes of interest.2. Develop the baseline population dose (person-rem, from the plant PRA or IPE, orcalculated based on leakage) for the applicable accident classes.3. Evaluate the risk impact (in terms of population dose rate and percentile change inpopulation dose rate) for the interval extension cases.4. Determine the risk impact in terms of the change in LERF and the change in CCFP.5. Consider both internal and external events.6. Evaluate the sensitivity of the results to assumptions in the liner corrosion analysis.The first three steps of the methodology deal with calculating the change in dose. The changein dose is the principal basis upon which the Type A ILRT interval extension was previouslygranted and is a reasonable basis for evaluating additional extensions. The fourth step in theP0247130002-47224-14 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacymethodology calculates the change in LERF and compares it to the guidelines in RegulatoryGuide 1.174. Because there is no change in CDF for IP2 and IP3, the change in LERF formsthe quantitative basis for a risk informed decision per current NRC practice, namely RegulatoryGuide 1.174. The fourth step of the methodology calculates the change in containment failureprobability, referred to as the conditional containment failure probability, CCFP. The NRC hasidentified a CCFP of less than 1.5% as the acceptance criteria for extending the Type A ILRTtest intervals as the basis for showing that the proposed change is consistent with the defensein depth philosophy [7]. As such, this step suffices as the remaining basis for a risk informeddecision per Regulatory Guide 1.174. Step 5 takes into consideration the additional risk due toexternal events, and Step 6 investigates the impact on results due to varying the assumptionsassociated with the liner corrosion rate and failure to visually identify pre-existing flaws.4.4 IMPACT OF EXTENSION ON DETECTION OF STEEL LINER CORROSION THAT LEADSTO LEAKAGEAn estimate of the likelihood and risk implications of corrosion-induced leakage of the steelliners occurring and going undetected during the extended test interval is evaluated using themethodology from the Calvert Cliffs liner corrosion analysis [5]. The Calvert Cliffs analysis wasperformed for a concrete cylinder and dome and a concrete basemat, each with a steel liner.IP2 and IP3 have similar containment types.The following approach is used to determine the change in likelihood, due to extending theILRT, of detecting corrosion of the containment steel liner. This likelihood is then used todetermine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the followingissues are addressed:* Differences between the containment basemat and the containment cylinder anddome" The historical steel liner flaw likelihood due to concealed corrosion* The impact of aging" The corrosion leakage dependency on containment pressure" The likelihood that visual inspections will be effective at detecting a flawP0247130002-47224-15 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyAssumptions" A half failure is assumed for the basemat concealed liner corrosion due to lack ofidentified failures.* The two corrosion events over a 5.5 year data period are used to estimate the linerflaw probability in the Calvert Cliffs analysis and are assumed to be applicable to theIP2 and IP3 containment analysis. These events, one at North Anna Unit 2 and oneat Brunswick Unit 2, were initiated from the non-visible (backside) portion of thecontainment liner. It is noted that two additional events have occurred in recentyears (based on a data search covering approximately 9 years documented inReference [21]). In November 2006, the Turkey Point 4 containment building linerdeveloped a hole when a sump pump support plate was moved. In May 2009, a holeapproximately 3/8" by 1" in size was identified in the Beaver Valley 1 containmentliner. For risk evaluation purposes, these two more recent events occurring over a 9year period are judged to be adequately represented by the two events in the 5.5year period of the Calvert Cliffs analysis incorporated in the EPRI guidance (SeeTable 4.4-1, Step 1)." Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumedto double every five years. This is based solely on judgment and is included in thisanalysis to address the increased likelihood of corrosion as the steel liner ages (SeeTable 4.4-1, Steps 2 and 3). Sensitivity studies are included that address doublingthis rate every two years and every ten years.* In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reachingthe outside atmosphere given that a liner flaw exists was estimated as 1.11% for thecylinder and dome region, and 0.11% (10% of the cylinder failure probability) for thebasemat. These values were determined from an assessment of the probability ofcontainment failure versus containment pressure, and the selected values areconsistent with a pressure that corresponds to the ILRT target pressure of 37 psig.For IP2 and IP3, the containment failure probabilities are less than these values at47 psig, which is the containment design pressure [18, 19]. The probabilities of 1%for the cylinder and dome, and 0.1% for the basemat, albeit conservative, are usedin this analysis. Sensitivity studies are included that increase and decrease theprobabilities by an order of magnitude (See Table 4.4-1, Step 4).* Consistent with the Calvert Cliffs analysis, a 5% visual inspection detection failurelikelihood given the flaw is visible and a total detection failure likelihood of 10% isused for the containment cylinder and dome. For the containment basemat, 100% isassumed unavailable for visual inspection. To date, all liner corrosion events havebeen detected through visual inspection (See Table 4.4-1, Step 5). Sensitivitystudies are included that evaluate total detection failure likelihood of 5% and 15%,respectively.* Consistent with the Calvert Cliffs analysis, all non-detectable containment failuresare assumed to result in early releases. This approach avoids a detailed analysis ofcontainment failure timing and operator recovery actions.P0247130002-47224-16 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.4-1STEEL LINER CORROSION BASE CASESTEP DESCRIPTION CONTAINMENT CONTAINMENTCYLINDER AND DOME BASEMATHistorical Steel Liner Events: 2 Events: 0 (assume half aFlaw Likelihood failure)Failure Data: Containment 2/(70

• 5.5) = 5.2E-3 0.5/(70

• 5.5) = 1.3E-3location specific(consistent with CalvertCliffs analysis).2 Age Adjusted Steel Year Failure Rate Year Failure RateLiner Flaw Likelihood 1 2.1E-3 1 5.OE-4During 15-year interval, avg 5-10 5.2E-3 avg 5-10 1.3E-3assume failure rate 15 1.E-2 15 3.5E-3doubles every five years(14.9% increase per year). 15 year average = 15 year average -The average for 5th to 10th 6.27E-3 1.57E-3year is set to the historicalfailure rate (consistentwith Calvert Cliffsanalysis).3 Flaw Likelihood at 3, 0.71% (1 to 3 years) 0.18% (1 to 3 years)10, and 15 years 4.06% (1 to 10 years) 1.04% (1 to 10 years)Uses age adjusted liner 9.40% (1 to 15 years) 2.42% (1 to 15 years)flaw likelihood (Step 2), (Note that the Calvert Cliffs (Note that the Calvertassuming failure rate analysis presents the delta Cliffs analysis presents thedoubles every five years between 3 and 15 years of delta between 3 and 15(consistent with Calvert 8.7% to utilize in the years of 2.2% to utilize inCliffs analysis -See Table estimation of the delta- the estimation of the delta-6 of Reference [5]). LERF value. For this LERF value. For thisanalysis, the values are analysis, however, valuescalculated based on the 3, are calculated based on10, and 15 year intervals.) the 3, 10, and 15 yearintervals.)P0247130002-47224-17 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE 4.4-1STEEL LINER CORROSION BASE CASESTEP DESCRIPTION CONTAINMENT CONTAINMENTCYLINDER AND DOME BASEMAT4 Likelihood of Breach in 1% 0.10/0Containment GivenSteel Liner FlawThe failure probability ofthe containment cylinderand dome is assumed tobe 1% (compared to 1.1%in the Calvert Cliffsanalysis). The basematfailure probability isassumed to be a factor often less, 0.1% (comparedto 0.11% in the CalvertCliffs analysis).5 Visual Inspection 100/% 100%Detection Failure 5% failure to identify visual Cannot be visuallyLikelihood flaws plus 5% likelihood inspected.Utilize assumptions that the flaw is not visibleconsistent with Calvert (not through-cylinder butCliffs analysis. could be detected by ILRT)All events have beendetected through visualinspection. 5% visiblefailure detection is aconservative assumption.6 Likelihood of Non- 0.000710/o (at 3 years) 0.000180/a (at 3 years)Detected Containment =0.71%

• 1%

• 10% =0.18%

• 0.1%

• 100%Leakage(Steps 3

• 4

• 5) 0.00406%/o (at 10 0.001040/a (at 10years) years)=4.06%

• 1%/a

• 10% =1.04%/a

• 0.1%/a

• 100%0.0094% (at 15 years) 0.00242% (at 15=9.40%

• 1%

• 10% years)=2.42%

• 0.1%

• 100%P0247130002-47224-18 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyThe total likelihood of the corrosion-induced, non-detected containment leakage that issubsequently added to the EPRI Class 3b contribution is the sum of Step 6 for the containmentcylinder and dome, and the containment basemat:At 3 years : 0.00071% + 0.00018% = 0.00089%At 10 years: 0.00406% + 0.00104% = 0.00510%At 15 years: 0.0094% + 0.00242% = 0.01182%P0247130002-47224-19 Risk Impact Assessment of Extending the Indian Point ILRT Intervals5.0 RESULTSThe application of the approach based on EPRI Guidance [3] has led to the following results.The results are displayed according to the eight accident classes defined in the EPRI report.Table 5.0-1 lists these accident classes.TABLE 5.0-1ACCIDENT CLASSESACCIDENTCLASSES(CONTAINMENTRELEASE TYPE) DESCRIPTION1 Containment Intact2 Large Isolation Failures (Failure to Close)3a Small Isolation Failures (liner breach)3b Large Isolation Failures (liner breach)4 Small Isolation Failures (Failure to seal -Type B)5 Small Isolation Failures (Failure to seal-Type C)6 Other Isolation Failures (e.g., dependent failures)7 Failures Induced by Phenomena (Early and Late)8 Bypass (SGTR and Interfacing System LOCA)CDF All CET End states (including very low and no release)The analysis performed examined IP2 and IP3 specific accident sequences in which thecontainment remains intact or the containment is impaired. Specifically, the categorization ofthe severe accidents contributing to risk was considered in the following manner:" Core damage sequences in which the containment remains intact initially and in thelong term (EPRI Class 1 sequences).* Core damage sequences in which containment integrity is impaired due to randomisolation failures of plant components other than those associated with Type B orType C test components. For example, liner breach or bellows leakage, if applicable.(EPRI Class 3 sequences)." Core damage sequences in which containment integrity is impaired due tocontainment isolation failures of pathways left "opened" following a plant post-maintenance test. (For example, a valve failing to close following a valve stroketest. (EPRI Class 6 sequences). Consistent with the EPRI Guidance, this class is notspecifically examined since it will not significantly influence the results of thisanalysis.P0247130002-47225-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAccident sequences involving containment bypass (EPRI Class 8 sequences), largecontainment isolation failures (EPRI Class 2 sequences), and small containmentisolation "failure-to-seal" events (EPRI Class 4 and 5 sequences) are accounted for inthis evaluation as part of the baseline risk profile. However, they are not affected bythe ILRT frequency change.Class 4 and 5 sequences are impacted by changes in Type B and C test intervals;therefore, changes in the Type A test interval do not impact these sequences.The steps taken to perform this risk assessment evaluation are as follows:Step 1 Quantify the base-line risk in terms of frequency per reactor year for each of theaccident classes presented in Table 5.0-1.Step 2 Develop plant-specific person-rem dose (population dose) per reactor year foreach of the accident classes.Step 3 Evaluate risk impact of extending Type A test interval from 3 to 15 and 10 to 15years.Step 4 Determine the change in risk in terms of Large Early Release Frequency (LERF)in accordance with RG 1.174.Step 5 Determine the impact on the Conditional Containment Failure Probability(CCFP).5.1 STEP 1 -QUANTIFY THE BASE-LINE RISK IN TERMS OF FREQUENCY PER REACTORYEARThis step involves the review of the IP2 and IP3 Level 2 release category frequency results [18,19]. As described in Section 4.2, the release categories were assigned to the EPRI classes asshown in Table 4.2-6a for IP2 and in Table 4.2-6b for IP3. This application combined with theIP2 and IP3 dose risk (person-rem/yr) also shown in Tables 4.2-6a and 4.2-6b, respectivelyforms the basis for estimating the increase in population dose risk.For the assessment of the impact on the risk profile due to the ILRT extension, the potentialfor pre-existing leaks is included in the model. These pre-existing leak events are representedby the Class 3 sequences in EPRI 1018243 [3]. Two failure modes were considered for theClass 3 sequences, namely Class 3a (small breach) and Class 3b (large breach).The determination of the frequencies associated with each of the EPRI categories listed inTable 5.0-1 is presented next.P0247130002-47225-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsClass 1 SequencesThis group represents the frequency when the containment remains intact (modeled asTechnical Specification Leakage). The frequency per year for these sequences is 7.74E-06/yrfor IP2 and 1.11E-05/yr for IP3 (refer to Table 5.1-1 for Containment Release Type 1) and isdetermined by subtracting all containment failure end states including the EPRI/NEI Class 3aand 3b frequency calculated below, from the total CDF. For this analysis, the associatedmaximum containment leakage for this group is iLa, consistent with an intact containmentevaluation. Note that the values for this Class reported in Table 5.1-1 are slightly lower thanthat reported in Tables 4.2-6a and 4.2-6b since the 3a and 3b frequencies are now subtractedfrom Class 1.Class 2 SequencesThis group consists of large containment isolation failures. For IP2, this frequency is1.11E-08/yr (refer to Table 5.1-1, Containment Release Type 2). For IP3, this frequency is3.99E-09/yr (refer to Table 5.1-1, Containment Release Type 2).Class 3 SequencesThis group represents pre-existing leakage in the containment structure (e.g., containmentliner). The containment leakage for these sequences can be either small (2La to 10OLa) orlarge (>1OOLa). In this analysis, a value of 1OLa was used for small pre-existing flaws and10OLa for relatively large flaws.The respective frequencies per year are determined as follows:PROBciass_3a = probability of small pre-existing containment liner leakage= 0.0092 (see Section 4.3)PROBciass_3b = probability of large pre-existing containment liner leakage= 0.0023 (see Section 4.3)As described in Section 4.3, additional consideration is made to not apply these failureprobabilities to those cases that are already considered LERF scenarios (i.e., the Class 2, Class7, and Class 8 LERF contributions). This adjustment is made for based on the frequencyinformation from Tables 4.2-6a and 4.2-6b for IP2 and IP3, respectively as shown below.P0247130002-47225-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsFor IP2:Class_3a = 0.0092 * [CDF -(Class 2 + Class 7 LERF + Class 8 SGTR + Class 8 ISLOCA)]= 0.0092 * [1.17E-05 -(1.11E-08 + 6.90E-08 + 1.05E-06 + 2.77E-08)]= 9.73E-08/yrClass_3b = 0.0023 * [CDF -(Class 2 + Class 7 LERF + Class 8 SGTR + Class 8 ISLOCA)]= 0.0023 * [1.17E-05 -(1.11E-08 + 6.90E-08 + 1.05E-06 + 2.77E-08)]= 2.43E-08/yrFor IP3:Class_3a = 0.0092 * [CDF -(Class 2 + Class 7 LERF + Class 8 SGTR + Class 8 ISLOCA)]= 0.0092 * [1.48E-05 -(3.99E-09 + 7.14E-08 + 9.77E-07 + 1.93E-07)]= 1.25E-07/yrClass_3b = 0.0023 * [CDF -(Class 2 + Class 7 LERF + Class 8 SGTR + Class 8 ISLOCA)]= 0.0023 * [1.48E-05 -(3.99E-09 + 7.14E-08 + 9.77E-07 + 1.93E-07)]= 3.13E-08/yrFor this analysis, the associated containment leakage for Class 3a is 1OLa and 10OLa for Class3b, which is consistent with the latest EPRI methodology [3] and the NRC SE [7].Class 4 SequencesThis group represents containment isolation failure-to-seal of Type B test components.Because these failures are detected by Type B tests which are unaffected by the Type A ILRT,this group is not evaluated any further in this analysis.Class 5 SequencesThis group represents containment isolation failure-to-seal of Type C test components.Because these failures are detected by Type C tests which are unaffected by the Type A ILRT,this group is not evaluated any further in this analysis.Class 6 SequencesThis group is similar to Class 2. These are sequences that involve core damage with a failure-to-seal containment leakage due to failure to isolate the containment. These sequences aredominated by misalignment of containment isolation valves following a test/maintenanceevolution. Consistent with the EPRI guidance, this accident class is not explicitly consideredsince it has a negligible impact on the results.P0247130002-47225-4 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsClass 7 SequencesThis group represents containment failure induced by early and late severe accidentphenomena. From Table 4.2-6a for IP2, the frequency for early Class 7 sequences is4.66E-09/yr + 6.90E-08/yr = 7.37E-08/yr, and the frequency for the late Class 7 sequences is2.71E-06/yr. From Table 4.2-6b for IP3, the frequency for early Class 7 sequences is1.17E-07/yr + 7.14E-08/yr = 1.88E-07/yr, and the frequency for the late Class 7 sequences is2.17E-06/yr.Class 8 SeauencesThis group represents sequences where containment bypass occurs (SGTR or ISLOCA). Fromthe frequency information provided in Table 4.2-6a for IP2, the total SGTR contribution to coredamage is 1.05E-06/yr and the ISLOCA contribution to core damage is 2.77E-08/yr. From thefrequency information provided in Table 4.2-6b for IP3, the total SGTR contribution to coredamage is 9.77E-07/yr and the ISLOCA contribution to core damage is 1.93E-07/yr.Summary of Accident Class FrequenciesIn summary, the accident sequence frequencies that can lead to release of radionuclides to thepublic have been derived in a manner consistent with the definition of accident classes definedin EPRI 1018243 [3] and are shown in Table 5.1-1 for IP2 and for IP3.P0247130002-47225-5 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.1-1RADIONUCLIDE RELEASE FREQUENCIES AS A FUNCTION OFACCIDENT CLASS (IP2 AND IP3 BASE CASE)ACCIDENT DESCRIPTION IP2 IP3CLASS FREQUENCY FREQUENCY(CONTAINMENT (1/YR) (1/YR)RELEASE TYPE)1 Containment Intact 7.74E-06 1.11E-052 Large Isolation Failures (Failure to Close) 1.11E-08 3.99E-093a Small Isolation Failures (liner breach) 9.73E-08 1.25E-073b Large Isolation Failures (liner breach) 2.43E-08 3.13E-084 Small Isolation Failures (Failure to seal -N/A N/AType B)5 Small Isolation Failures (Failure to seal- N/A N/AType C)6 Other Isolation Failures (e.g., dependent N/A N/Afailures)7-CFE Failures Induced by Phenomena (Early) 7.37E-08 1.88E-077-CFL Failures Induced by Phenomena (Late) 2.71E-06 2.17E-068-SGTR Containment Bypass (Steam Generator 1.05E-06 9.77E-07Tube Rupture)8-ISLOCA Containment Bypass (Interfacing System 2.77E-08 1.93E-07LOCA)CDF All CET End States (Including Intact 1.17E-05 1.48E-05Case)5.2 STEP 2 -REACTOR YEARDEVELOP PLANT-SPECIFIC PERSON-REM DOSE (POPULATION DOSE) PERPlant-specific release analyses were performed to estimate the weighted average person-remdoses to the population within a 50-mile radius from the plant. The releases are based on acombination of the information provided by the IP2 and IP3 SAMA re-analysis [10], additionalpopulation dose runs for the intact containment scenarios [11], and the Level 2 containmentfailure release frequencies [18, 19] (see Tables 4.2-6a and 4.2-6b of this analysis). Theresults of applying these releases to the EPRI containment failure classifications areP0247130002-47225-6 Risk Impact Assessment of Extending the Indian Point ILRT Intervalssummarized below. Note that the 7-CFE release category is further refined to be the weightedaverage of the two contributors for moving forward in the ILRT methodology since it is notimpacted by the change to the ILRT interval.For IP2:Class 1Class 2Class 3aClass 3bClass 4Class 5Class 6Class 7-CFEClass 7-CFLClass 8-SGTR= 4.41E+04 person-rem (at 1.OLa)= 6.51E+07 person-rem= 4.41E+04 person-rem x 1OLa = 4.41E+05 person-rem= 4.41E+04 person-rem x 10OLa = 4.41E+06 person-rem= Not analyzed= Not analyzed= Not analyzed= (4.66E-09

• 1.94E+07 + 6.90E-08

• 6.51E+07) /(4.66E-09 + 6.90E-08) = 6.22E+07 person-rem= 6.87E+06 person-rem= 6.51E+07 person-remClass 8-ISLOCA = 6.51E+07 person-remFor IP3:Class 1Class 2Class 3aClass 3bClass 4Class 5Class 6Class 7-CFEClass 7-CFLClass 8-SGTR= 4.41E+04 person-rem (at 1.OLa)= 5.08E+07 person-rem= 4.41E+04 person-rem x 1OLa = 4.41E+05 person-rem= 4.41E+04 person-rem x 10OLa = 4.41E+06 person-rem= Not analyzed= Not analyzed= Not analyzed= (1.17E-07

• 2.OOE+07 + 7.14E-08

• 5.08E+07) /(1.17E-07 + 7.14E-08) = 3.17E+07 person-rem= 6.85E+06 person-rem= 5.08E+07 person-remClass 8-ISLOCA = 5.08E+07 person-remP0247130002-47225-7 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsIn summary, the population dose estimates derived for use in the risk evaluation per the EPRImethodology [3] for all EPRI classes are provided in Table 5.2-1, which includes the valuespreviously presented in Table 4.2-6a and 4.2-6b as well as the Class 3a, 3b, and 7-CFEpopulation doses calculated above.TABLE 5.2-1IP2 AND IP3 POPULATION DOSEFOR POPULATION WITHIN 50 MILESACCIDENT DESCRIPTION IP2 IP3CLASS PERSON- PERSON-(CONTAINMENT REM REMRELEASE TYPE) (0-50 (0-50MILES) MILES)1 Containment Intact 4.41E+04 4.41E+042 Large Isolation Failures (Failure to 6.51E+07 5.08E+07Close)3a Small Isolation Failures (liner breach) 4.41E+05 4.41E+053b Large Isolation Failures (liner breach) 4.41E+06 4.41E+064 Small Isolation Failures (Failure to seal -N/A N/AType B)5 Small Isolation Failures (Failure to seal -N/A N/AType C)6 Other Isolation Failures (e.g., dependent N/A N/Afailures)7-CFE Failures Induced by Phenomena (Early) 6.22E+07 3.17E+077-CFL Failures Induced by Phenomena (Late) 6.87E+06 6.85E+068-SGTR Containment Bypass (Steam Generator 6.51E+07 5.08E+07Tube Rupture)8-ISLOCA Containment Bypass (Interfacing 6.51E+07 5.08E+07System LOCA)The above population doses, when multiplied by the frequency results presented in Table5.1-1, yield the IP2 and IP3 baseline mean dose risk for each EPRI accident class. Theseresults are presented in Table 5.2-2a for IP2 and in Table 5.2-2b for IP3. Note that theadditional contribution to EPRI Class 3b from the corrosion analysis as described in Section 4.4is also included in these tables.P0247130002-47225-8 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.2-2AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON-REM EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES (0-50 PLUS CORROSION CORROSION(CONTAINMENT MILES) (PERSON-RELEASE TYPE) FREQUENCY PERSON- FREQUENCY PERSON- REM/YR) (1)(1/YR) REM/YR (1/YR) REM/YR(0-50 MILES) (0-50MILES)1 Containment 4.41E+04 7.74E-06 3.41E-01 7.74E-06 3.41E-01 -4.14E-06Intact (2)2 Large Isolation 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 9.73E-08 4.29E-02 9.73E-08 4.29E-02 --Failures (linerbreach)3b Large Isolation 4.41E+06 2.43E-08 1.07E-01 2.44E-08 1.08E-01 4.14E-4Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-9 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.2-2AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON-REM EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES (0-50 PLUS CORROSION CORROSION(CONTAINMENT MILES) (PERSON-RELEASE TYPE) FREQUENCY PERSON- FREQUENCY PERSON- REM/YR) (1)(1/YR) REM/YR (1/YR) REM/YR(0-50 MILES) (0-50MILES)7-CFE Failures Induced 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00 --by Phenomena(Early)7-CFL Failures Induced 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01 --by Phenomena(Late)8-SGTR Containment 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01 --Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00 --Bypass(Interfacing_System LOCA) ICDF All CET end 1.17E-05 9.426E+01 1.17E-05 9.426E+01 4.10E-4states) Only release Classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as ILa release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-10 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.2-2BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (PERSON-RELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- RERSO(1(1/YR) REM/YR (1/YR) REM/YR REM/YR)(0-50 MILES) (0-50MILES)1 Containment 4.41E+04 1.11E-05 4.91E-01 1.11E-05 4.91E-01 -5.32E-6Intact (2)2 Large Isolation 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 1.25E-07 5.51E-02 1.25E-07 5.51E-02 --Failures (linerbreach)3b Large Isolation 4.41E+06 3.13E-08 1.38E-01 3.14E-08 1.38E-01 5.32E-4Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-11 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.2-2BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (PERSON-RELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- REM/YR)(1/YR) REM/YR (1I/YR) REM/YR(0-50 MILES) (0-50MILES)7-CFE Failures Induced 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00 --by Phenomena(Early)7-CFL Failures Induced 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01 --by Phenomena(Late)8-SGTR Containment 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01 --Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00 --Bypass(Interfacing_System LOCA) ICDF All CET end 1.48E-05 8.114E+01 1.48E-05 8.115E+01 5.27E-4states(1) Only release Classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-12 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe baseline IP2 and IP3 doses compare reasonably with other plants given the relativepopulation densities surrounding each location:PLANT ANNUAL DOSE REFERENCE(PERSON-REM/YR)Indian Point 2 94.3 [Table 5.2-2a]Indian Point 3 81.1 [Table 5.2-2b]Peach Bottom 2 8.6 [22]Farley Unit 1, 2 1.5, 2.4 [23]Crystal River 1.4 [24]5.3 STEP 3 -EVALUATE RISK IMPACT OF EXTENDING TYPE A TEST INTERVAL FROM 10-TO-15 YEARSThe next step is to evaluate the risk impact of extending the test interval from its current ten-year value to fifteen-years. To do this, an evaluation must first be made of the risk associatedwith the ten-year interval since the base case applies to a 3-year interval (i.e., a simplifiedrepresentation of a 3-in- 10 year interval).Risk Impact Due to 10-year Test IntervalAs previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, therelease magnitude is not impacted by the change in test interval (a small or large breachremains the same, even though the probability of not detecting the breach increases). Thus,only the frequency of Class 3a and 3b sequences is impacted. The risk contribution is changedbased on the EPRI guidance as described in Section 4.3 by a factor of 3.33 compared to thebase case values. The results of the calculation for a 10-year interval are presented in Table5.3-1a for IP2 and in Table 5.3-1b for IP3.Risk Imoact Due to 15-Year Test IntervalThe risk contribution for a 15-year interval is calculated in a manner similar to the 10-yearinterval. The difference is in the increase in probability of not detecting a leak in Classes 3aand 3b. For this case, the value used in the analysis is a factor of 5.0 compared to the 3-yearinterval value, as described in Section 4.3. The results for this calculation are presented inTable 5.3-2a for IP2 and in Table 5.3-2b for IP3.P0247130002-47225-13 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-1AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-RELEASE TYPE) MILES) (1/YR) REM/YR (1/YR) REM/YR REM/YR) (1)(0-50 (0-50MILES) MILES)1 Containment 4.41E+04 7.46E-06 3.29E-01 7.45E-06 3.29E-01 -2.38E-05Intact (2)2 Large Isolation 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 3.24E-07 1.43E-01 3.24E-07 1.43E-01 --Failures (linerbreach)3b Large Isolation 4.41E+06 8.1OE-08 3.57E-01 8.15E-08 3.60E-01 2.38E-3Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-14 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-1AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 FREQUENCY PERSON FREQUENCY PERSON (PERSON-RELEASE TYPE) MILES) (1/YR) REM/YR (1/YR) REM/YR REM/YR)(1)(0-S0 (0-50MILES) MILES)7-CFE Failures Induced 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00by Phenomena(Early)7-CFL Failures Induced 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01by Phenomena(Late)8-SGTR Containment 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00Bypass(InterfacingSystem LOCA) ICDF All CET end 1.17E-05 9.460E+01 1.17E-05 9.460E+01 2.35E-3states(1) Only release classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 1L. release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3aand 3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-15 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-1BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 FREQUENCY PERSON FREQUENCY PERSON (PERSON-RELEASE TYPE) MILES) (1/YR) REM/YR (1/YR) REM/YR REM/YR)(0-50 (0-50MILES) MILES)1 Containment 4.41E+04 1.08E-05 4.75E-01 1.08E-05 4.75E-01 -3.05E-5Intact (2)2 Large Isolation 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 4.16E-07 1.84E-01 4.16E-07 1.84E-01 --Failures (linerbreach)3b Large Isolation 4.41E+06 1.04E-07 4.59E-01 1.05E-07 4.62E-01 3.05E-3Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-16 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-1BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 FREQUENCY PERSON FREQUENCY PERSON (PERSON-RELEASE TYPE) MILES) (1I/YR) REM/YR (1/YR) REM/YR REM/YR)(0-50 (0-50MILES) MILES)7-CFE Failures Induced 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00by Phenomena(Early)7-CFL Failures Induced 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01by Phenomena(Late)8-SGTR Containment 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00Bypass(InterfacingSystem LOCA) ICDF All CET end 1.48E-05 8.158E+01 1.48E-05 8.158E+01 3.02E-3states(1) Only release classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 11 release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3aand 3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-17 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-2AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSIONS(CONTAINMENT (0-50 (ESNRELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-(1/YR) REM/YR (1/YR) REM/YR REM/YR)(0-50 (0-50MILES) MILES)1 Containment Intact 4.41E+04 7.25E-06 3.20E-01 7.25E-06 3.20E-01 -5.51E-05(2)2 Large Isolation 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 4.86E-07 2.15E-01 4.86E-07 2.15E-01 --Failures (linerbreach)3b Large Isolation 4.41E+06 1.22E-07 5.36E-01 1.23E-07 5.42E-01 5.51E-3Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependent failures)P0247130002-47225-18 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-2AIP2 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (ESNRELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-(1/YR) REM/YR (1/YR) REM/YR REM/YR)(1)(0-50 (0-50MILES) MILES)7-CFE Failures Induced 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00 --by Phenomena(Early)7-CFL Failures Induced 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01 --by Phenomena(Late)8-SGTR Containment 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01 --Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00 --Bypass(InterfacingSystem LOCA)CDF All CET end 1.17E-05 9.484E+01 1.17E-05 9.484E+01 5.46E-3states(1) Only release classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in asmall reduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3aand 3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-19 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-2BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (ESNRELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-(1/YR) REM/YR (1/YR) REM/YR REM/YR)(1)(0-50 (0-50MILES) MILES)1 Containment 4.41E+04 1.05E-05 4.64E-01 1.05E-05 4.64E-01 -7.08E-5Intact (2)2 Large Isolation 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 --Failures (Failure toClose)3a Small Isolation 4.41E+05 6.25E-07 2.76E-01 6.25E-07 2.76E-01 --Failures (linerbreach)3b Large Isolation 4.41E+06 1.56E-07 6.89E-01 1.58E-07 6.96E-01 7.08E-3Failures (linerbreach)4 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal -Type B)5 Small Isolation N/A N/A N/A N/A N/A N/AFailures (Failure toseal-Type C)6 Other Isolation N/A N/A N/A N/A N/A N/AFailures (e.g.,dependentfailures)P0247130002-47225-20 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.3-2BIP3 ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS;CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCYACCIDENT DESCRIPTION PERSON- EPRI METHODOLOGY EPRI METHODOLOGY CHANGE DUE TOCLASSES REM PLUS CORROSION CORROSION(CONTAINMENT (0-50 (ESNRELEASE TYPE) MILES) FREQUENCY PERSON- FREQUENCY PERSON- (PERSON-(1/YR) REM/YR (1/YR) REM/YR REM/YR)(1)(0-S0 (0-50MILES) MILES)7-CFE Failures Induced 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00 --by Phenomena(Early)7-CFL Failures Induced 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01 --by Phenomena(Late)8-SGTR Containment 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01 --Bypass (SteamGenerator TubeRupture)8-ISLOCA Containment 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00 --Bypass(InterfacingSystem LOCA)CDF All CET end 1.48E-05 8.189E+01 1.48E-05 8.190E+01 7.01E-3statesIII(1) Only release classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every fiveyears. The additional frequency added to Class 3b is subtracted from Class 1 and the population dose rates are recalculated. This results in a smallreduction to the Class 1 dose rate and an increase to the Class 3b dose rate.(2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and3b include failures of containment to meet the Technical Specification leak rate.P0247130002-47225-21 Risk Impact Assessment of Extending the Indian Point ILRT Intervals5.4 STEP 4 -DETERMINE THE CHANGE IN RISK IN TERMS OF LARGE EARLY RELEASEFREQUENCYRegulatory Guide 1.174 provides guidance for determining the risk impact of plant-specificchanges to the licensing basis. RG 1.174 defines very small changes in risk as resulting inincreases of core damage frequency (CDF) below 1E-06/yr and increases in LERF below1E-07/yr, and small changes in LERF as below 1E-06/yr. Because the ILRT does not impactCDF for IP2 and IP3, the relevant metric is LERF.For IP2 and IP3, 100% of the frequency of Class 3b sequences can be used as a conservativefirst-order estimate to approximate the potential increase in LERF from the ILRT intervalextension (consistent with the EPRI guidance methodology and the NRC SE). Based on theoriginal 3-in-10 year test interval assessment from Tables 5.2-2a and 5.2-2b, the Class 3bfrequency is 2.44E-08/yr for IP2 and 3.14E-08/yr for IP3, which includes the corrosion effect ofthe containment liner. Based on a ten-year test interval from Tables 5.3-1a and 5.3-1b, theClass 3b frequency is 8.15E-08/yr for IP2 and 1.05E-07/yr for IP3; and, based on a fifteen-year test interval from Tables 5.3-2a and 5.3-2b, it is 1.23E-07/yr for IP2 and 1.58E-07/yr forIP3. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is dueto increasing the ILRT test interval from 3 to 15 years (including corrosion effects) is 9.84E-08/yr for IP2 and 1.26E-07/yr for IP3. Similarly, the increase in LERF due to increasing theinterval from 10 to 15 years (including corrosion effects) is 4.13E-08/yr for IP2 and 5.31E-08/yr for IP3. As can be seen, even with the conservatisms included in the evaluation (per theEPRI methodology), the estimated change in LERF is well within Region II of Figure 4 ofReference [4] (i.e., the acceptance criteria for small changes in LERF) when comparing the 15year results to the original 3-in-10 year requirement.5.5 STEP 5 -DETERMINE THE IMPACT ON THE CONDITIONAL CONTAINMENT FAILUREPROBABILITYAnother parameter that can provide input into the decision-making process is the change inthe conditional containment failure probability (CCFP). The change in CCFP is indicative of theeffect of the ILRT on all radionuclide releases, not just LERF. The CCFP can be calculated fromthe results of this analysis. One of the difficult aspects of this calculation is providing adefinition of the "failed containment." In this assessment, the CCFP is defined such thatcontainment failure includes all radionuclide release end states other than the intact state and,consistent with the EPRI guidance, the small isolation failures (Class 3a). The conditional partof the definition is conditional given a severe accident (i.e., core damage).P0247130002-47225-22 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe change in CCFP can be calculated by using the method specified in the EPRI methodology[3]. The NRC SE has noted a change in CCFP of <1.5% as the acceptance criterion to be usedas the basis for showing that the proposed change is consistent with the defense-in-depthphilosophy. Table 5.5-1 shows the CCFP values that result from the assessment for thevarious testing intervals including corrosion effects in which the flaw rate is assumed to doubleevery five years.TABLE 5.5-1IP2 AND IP3 ILRT CONDITIONAL CONTAINMENT FAILURE PROBABILITIESUNIT CCFP CCFP CCFP3 IN 10 1 IN 10 1 IN 15 ACCFP15-3 ACCFP15-10YRS YRS YRSIndian Point 2 33.19% 33.67% 34.03% 0.84% 0.35%Indian Point 3 24.03% 24.52% 24.88% 0.85% 0.36%CCFP = [1 -(Class 1 frequency + Class 3a frequency)/CDF] x 100%The change in CCFP of less than 1% as a result of extending the test interval to 15 years fromthe original 3-in-10 year requirement is judged to be relatively insignificant, and is less thanthe NRC SE acceptance criteria of <. 1.5%.5.6 SUMMARY OF INTERNAL EVENTS RESULTSTable 5.6-1a summarizes the internal events results of this ILRT extension risk assessment forIP2. Table 5.6-1b summarizes the internal events results of this ILRT extension riskassessment for IP3. The results between the 3-in-10 year interval and the 15 year intervalcompared to the acceptance criteria are then shown in Table 5.6-2 for IP2 and IP3, and it isdemonstrated that the acceptance criteria are met.P0247130002-47225-23 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.6-1AIP2 ILRT CASES:BASE, 3 TO 10, AND 3 TO 15 YR EXTENSIONS(INCLUDING AGE ADJUSTED STEEL LINER CORROSION LIKELIHOOD)EPRI DOSE BASE CASE EXTEND TO EXTEND TOCLASS PER-REM 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARSCDF PERSON- CDF PERSON- CDF PERSON-(1/YR) REM/YR (1/YR) REM/YR (1/YR) REM/YR1 4.41E+04 7.74E-06 3.41E-01 7.45E-06 3.29E-01 7.25E-06 3.20E-012 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 1.11E-08 7.23E-013a 4.41E+05 9.73E-08 4.29E-02 3.24E-07 1.43E-01 4.86E-07 2.15E-013b 4.41E+06 2.44E-08 1.08E-01 8.15E-08 3.60E-01 1.23E-07 5.42E-017-CFE 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00 7.37E-08 4.58E+007-CFL 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01 2.71E-06 1.86E+018-SGTR 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01 1.05E-06 6.80E+018-ISLOCA 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00 2.77E-08 1.80E+00Total [_1.17E-05 9.426E+01 1. 17E-0-9 19.4 .484E+01ILRT Dose Rate 1.51E-01 5.02E-01 7.56E-01(person-rem/yr) from3a and 3bDelta From 3 yr --- 3.39E-01 5.84E-01TotalIDose From 10 yr 2.45E-01DoseRate*1)3b Frequency (LERF) 2.44E-08 8.15E-08 1.23E-07Delta 3b From 3 yr --- 5.71E-08 9.84E-08LERF From 10 yr ......_4.13E-08CCFP % 33.19% 33.67% 34.03%Delta From 3 yr --- 0.49% 0.84%CCFP %From 10 yr ...0.35%( The overall difference in total dose rate is less than the difference of only the 3a and 3bcategories between two testing intervals. This is due to the fact that the Class 1 person-rem/yr decreases when extending the ILRT frequency.P0247130002-47225-24 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.6-1BIP3 ILRT CASES:BASE, 3 TO 10, AND 3 TO 15 YR EXTENSIONS(INCLUDING AGE ADJUSTED STEEL LINER CORROSION LIKELIHOOD)EPRI DOSE BASE CASE EXTEND TO EXTEND TOCLASS PER-REM 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARSCDF PERSON- CDF PERSON- CDF PERSON-(1/YR) REM/YR (1/YR) REM/YR (1/YR) REM/YR1 4.41E+04 1.11E-05 4.91E-01 1.08E-05 4.75E-01 1.05E-05 4.64E-012 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 3.99E-09 2.03E-013a 4.41E+05 1.25E-07 5.51E-02 4.16E-07 1.84E-01 6.25E-07 2.76E-013b 4.41E+06 3.14E-08 1.38E-01 1.05E-07 4.62E-01 1.58E-07 6.96E-017-CFE 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00 1.88E-07 5.97E+007-CFL 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01 2.17E-06 1.49E+018-SGTR 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01 9.77E-07 4.96E+018-ISLOCA 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00 1.93E-07 9.80E+00Total 1.48E-05 8.115E+01 I 1.48E-05 18.158E+011 1.48E-05 18.190E+01ILRT Dose Rate 1.93E-01 6.46E-01 9.72E-01(person-rem/yr) from3a and 3bDelta From 3 yr --- 4.36E-01 7.51E-01TotalDose From 10 yr --- 3.15E-01DoseRate(l)3b Frequency (LERF) 3.14E-08 1.05E-07 1.58E-07Delta 3b From 3 yr --- 7.34E-08 1.26E-07LERFtFrom 10 yr ...... 5.31E-08CCFP % 24.03% 24.52% 24.88%Delta From 3 yr --- 0.49% 0.85%CCFP %From 10 yr --- 0.36%(1) The overall difference in total dose rate is less than the difference of only the 3a and 3bcategories between two testing intervals. This is due to the fact that the Class 1 person-rem/yr decreases when extending the ILRT frequency.P0247130002-47225-25 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.6-2IP2 AND IP3 ILRT EXTENSION COMPARISON TO ACCEPTANCE CRITERIAUnit ALERF APerson-rem/yr ACCFPIndian Point 2 9.84E-8/yr 0.584/yr (0.62%) 0.84%Indian Point 3 1.26E-7/yr 0.751/yr (0.93%) 0.85%Acceptance < 1.OE-6/yr <1.0 person- <1.50/oCriteria rem/yr or <1.0%5.7 EXTERNAL EVENTS CONTRIBUTIONSince the risk acceptance guidelines in RG 1.174 are intended for comparison with a full-scopeassessment of risk, including internal and external events, a bounding analysis of the potentialimpact from external events is presented here.The method chosen to account for external events contributions is similar to that used in theSAMA analysis [20] in which a multiplier was applied to the internal events results based oninformation from the IPEEE [8, 9]. Similar to that provided in the SAMA analysis, a descriptionof the external events contribution to risk at IP2 and IP3 is provided below.5.7.1 Indian Point 2 External Events DiscussionThe IP2 Individual Plant Examination of External Events (IPEEE) included quantitative CDFresults for high winds, seismic, and fire contributors. Each of these is discussed below.A high wind analysis was performed for the IP2 IPEEE. Conservative assumptions in the highwind PRA analysis included the following.* Offsite power was assumed to be lost for all high wind events." Building frame failures were assumed to cause failure of all equipment within thebuilding.* Missile (high wind projectile) impact on a structure was assumed to cause failure ofall equipment within that structure.* Likelihood of missile (high wind projectile) strikes was assumed to be independent ofthe intensity of the hazard.* Both onsite and offsite alternate power sources (gas turbines) were assumed to failgiven failure of a more robust structure.P0247130002-47225-26 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe core damage frequency contribution associated with high wind events was estimated to be3.03E-05/yr. As described above, this is a conservative value. In addition, plant changes,improved equipment performance data, and modeling improvements since the issuance of theIP2 IPEEE have demonstrated that the response of plant systems as modeled at that time wasconservative. This can be seen from the reduction in internal events CDF from 2.85E-05/yr atthe time the IPEEE was developed to the present value of 1.17E-05/yr. Although conservative,consistent with the SAMA analysis, the wind risk contribution of 3.03E-05/yr is maintained todetermine the potential external events impact in the ILRT extension assessment.A seismic PRA analysis was performed for the seismic portion of the IP2 IPEEE. The seismicPRA analysis was a conservative analysis. Therefore, its results should not be compareddirectly with the best-estimate internal events results. Conservative assumptions in the seismicPRA analysis included the following.* Sequences in the seismic PRA involving loss of off-site power were assumed to beunrecoverable. If off-site power was recovered following a seismic event, there wouldbe many more systems available to maintain core cooling and containment integritythan were credited for those sequences.* A single, conservative, surrogate element whose failure leads directly to coredamage was used in the seismic risk quantification to model the most seismicallyrugged components.* Seismic-induced ATWS was considered in the analysis, but no credit was included formanual scram or mitigation of ATWS using the boration system. This conservativelyresulted in most seismic-induced ATWS events leading to consequential coredamage.* Redundant components were conservatively assumed to be completely correlated bytreating them as if they were one component for the purpose of determining theprobability of seismic induced failures." Several systems were assumed to be unavailable during a seismic event, including:a. the city water system, which can be used to supply backup cooling to thecharging pumps if CCW is lost, as an alternate source of suction to the AFWpumps and to provide alternate cooling to the RHR and SI pumps;b. the primary water system, which can also be used as a backup to CCW tosupply cooling to the RHR and SI pumps; andc. the onsite and offsite gas turbine generators, which can provide alternatestation power.* No credit was taken for recovery of power through the alternate safe shutdownsystem (ASSS).The seismic CDF in the IPEEE was originally estimated to be 1.46E-05/yr. As a result of anIPEEE recommendation, the CCW surge tank hold-down bolts were upgraded, reducing theseismic CDF to 1.06E-05/yr. Although it remains conservative, consistent with the SAMAP0247130002-47225-27 Risk Impact Assessment of Extending the Indian Point ILRT Intervalsanalysis, the seismic risk contribution of 1.06E-05/yr is maintained to determine the potentialexternal events impact in the ILRT extension assessment.The conservative EPRI FIVE methodology was used for initial screening of fire zones in the IP2IPEEE fire analysis. Unscreened fire zones were then analyzed in more detail using a fire PRAapproach. The sum of the resulting fire zone CDF values is approximately 1.84E-05/yr.Conservative assumptions in the IP2 IPEEE fire analysis include the following." The frequency and severity of fires were generally conservatively overestimated inthe generic IPEEE fire analysis methods. A revised NRC fire events databaseindicates a trend toward lower frequency and less severe fires. This trend reflectsimproved housekeeping, reduction in transient fire hazards, and other improved fireprotection steps at utilities.* Cable failure due to fire damage was assumed to arise from open circuits, hot shortcircuits, and short circuits to ground. In damaging a cable, the analysis addressedthe ability of the fire to induce the conductor failure mode of concern. Hot shortswere conservatively assigned a probability of 0.1, which was applied to all singlephase, AC control circuit or DC power and control circuit cases regardless of whetherthe wires were in the same multi-conductor." A plant trip was assumed for all fires, including those for which immediate operatoractions are not specified in emergency response procedures." PORV block valves were assumed to be in the more limiting position (open or closed)to maximize the impact of the fire.* The main feedwater and condensate systems were assumed to be unavailable in allscenarios, even when their power source was not impacted by the fire scenario. Useof these systems for recovery, following a failure of AFW, is addressed in currentplant procedures.* All sequences involving induced RCP seal LOCAs were assumed to lead to completeseal failure. Although casualty cables exist for powering ECCS pumps from the ASSSpower source, the ASSS was assumed to be ineffective in mitigating induced LOCAs.* The currently accepted RCP seal LOCA methodology is more detailed and providessequences with varying leakage rates. Under that current methodology, a majority ofseal LOCAs remain within the capability of a charging pump (which has hardwiredASSS transfer capability) to provide makeup.As noted previously, plant changes, improved equipment performance data and modelingimprovements since the issuance of the IP2 IPEEE have demonstrated that the response ofplant systems as modeled at that time was conservative. This can be seen from the reductionin internal events CDF from 2.85E-05/yr at the time the IPEEE was developed to the presentvalue of 1.17E-5/yr., a reduction factor of 2.4. Factoring in the additional conservatisms in thefire analysis noted above, an overall reduction factor of 2 is reasonable which is consistent withthe assumption used in the SAMA analysis [20]. The IPEEE fire CDF value, reduced by a factorof two, is 9.20E-06/yr.P0247130002-47225-28 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe IP2 Individual Plant Examination of External Events (IPEEE) concluded for "Other" externalevents, with the exception of "high wind" events as noted above, that no undue risks arepresent that might contribute to CDF with a predicted frequency in excess of 1.OE-06/yr. Asthese events are not dominant contributors to external event risk and quantitative analysis ofthese events is not practical, they are considered negligible in estimation of the external eventsimpact on the ILRT extension assessment.In summary, the combination of the IPEEE high wind CDF and the reduced seismic and fireCDF values described above results in an external events risk estimate of 5.01E-05/yr, which is4.3 times higher than the internal events CDF (1.17E-05/yr).5.7.2 Indian Point 3 External Events DiscussionThe IP3 Individual Plant Examination of External Events (IPEEE) concluded for high winds,floods, and "Other" external events that no undue risks are present that might contribute toCDF with a predicted frequency in excess of 1.OE-06/yr. Note that at IP3 (compared to IP2),the EDGs are in separate concrete bunkered cells and as such are not susceptible to highwinds. In any event, as these other events are not dominant contributors to external eventrisk and quantitative analysis of these events is not practical, they are considered negligible inestimation of the external events impact on the ILRT extension assessment. The IPEEEanalyses using the seismic PRA and fire PRA provided quantitative, but conservative, results.Therefore, the results were combined as described below to represent the total external eventsrisk.A seismic PRA analysis was performed for the seismic portion of the IP3 IPEEE. The seismicPRA analysis is a conservative analysis. Therefore, its results should not be compared directlywith the best-estimate internal events results. Conservative assumptions in the seismic PRAanalysis included the following." Each of the sequences in the seismic PRA assumes unrecoverable loss of off-sitepower. If off-site power was maintained, or recovered, following a seismic event,there would be many more systems available to maintain core cooling andcontainment integrity than were credited in the analysis.* Seismic events were assumed to induce a small loss of coolant accident (LOCA) inaddition to a loss of offsite power." A single, conservative, surrogate element whose failure leads directly to coredamage was used in the seismic risk quantification to model the most seismicallyrugged components." Redundant components were conservatively assumed to be completely correlated bytreating them as if they were one component for the purpose of determining theprobability of seismic induced failures.P0247130002-47225-29 Risk Impact Assessment of Extending the Indian Point ILRT Intervals* The ATWS event tree was conservatively simplified so that all conditions which leadto a failure to trip result in core damage, without the benefit of emergency borationor other mitigating systems." Because there is little industry experience with crew actions following seismic events,human actions were conservatively characterized.The seismic CDF in the IPEEE was conservatively estimated to be 4.40E-05/yr. As describedabove, this is a conservative value. The seismic PRA CDF has been re-evaluated to reflectupdated random component failure probabilities and to model recovery of onsite power andlocal operation of the turbine-driven AFW pump. The updated seismic CDF is 2.65E-05/yr.Although it remains conservative, consistent with the SAMA analysis, the seismic riskcontribution of 2.65E-05/yr is maintained to determine the external events impact on the ILRTextension assessment.The EPRI Fire PRA Implementation Guide was followed for the IP3 IPEEE fire analysis. The EPRIFire Induced Vulnerability Evaluation (FIVE) method was used for the initial screening, fortreatment of transient combustibles, and as the source of fire frequency data. The sum of theresulting fire zone CDF values is approximately 5.58E-05/yr. Conservatisms in the IP3 IPEEEfire analysis include the following.* The frequency and severity of fires were generally conservatively overestimated. Arevised NRC fire events database indicates a trend toward lower frequency and lesssevere fires. This trend reflects improved housekeeping, reduction in transient firehazards, and other improved fire protection steps at utilities." There is little industry experience with crew actions following fires. This led toconservative characterization of crew actions in the IPEEE fire analysis. Because CDFis strongly correlated with crew actions, this conservatism has a profound effect onfire results.* Hot gas layer temperature timing calculations were based on simplified analyses(versus more detailed calculations such as GOTHIC or even COMPBURN) which arebelieved to result in more severe timing (i.e., shorter time to equipment failure).* Heat and combustion products from a fire within a zone were assumed to beconfined within the zone. Heat loss through separating zones was not considered;nor was heat loss through open equipment hatches, ladder ways, open doorways, orunsealed penetrations." Cable failure due to fire damage was assumed to arise from open circuits, hot shortscircuits, and short circuits to ground. In damaging a cable, the fire was alwaysassumed to induce the conductor failure mode of concern." A plant trip was assumed for all fires, including those for which immediate operatoractions are not specified in emergency response procedures." For several fire zones, a minimum heat requirement for target damage wasestimated." Propagation of fires in cable spreading room trays and electrical tunnels was modeledusing a maximum heat release rate. This results in a shorter time to damage thanP0247130002-47225-30 Risk Impact Assessment of Extending the Indian Point ILRT Intervalsthe five-minute delay using heat release rate scaling factors as a function of distancerecommended in the EPRI fire PRA implementation guide.Implementation of the IP3 IPEEE recommendations reduced the fire risk. The fire suppressionsystem in the 480V switchgear room was restored to automatic actuation, and realignmentand rerouting of the power feeds to the EDG exhaust fans and engine auxiliaries in emergencydiesel generator room 31, emergency diesel generator room 32, and emergency dieselgenerator room 33 significantly reduce the respective fire zone's CDF. In addition, restorationof the 480V switchgear room fire suppression system to automatic actuation results in a similarreduction in the fire zone 14/37A multiple compartment fire CDF. Consequently, the IPEEE fireCDF value was reduced from 5.58E-05/yr to 2.55E-05/yr. Although it remains conservative,consistent with the SAMA analysis, the fire risk contribution of 2.55E-05/yr is maintained todetermine the potential external event impact on the ILRT extension assessment.In summary, combining the reduced seismic and fire CDF values results in an external eventsrisk estimate of 5.20E-05/yr, which is 3.5 times higher than the internal events CDF (1.48E-05/yr).5.7.3 Additional Seismic Risk DiscussionAs an additional consideration, it can be noted that in June 2013, Entergy submittedinformation to the NRC that addressed some conservatisms in the original IPEEE analyses, andindicated that the seismic CDF risk at IP2 and IP3 are both actually less than 1.OE-05/yr [25].However, to maintain consistency with the approach utilized in the SAMA analysis, theadditional information will not be factored into this analysis but is noted here for completeness.5.7.4 External Events Impact SummaryTable 5.7-1 summarizes the external events CDF contribution for IP2 and 1P3. Although notedas conservative, these values are consistent with that used in the SAMA analysis [20].P0247130002-47225-31 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-1EXTERNAL EVENTS CONTRIBUTOR SUMMARY [20]EXTERNAL EVENT INITIATOR GROUP IP2 CDF (1/YR)_ 1 1P3 CDF (1/YR)Seismic 1.06E-05 2.65E-05Internal Fire 9.20E-06 2.55E-05High Winds 3.03E-05 ScreenedOther Hazards Screened ScreenedTotal (for initiators with CDF available) 5.01E-05 5.20E-05Internal Events CDF 1.17E-05 1.48E-05External Events Multiplier 4.28 3.51From Table 5.7-1, the external events multiplier for IP2 is conservatively estimated to be 4.28and for IP3, it is conservatively estimated to be 3.51.5.7.5 External Events Impact on ILRT Extension AssessmentThe EPRI Category 3b frequency for the 3-per-10 year, 1-per-10 year, and 1-per-15 year ILRTintervals are shown in Table 5.6-1a for IP2 as 2.44E-08/yr, 8.15E-08/yr, and 1.23E-07/yr,respectively. Using an external events multiplier of 4.28 for IP2, the change in the LERF riskmeasure due to extending the ILRT from 3-per-l.0 years to 1-per-15 years, including bothinternal and external hazards risk, is estimated as shown in Table 5.7-2a. Similarly, the EPRIClass 3b frequencies shown in Table 5.6-1b for IP3 are 3.14E-08/yr, 1.05E-07/yr, and1.58E-07/yr. Using an external events multiplier of 3.51 for IP3, the change in the LERF riskmeasure due to extending the ILRT from 3-per-10 years to 1-per-15 years, including bothinternal and external hazards risk, is estimated as shown in Table 5.7-2b.P0247130002-47225-32 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-2AIP2 3B (LERF/YR) AS A FUNCTION OF ILRT FREQUENCYFOR INTERNAL AND EXTERNAL EVENTS(INCLUDING AGE ADJUSTED STEEL LINER CORROSION LIKELIHOOD)3B B 3B LERFFREQUENCY FREQUENCY FREQUENCY INCREASE"1)(3-PER-10 (1-PER-10 (1-PER-15YR ILRT) YEAR ILRT) YEAR ILRT)Internal Events 2.44E-08 8.15E-08 1.23E-07 9.84E-08ContributionExternal EventsContribution (Internal 1.05E-07 3.49E-07 5.26E-07 4.22E-07Events CDF x 4.28)Combined (Internal +1.29E-07 4.31E-07 6.49E-7 5.20E-07External)(1) Associated with the change from the baseline 3-per-10 year frequency to the proposed 1-per-15year frequency.Thus for IP2, the total increase in LERF (measured from the baseline 3-per-10 year ILRTinterval to the proposed 1-per-15 year frequency) due to the combined internal and externalevents contribution is estimated as 5.20E-07/yr, which includes the age adjusted steel linercorrosion likelihood.TABLE 5.7-2B1P3 3B (LERF/YR) AS A FUNCTION OF ILRT FREQUENCYFOR INTERNAL AND EXTERNAL EVENTS(INCLUDING AGE ADJUSTED STEEL LINER CORROSION LIKELIHOOD)3B 3B 3B LERFFREQUENCY FREQUENCY FREQUENCY INCREASE"1)(3-PER-10 (1-PER-10 (1-PER-15YR ILRT) YEAR ILRT) YEAR ILRT)Internal Events 3.14E-08 1.05E-07 1.58E-07 1.26E-07ContributionExternal EventsContribution (Internal 1.10E-07 3.67E-07 5.53E-07 4.43E-07Events CDF x 3.51)ombined (Internal + 1.41E-07 4.72E-07 7.11E-7 5.70E-07External) _Associated with the change from the baselineyear frequency.3-per-10 year frequency to the proposed 1-per-15P0247130002-47225-33 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThus for IP3, the total increase in LERF (measured from the baseline 3-per-10 year ILRTinterval to the proposed 1-per-15 year frequency) due to the combined internal and externalevents contribution is estimated as 5.70E-07/yr, which includes the age adjusted steel linercorrosion likelihood.The other acceptance criteria for the ILRT extension risk assessment can be similarly derivedusing the multiplier approach. The results between the 3-in-10 year interval and the 15 yearinterval compared to the acceptance criteria are shown in Table 5.7-3. As can be seen, theimpact from including the external events contributors would not change the conclusion of therisk assessment. That is, the acceptance criteria are all met such that the estimated riskincrease associated with permanently extending the ILRT surveillance interval to 15 years hasbeen demonstrated to be small. Note that a bounding analysis for the total LERF contributionfollows Table 5.7-3 to demonstrate that the total LERF value for IP2 and IP3 is less than1.OE-5/yr consistent with the requirements for a "Small Change" in risk of the RG 1.174acceptance guidelines.P0247130002-47225-34 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-3COMPARISON TO ACCEPTANCE CRITERIA INCLUDING EXTERNALEVENTS CONTRIBUTION FOR IP2 AND IP3Contributor ALERF APerson-rem/yr ACCFPIP2 Internal 9.84E-8/yr 0.584/yr (0.62%) 0.84%EventsIP2 External 4.22E-7/yr 2.50/yr (0.62%) 0.84%EventsIndian Point 2 5.20E-7/yr 3.09/yr (0.62%) 0.84%TotalIP3 Internal 1.26E-7/yr 0.751/yr (0.93%) 0.85%EventsIP3 External 4.43E-7/yr 2.63/yr (0.93%) 0.85%EventsIndian Point 3 5.70E-7/yr 3.38/yr (0.93%/) 0.850/0TotalAcceptance < 1.OE-6/yr <1.0 person- <1.50/0Criteria rem/yr or <1.0%The 5.20E-07/yr increase in LERF for IP2 and the 5.70E-07/yr increase in LERF for IP3 due tothe combined internal and external events from extending the ILRT frequency from 3-per-10years to 1-per-15 years falls within Region II between 1.OE-7 to 1.OE-6 per reactor year("Small Change" in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when thecalculated increase in LERF due to the proposed plant change is in the "Small Change" range,the risk assessment must also reasonably show that the total LERF is less than 1.OE-5/yr.Similar bounding assumptions regarding the external event contributions that were madeabove are used for the total LERF estimate.From Table 4.2-1, the total LERF due to postulated internal event accidents is 1.16E-06/yr forIP2 and 1.25E-06/yr for IP3. Although some of the LERF contributors may not be applicable toexternal events initiators, the base LERF distribution due to external events is assumed to bethe same as the internal events contribution. The total LERF values for IP2 and IP3 are thenshown in Table 5.7-4.P0247130002-47225-35 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-4IMPACT OF 15-YR ILRT EXTENSION ON LERF FOR IP2 AND IP3LERF CONTRIBUTOR IP2 (1/YR) IP3 (1/YR)Internal Events LERF 1.16E-06 1.25E-064.97E-06 4.38E-06External Events LERF [Internal Events LERF * [Internal Events LERF *4.28] 3.51]Internal Events LERF due to 1.23E-07 1.58E-07ILRT (at 15 years) (1)External Events LERF due to 5.26E-07 5.53E-07ILRT (at 15 years) (1)Total 6.78E-06/yr 6.34E-06/yr) Including age adjusted steel liner corrosion likelihood as reported in Table 5.7-2a for IP2and Table 5.7-2b for IP3.As can be seen, the estimated upper bound LERF for IP2 is estimated as6.78E-06/yr and for IP3 it is 6.34E-06/yr. These values are both less than the RG 1.174requirement to demonstrate that the total LERF due to internal and external events is less than1.OE-5/yr.5.7.6 Alternative Approach for External Events Impact on ILRT Extension AssessmentThe approach above described in Section 5.7.5 for the external events impact is consistentwith that used in the Palisades ILRT extension risk assessment evaluation that was submittedby Entergy [26] and approved by the NRC [27]. As shown, the IP2 and IP3 results fall withinthe value in the NRC SER for a small increase in population dose, as defined by percentincrease in dose (i.e., <1.0% person-rem/yr). However, since the IP2 and IP3 results rely onthat criterion rather than the absolute increase in dose criteria (i.e., < 1.0 person-rem/yr),additional information is provided to further demonstrate that the percent increase in dosecriteria is not exceeded.To do this, a reasonable estimate for the base case dose risk associated with external eventsmust be determined. In this case, each EPRI accident class is re-examined considering thepotential contribution for external events. Since the Class 1 frequency is determined based onremaining contribution not assigned to other classes, the discussion appears in reverse orderstarting with EPRI Class 8 and ending with EPRI Class 1. However, EPRI Class 2 is discussedprior to Class 3 since its value is used in the final determination of the Class 3 frequencies.P0247130002-47225-36 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsClass 8 SequencesThis group represents sequences where containment bypass occurs (SGTR or ISLOCA).ISLOCA and SGTR initiators are deemed inapplicable to the external events assessment so onlyinduced SGTR scenarios need to be considered. From the frequency information provided inTable 4.2-1 for IP2, the induced SGTR contribution to core damage is about 0.75% and for IP3it represented about 0.39%. A value of 0.5% is assumed for the external events contributionfor both IP2 and IP3. A High Early release magnitude dose is assigned.For IP2:Class_8 = 0.005 * [IP2 External Events CDF]= 0.005 * [5.01E-05]= 2.51E-07/yrFor IP3:Class_8 = 0.005 * [IP3 External Events CDF]= 0.005 * [5.20E-05]= 2.60E-07/yrClass 7 SeauencesThis group represents containment failure induced by early and late severe accidentphenomena. From Table 5.1-1 for IP2, the contribution from the early Class 7 sequences isabout 0.6% and for IP3 it represented about 1.3%. A value of 1.0% is assumed for theexternal events contribution for both IP2 and IP3. A High Early release magnitude dose isassigned. From Table 5.1-1 for IP2, the contribution from the late Class 7 sequences is about23% and for IP3 it represented about 15%. However, since the external events contributorsare more dominated by unrecoverable SBO-like scenarios, a value of 50% is assumed for theexternal events contribution for both IP2 and IP3. A High Late release magnitude dose isassigned.P0247130002-47225-37 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsFor IP2:Class_7-CFE = 0.01 * [IP2 External Events CDF]= 0.01 * [5.01E-05]= 5.01E-07/yrClass_7-CFL = 0.50 * [IP2 External Events CDF]= 0.50 * [5.01E-05]= 2.51E-05/yrForlP3:Class_7-CFE = 0.01 * [IP3 External Events CDF]= 0.01 * [5.20E-05]= 5.20E-07/yrClass_7-CFL = 0.50 * [IP3 External Events CDF]= 0.50 * [5.20E-05]= 2.60E-05/yrClass 4, 5. and 6 SequencesSimilar to the internal events assessment, because these failures are unaffected by the Type AILRT, these groups are not evaluated any further in this analysis.Class 2 SequencesThis group consists of large containment isolation failures. From the frequency informationprovided in Table 4.2-1 for IP2, the internal events contribution to this accident class wasapproximately 0.1% of the CDF and for IP3 it represented about 0.03%. Since seismic andfire initiated events would likely be more susceptible to this failure mode, the largercontribution of 0.1% is assumed for both IP2 and IP3. The population doses are assigned thesame as the Class 2 scenarios in the internal events assessment.P0247130002-47225-38 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsForIP2:Class_2 = 0.001 * [IP2 External Events CDF]= 0.001 * [5.01E-05]= 5.01E-08/yrFor IP3:Class_2 = 0.001 * [IP3 External Events CDF]= 0.001 * [5.20E-05]= 5.20E-08/yrClass 3 SequencesSimilar to the internal events assessment, the respective frequencies peras follows:year are determinedPROBciass_3aPROBclass_3b= probability of small pre-existing containment liner leakage= 0.0092 (see Section 4.3)= probability of large pre-existing containment liner leakage= 0.0023 (see Section 4.3)As described in Section 4.3, additional consideration is made to not apply these failureprobabilities to those cases that are already considered LERF scenarios (i.e., the Class 2, Class7, and Class 8 LERF contributions). This adjustment is made for based on the frequencyinformation described above for IP2 and IP3, respectively as shown below.For IP2:Class_3a = 0.0092 * [CDF -(Class 2 + Class 7-CFE + Class 8)]= 0.0092 * [5.01E-05 -(5.01E-08 + 5.01E-07 + 2.51E-07)]= 4.54E-07/yrClass_3b = 0.0023 * [CDF -(Class 2 + Class 7-CFE + Class 8)]= 0.0023 * [5.01E-05 -(5.01E-08 + 5.01E-07 + 2.51E-07)]= 1.13E-07/yrP0247130002-47225-39 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsFor IP3:Class_3a = 0.0092 * [CDF -(Class 2 + Class 7-CFE + Class 8)]= 0.0092 * [5.20E-05 -(5.20E-08 + 5.20E-07 + 2.60E-07)]= 4.71E-07/yrClass_3b = 0.0023 * [CDF -(Class 2 + Class 7-CFE + Class 8)]= 0.0023 * [5.20E-05 -(5.20E-08 + 5.20E-07 + 2.60E-07)]= 1.18E-07/yrFor this analysis, the associated containment leakage for Class 3a is 1OLa and 10OLa for Class3b, which is consistent with the latest EPRI methodology [3] and the NRC SE [7].Class 1 SequencesSimilar to the internal events assessment, the frequency is determined by subtracting allcontainment failure end states including the EPRI/NEI Class 3a and 3b frequency calculatedbelow, from the total CDF. The internal events intact containment dose of 4.41E+04person-rem for IP2 and IP3 is also utilized.Summary of Alternative External Events Base Case Dose AssessmentIn summary, the accident sequence frequencies that can lead to release of radionuclides to thepublic have been derived in a manner consistent with the definition of accident classes definedin EPRI 1018243 [3]. These frequencies have been combined with reasonable assumptionsregarding the population dose associated with each class to determine the base casepopulation dose risk for external events. This information is provided in Table 5.7-5a for IP2and in Table 5.7-5b for IP3. Additionally, following the same EPRI methodology utilized forinternal events to determine the risk impact assessment of extending the ILRT interval, theexternal events accident class frequencies indicative of a 15 year ILRT interval are provided inTable 5.7-6a for IP2 and in Table 5.7-6b for IP3.Table 5.7-7 then shows the changes due to the ILRT extension from 3 year to a 15 yearinterval in the LERF, person-rem/yr, and CCFP figures of merit. When these values are addedto the internal events results, the acceptance criteria are all still met by using this detailedalternative external events evaluation instead of the simple evaluation that was utilized inSection 5.7.5. A comparison to the acceptance criteria is also shown in Table 5.7-7. Note thatthe ALERF, person-rem/yr, and change in CCFCP shown in Table 5.7-7 are all slightly higherthan the corresponding values shown in Table 5.7-3. This is because the simple method inTable 5.7-3 assumes the same distribution of LERF contributors exists between the internalP0247130002-47225-40 Risk Impact Assessment of Extending the Indian Point ILRT Intervalsand external events models whereas the alternative assessment re-apportions the base caseLERF contributions based on more realistic assumptions while conservatively maintaining thetotal CDF value. That is, since the contribution from SGTR initiators and ISLOCA initiators(which contribute to the base LERF value) are not applicable to the external eventscontribution, more of the remaining CDF distribution is potentially affected by the ILRTextension as represented by the Class 3b multiplier on CDF (that is not already LERF).Additionally, the alternative detailed assessment leads to slightly different percent increases inperson-rem/yr which are a function of the base case dose estimates.TABLE 5.7-5APOPULATION DOSE RISK AS A FUNCTION OF ACCIDENT CLASS(IP2 ALTERNATIVE EXTERNAL EVENTS BASE CASE)ACCIDENT DESCRIPTION FREQUENCY DOSE DOSE RISKCLASS (1/YR) (PERSON- (PERSON-(CONTAINMENT REM) REM/YR)RELEASE TYPE)1 Containment Intact 2.37E-05 4.41E+04 1.04E+002 Large Isolation Failures 5.01E-08 6.51E+07 3.26E+00(Failure to Close)3a Small Isolation Failures (liner 4.54E-07 4.41E+05 2.OOE-01breach)3b Large Isolation Failures (liner 1.13E-07 4.41E+06 5.OOE-01breach)4 Small Isolation Failures N/A N/A N/A(Failure to seal -Type B)5 Small Isolation Failures N/A N/A N/A(Failure to seal-Type C)6 Other Isolation Failures (e.g., N/A N/A N/Adependent failures)7-CFE Failures Induced by 5.01E-07 6.51E+07 3.26E+01Phenomena (Early)7-CFL Failures Induced by 2.51E-05 1.63E+07 4.08E+02Phenomena (Late)8-SGTR Containment Bypass (Steam 2.51E-07 6.51E+07 1.63E+01Generator Tube Rupture)8-ISLOCA Containment Bypass O.OOE+00 6.51E+07 O.OOE+00(Interfacing System LOCA)CDF All CET End States 5.01E-05 462.2(Including Intact Case)P0247130002-47225-41 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-5BPOPULATION DOSE RISK AS A FUNCTION OF ACCIDENT CLASS(IP3 ALTERNATIVE EXTERNAL EVENTS BASE CASE)ACCIDENT DESCRIPTION FREQUENCY DOSE DOSE RISKCLASS (1/YR) (PERSON- (PERSON-(CONTAINMENT REM) REM/YR)RELEASE TYPE)1 Containment Intact 2.46E-05 4.41E+04 1.08E+002 Large Isolation Failures 5.20E-08 5.08E+07 2.64E+00(Failure to Close)3a Small Isolation Failures (liner 4.71E-07 4.41E+05 2.08E-01breach)3b Large Isolation Failures (liner 1.18E-07 4.41E+06 5.19E-01breach)4 Small Isolation Failures N/A N/A N/A(Failure to seal -Type B)5 Small Isolation Failures N/A N/A N/A(Failure to seal-Type C)6 Other Isolation Failures (e.g., N/A N/A N/Adependent failures)7-CFE Failures Induced by 5.20E-07 5.08E+07 2.64E+01Phenomena (Early)7-CFL Failures Induced by 2.60E-05 1.63E+07 4.24E+02Phenomena (Late)8-SGTR Containment Bypass (Steam 2.60E-07 5.08E+07 1.32E+01Generator Tube Rupture)8-ISLOCA Containment Bypass O.OOE+00 5.08E+07 O.OOE+00(Interfacing System LOCA)CDF All CET End States 5.20E-05 467.9(Including Intact Case)P0247130002-47225-42 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-6APOPULATION DOSE RISK AS A FUNCTION OF ACCIDENT CLASS (IP2 ALTERNATIVEEXTERNAL EVENTS EVALUATION CHARACTERISTIC OF CONDITIONS FOR 1 IN 15YEAR ILRT FREQUENCY)ACCIDENT DESCRIPTION FREQUENCY DOSE DOSE RISKCLASS (1/YR) (PERSON- (PERSON-(CONTAINMENT REM) REM/YR)RELEASE TYPE)1 Containment Intact 2.14E-05 4.41E+04 9.44E-012 Large Isolation Failures 5.01E-08 6.51E+07 3.26E+00(Failure to Close)3a Small Isolation Failures (liner 2.27E-06 4.41E+05 1.OOE+00breach)3b Large Isolation Failures (liner 5.67E-07 4.41E+06 2.50E+00breach)4 Small Isolation Failures N/A N/A N/A(Failure to seal -Type B)5 Small Isolation Failures N/A N/A N/A(Failure to seal-Type C)6 Other Isolation Failures (e.g., N/A N/A N/Adependent failures)7-CFE Failures Induced by 5.01E-07 6.51E+07 3.26E+01Phenomena (Early)7-CFL Failures Induced by 2.51E-05 1.63E+07 4.08E+02Phenomena (Late)8-SGTR Containment Bypass (Steam 2.51E-07 6.51E+07 1.63E+01Generator Tube Rupture)8-ISLOCA Containment Bypass O.OOE+00 6.51E+07 O.OOE+00_ (Interfacing System LOCA)CDF All CET End States 5.01E-05 [ 464.9_ (Including Intact Case) IP0247130002-47225-43 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-6BPOPULATION DOSE RISK AS A FUNCTION OF ACCIDENT CLASS (IP3 ALTERNATIVEEXTERNAL EVENTS EVALUATION CHARACTERISTIC OF CONDITIONS FOR 1 IN 15YEAR ILRT FREQUENCY)ACCIDENT DESCRIPTION FREQUENCY DOSE DOSE RISKCLASS (1/YR) (PERSON- (PERSON-(CONTAINMENT REM) REM/YR)RELEASE TYPE)1 Containment Intact 2.22E-05 4.41E+04 9.80E-012 Large Isolation Failures 5.20E-08 5.08E+07 2.64E+00(Failure to Close)3a Small Isolation Failures (liner 2.35E-06 4.41E+05 1.04E+00breach)3b Large Isolation Failures (liner 5.88E-07 4.41E+06 2.59E+00breach)4 Small Isolation Failures N/A N/A N/A(Failure to seal -Type B)5 Small Isolation Failures N/A N/A N/A(Failure to seal-Type C)6 Other Isolation Failures (e.g., N/A N/A N/Adependent failures)7-CFE Failures Induced by 5.20E-07 5.08E+07 2.64E+01Phenomena (Early)7-CFL Failures Induced by 2.60E-05 1.63E+07 4.24E+02Phenomena (Late)8-SGTR Containment Bypass (Steam 2.60E-07 5.08E+07 1.32E+01Generator Tube Rupture)8-ISLOCA Containment Bypass O.OOE+00 5.08E+07 O.OOE+00(Interfacing System LOCA)CDF All CET End States I 5.20E-05 470.7(Including Intact Case) IP0247130002-47225-44 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-7COMPARISON TO ACCEPTANCE CRITERIA INCLUDING ALTERNATIVEEXTERNAL EVENTS EVALUATION CONTRIBUTION FOR IP2 AND IP3Contributor ALERF APerson-rem/yr ACCFPIP2 Internal 9.84E-8/yr 0.584/yr (0.62%) 0.84%EventsIP2 External 4.54E-7/yr 2.70/yr (0.58%) 0.91%EventsIndian Point 2 5.52E-7/yr 3.28/yr (0.59%) 0.89%TotalIP3 Internal 1.26E-7/yr 0.751/yr (0.93%) 0.85%EventsIP3 External 4.71E-7/yr 2.80/yr (0.60%) 0.91%EventsIndian Point 3 5.96E-7/yr 3.55/yr (0.65%) 0.89%TotalAcceptance < 1.OE-6/yr <1.0 person- < 1.5%0/Criteria rem/yr or <1.0%The 5.52E-07/yr increase in LERF for IP2 and the 5.97E-07/yr increase in LERF for IP3 due tothe combined internal and external events from extending the ILRT frequency from 3-per-10years to 1-per-15 years falls within Region II between 1.0E-7 to 1.0E-6 per reactor year("Small Change" in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when thecalculated increase in LERF due to the proposed plant change is in the "Small Change" range,the risk assessment must also reasonably show that the total LERF is less than 1.0E-5/yr.From Table 4.2-1, the total LERF due to postulated internal event accidents is 1.16E-06/yr forIP2 and 1.25E-06/yr for IP3. From Table 5.7-5a for IP2, the base external events LERF can bederived from the Class 2, Class 3b, Class 7-CFE, and Class 8 contributions. From the individualcontributions of 5.01E-08/yr + 1.13E-07/yr + 5.01E-07/yr + 2.51E-07/yr, this equates to9.15E-07/yr. From Table 5.7-5b for IP3, the individual contributions of 5.20E-08/yr +1.18E-07/yr + 5.20E-07/yr + 2.60E-07/yr result in a total base case LERF from externalevents of 9.50E-07/yr. The total LERF values for IP2 and IP3 using the alternative externalevents evaluation are then shown in Table 5.7-8.P0247130002-47225-45 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 5.7-8IMPACT OF 15-YR ILRT EXTENSION ON LERF FOR IP2 AND IP3LERF CONTRIBUTOR IP2 (1/YR) IP3 (1/YR)Internal Events LERF 1.16E-06 1.25E-06External Events LERF 9.15E-07 9.50E-07Internal Events LERF due to 1.23E-07 1.58E-07ILRT (at 15 years) (1)External Events LERF increasedue to ILRT extension (2) 4.54E-07 4.71E-07Total 2.65E-06/yr 2.83E-06/yr(1) Including age adjusted steel liner corrosion likelihood as reported in Table 5.7-2a for IP2and Table 5.7-2b for IP3.(2) As shown in Table 5.7-7. This did not include the age adjusted steel liner corrosionlikelihood, but this was demonstrated to be a small contributor for IP2 and IP3.As can be seen, the total LERF for IP2 is estimated as 2.65E-06/yr and for IP3 it is2.83E-06/yr. These values are both less than the RG 1.174 requirement to demonstrate thatthe total LERF due to internal and external events is less than 1.OE-5/yr.P0247130002-47225-46 Risk Impact Assessment of Extending the Indian Point ILRT Intervals5.8 CONTAINMENT OVERPRESSURE IMPACTS ON CDFFor IP2 and IP3, ECCS NPSH calculations made in support of the GSI-191 effort [28, 29]confirmed that containment overpressure is not required to obtain adequate NPSH [30]. Thisis consistent with the PRA models which indicate there is no impact on CDF from the ILRTextension risk assessment.In IP-CALC-06-000231 [28], the NPSHA / NPSHR relationship for IP2 ECCS pumpswas being evaluated. For conservatism in obtaining the NPSHA and NPSHR, themaximum volumetric flow rate was used. The greatest volumetric flow rate occurswhen the least dense fluid is being pumped. This is at the highest temperature in therecirculation phase of the accident. For IP2, this temperature was 264.4 F whichoccurs at start of recirculation. Since 264.4 F is higher than 212 F, a boundarycondition pressure of 37.6 psia is inputted. This is close to the saturation pressure at264.4 F so there is essentially no containment overpressure being invoked. In otherwords, 264.4 F and 37.6 psia is basically equivalent to 212 F and 14.7 psia (0 psig).* The same issue was addressed in IP-CALC-07-00054 [29] for the TP3 NPSHA /NPSHR evaluation. Again, to be most conservative with respect to NPSHA andNPSHR, the maximum volumetric flow rate has to be used. This entails that thehighest temperature during recirculation applies. This is 242.8 F at commencementof recirculation. The saturation pressure at 242.8 F is close to 26.1 psia, which is theboundary condition pressure input in the calculation. Again, essentially nocontainment overpressure is being invoked since 242.8 F and 26.1 psia is basicallyequivalent to 212 F and 14.7 psia (0 psig).P0247130002-47225-47 Risk Impact Assessment of Extending the Indian Point ILRT Intervals6.0 SENSITIVITIES6.1 SENSITIVITY TO CORROSION IMPACT ASSUMPTIONSThe results in Tables 5.2-2a(b), 5.3-la(b), and 5.3-2a(b) show that including corrosion effectscalculated using the assumptions described in Section 4.4 does not significantly affect theresults of the ILRT extension risk assessment. In any event, sensitivity cases were developedto gain an understanding of the sensitivity of the results to the key parameters in the corrosionrisk analysis. The time for the flaw likelihood to double was adjusted from every five years toevery two and every ten years. The failure probabilities for the cylinder, dome and basematwere increased and decreased by an order of magnitude. The total detection failure likelihoodwas adjusted from 10% to 15% and 5%. The results are presented in Table 6.1-1a for IP2and in Table 6.1-1b for IP3. In every case, the impact from including the corrosion effects isvery minimal. Even the upper bound estimates with very conservative assumptions for all ofthe key parameters yield increases in LERF due to corrosion of only 3.68E-8/yr for IP2 and4.72E-08/yr for IP3. The results indicate that even with very conservative assumptions, theconclusions from the base analysis would not change.TABLE 6.1-1ASTEEL LINER CORROSION SENSITIVITY CASES FOR IP2AGE CONTAINMENT VISUAL INCREASE IN CLASS 3B(STEP 3 IN THE BREACH INSPECTION FREQUENCY (LERF)CORROSION (STEP 4 IN THE & NON- FOR ILRT EXTENSIONANALYSIS) CORROSION VISUAL FROM 3 IN 10 TO 1 IN 15 YEARSANALYSIS) FLAWS (PER YEAR)(STEP 5 IN THECORROSION TOTAL INCREASE DUE TOANALYSIS) INCREASE CORROSIONBase Case Base Case Base Case 9.84E-08 1.16E-09Doubles every (1.0% Cylinder- (10% Cylinder-5 yrs Dome, Dome,0.1% Basemat) 100% Basemat)Doubles every Base Base 9.99E-08 2.63E-092 yrsDoubles every Base Base 9.83E-08 9.68E-1010 yrsBase Base 15% Cylinder- 9.89E-08 1.62E-09DomeP0247130002-47226-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 6.1-1ASTEEL LINER CORROSION SENSITIVITY CASES FOR IP2AGE CONTAINMENT VISUAL INCREASE IN CLASS 3B(STEP 3 IN THE BREACH INSPECTION FREQUENCY (LERF)CORROSION (STEP 4 IN THE & NON- FOR ILRT EXTENSIONANALYSIS) CORROSION VISUAL FROM 3 IN 10 TO 1 IN 15 YEARSANALYSIS) FLAWS (PER YEAR)(STEP 5 IN THECORROSION TOTAL INCREASE DUE TOANALYSIS) INCREASE CORROSIONBase Base 5% Cylinder- 9.79E-08 6.97E-10DomeBase 10% Cylinder- Base 1.09E-07 1.16E-08Dome,1% BasematBase 0.1% Cylinder- Base 9.74E-08 1.16E-10Dome,0.01% BasematLOWER BOUNDDoubles every 0.1% Cylinder- 5% Cylinder- 9.73E-08 5.81E-1110 yrs Dome, Dome,0.01% Basemat 100% BasematUPPER BOUNDDoubles every 10% Cylinder- 15% Cylinder- 1.34E-07 3.68E-082 yrs Dome, Dome,1% Basemat 100% BasematP0247130002-47226-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 6.1-1BSTEEL LINER CORROSION SENSITIVITY CASES FOR IP3AGE CONTAINMENT VISUAL INCREASE IN CLASS 3B(STEP 3 IN THE BREACH INSPECTION FREQUENCY (LERF)CORROSION (STEP 4 IN THE & NON- FOR ILRT EXTENSIONANALYSIS) CORROSION VISUAL FROM 3 IN 10 TO 1 IN 15 YEARSANALYSIS) FLAWS (PER YEAR)(STEP 5 IN THECORROSION TOTAL INCREASE DUE TOANALYSIS) INCREASE CORROSIONBase Case Base Case Base Case 1.26E-07 1.49E-09Doubles every (1.0% Cylinder- (10% Cylinder-5 yrs Dome, Dome,0.1% Basemat) 100% Basemat)Doubles every Base Base 1.28E-07 3.37E-092 yrsDoubles every Base Base 1.26E-07 1.24E-0910 yrsBase Base 15% Cylinder- 1.27E-07 2.08E-09DomeBase Base 5% Cylinder- 1.26E-07 8.95E-10DomeBase 10% Cylinder- Base 1.40E-07 1.49E-08Dome,1% BasematBase 0.1% Cylinder- Base 1.25E-07 1.49E-10Dome,0.01% BasematLOWER BOUNDDoubles every 0.1% Cylinder- 5% Cylinder- 1.25E-07 7.47E-1110 yrs Dome, Dome,0.01% Basemat 100% BasematUPPER BOUNDDoubles every 100/a Cylinder- 15% Cylinder- 1.72E-07 4.72E-082 yrs Dome, Dome,1% Basemat 100% BasematP0247130002-47226-3 Risk Impact Assessment of Extending the Indian Point ILRT Intervals6.2 EPRI EXPERT ELICITATION SENSITIVITYAn expert elicitation was performed to reduce excess conservatisms in the data associated withthe probability of undetected leaks within containment [3]. Since the risk impact assessmentof the extensions to the ILRT interval is sensitive to both the probability of the leakage as wellas the magnitude, it was decided to perform the expert elicitation in a manner to solicit theprobability of leakage as a function of leakage magnitude. In addition, the elicitation wasperformed for a range of failure modes which allowed experts to account for the range offailure mechanisms, the potential for undiscovered mechanisms, inaccessible areas of thecontainment as well as the potential for. detection by alternate means. The expert elicitationprocess has the advantage of considering the available data for small leakage events, whichhave occurred in the data, and extrapolate those events and probabilities of occurrence to thepotential for large magnitude leakage events.The basic difference in the application of the ILRT interval methodology using the expertelicitation is a change in the probability of pre-existing leakage within containment. The basecase methodology uses the Jeffrey's non-informative prior for the large leak size and theexpert elicitation sensitivity study uses the results from the expert elicitation. In addition,given the relationship between leakage magnitude and probability, larger leakage that is morerepresentative of large early release frequency can be reflected. For the purposes of thissensitivity, the same leakage magnitudes that are used in the base case methodology (i.e.,1OLa for small and 10OLa for large) are used here. Table 6.2-1 illustrates the magnitudes andprobabilities of a pre-existing leak in containment associated with the base case and the expertelicitation statistical treatments. These values are used in the ILRT interval extension for thebase methodology and in this sensitivity case. Details of the expert elicitation process,including the input to expert elicitation as well as the results of the expert elicitation, areavailable in the various appendices of EPRI 1018243 [3].TABLE 6.2-1EPRI EXPERT ELICITATION RESULTSLEAKAGE SIZE (LA) BASE CASE MEAN EXPERT PERCENTPROBABILITY OF ELICITATION MEAN REDUCTIONOCCURRENCE PROBABILITY OFOCCURRENCE [3]10 9.2E-03 3.88E-03 58%100 2.3E-03 2.47E-04 89%P0247130002-47226-4 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsThe summary of results using the expert elicitation values for probability of containmentleakage is provided in Table 6.2-2a for IP2 and in Table 6.2-2b for 1P3. As mentionedpreviously, probability values are those associated with the magnitude of the leakage used inthe base case evaluation (1OLa for small and 10OLa for large). The expert elicitation processproduces a relationship between probability and leakage magnitude in which it is possible toassess higher leakage magnitudes that are more reflective of large early releases; however,these evaluations are not performed in this particular study.The net effect is that the reduction in the multipliers shown above also leads to a dramaticreduction on the calculated increases in the LERF values. As shown in Table 6.2-2a for IP2, theincrease in the overall value for LERF due to Class 3b sequences that is due to increasing theILRT test interval from 3 to 15 years is just 1.05E-08/yr. Similarly, the increase due toincreasing the interval from 10 to 15 years is just 4.40E-09/yr. As shown in Table 6.2-2b for1P3, the increase in the overall value for LERF due to Class 3b sequences that is due toincreasing the ILRT test interval from 3 to 15 years is just 1.34E-08/yr. Similarly, the increasedue to increasing the interval from 10 to 15 years is just 5.60E-09/yr. As such, if the expertelicitation probabilities of occurrence are used instead of the non-informative prior estimates,the change in LERF for IP2 and IP3 is within the range of a "very small" change in risk whencompared to the current 1-in-10, or baseline 3-in-10 year requirement. Additionally, as shownin Table 6.2-2a for IP2 and Table 6.2-2b for IP3, the increase in dose rate and CCFP aresimilarly reduced to much smaller values. The results of this sensitivity study are judged to bemore indicative of the actual risk associated with the ILRT extension than the results from theassessment as dictated by the values from the EPRI methodology [3], and yet are stillconservative given the assumption that all of the Class 3b contribution is considered to beLERF.P0247130002-47226-5 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 6.2-2AIP2 ILRT CASES:3 IN 10 (BASE CASE), 1 IN 10, AND 1 IN 15 YR INTERVALS(BASED ON EPRI EXPERT ELICITATION LEAKAGE PROBABILITIES)EPRI DOSE BASE CASE EXTEND TO EXTEND TOCLASS PER-REM 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARSCDF PERSON- CDF PERSON- CDF PERSON-(1/YR) REM/YR (1/YR) REM/YR (1/YR) REM/YR1 4.41E+04 7.82E-06 3.45E-01 7.71E-06 3.40E-01 7.64E-06 3.37E-012 6.51E+07 1.11E-08 7.23E-01 1.11E-08 7.23E-01 1.11E-08 7.23E-013a 4.41E+05 4.10E-08 1.81E-02 1.37E-07 6.03E-02 2.05E-07 9.05E-023b 4.41E+06 2.61E-09 1.15E-02 8.70E-09 3.84E-02 1.31E-08 5.76E-027-CFE 6.22E+07 7.37E-08 4.58E+00 7.37E-08 4.58E+00 7.37E-08 4.58E+007-CFL 6.87E+06 2.71E-06 1.86E+01 2.71E-06 1.86E+01 2.71E-06 1.86E+018-SGTR 6.51E+07 1.05E-06 6.80E+01 1.05E-06 6.80E+01 1.05E-06 6.80E+018-ISLOCA 6.51E+07 2.77E-08 1.80E+00 2.77E-08 1.80E+00 2.77E-08 1.80E+00Total 1.17E-05 9.414E+01 1.17E-05 9.421E+01 1.17E-05 19.425E+01ILRT Dose Rate from 2.96E-02 9.86E-02 1.48E-013a and 3bDelta From 3 yr --- 6.45E-02 1.11E-01TotalDose From 10 yr --- 4.62E-02DoseRate(1)3b Frequency (LERF) 2.61E-09 8.70E-09 1.31E-08Delta 3b From 3 yr --- 6.09E-09 1.05E-08LERF From 10 yr .... -- 4.40E-09CCFP % 33.00% 33.05% 33.09%Delta From 3 yr --- 0.05% 0.09%CCFP %From 10 yr --- 0.04%(1) The overall difference in total dose rate is less than the difference of only the 3a and 3bcategories between two testing intervals. This is due to the fact that the Class 1 person-rem/yr decreases when extending the ILRT frequency.P0247130002-47226-6 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsTABLE 6.2-2BIP3 ILRT CASES:3 IN 10 (BASE CASE), 1 IN 10, AND 1 IN 15 YR INTERVALS(BASED ON EPRI EXPERT ELICITATION LEAKAGE PROBABILITIES)EPRI DOSE BASE CASE EXTEND TO EXTEND TOCLASS PER-REM 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARSCDF PERSON- CDF PERSON- CDF PERSON-(1/YR) REM/YR (1/YR) REM/YR (1/YR) [ REM/YR1 4.41E+04 1.12E-05 4.96E-01 1.11E-05 4.90E-01 1.10E-05 4.86E-012 5.08E+07 3.99E-09 2.03E-01 3.99E-09 2.03E-01 3.99E-09 2.03E-013a 4.41E+05 5.27E-08 2.32E-02 1.76E-07 7.74E-02 2.64E-07 1.16E-013b 4.41E+06 3.36E-09 1.48E-02 1.12E-08 4.93E-02 1.68E-08 7.40E-027-CFE 3.17E+07 1.88E-07 5.97E+00 1.88E-07 5.97E+00 1.88E-07 5.97E+007-CFL 6.85E+06 2.17E-06 1.49E+01 2.17E-06 1.49E+01 2.17E-06 1.49E+018-SGTR 5.08E+07 9.77E-07 4.96E+01 9.77E-07 4.96E+01 9.77E-07 4.96E+018-ISLOCA 5.08E+07 1.93E-07 9.80E+00 1.93E-07 9.80E+00 1.93E-07 9.80E+00Total 1.48E-05 8.099E+01 1.48E-05 18.108E+01 I 1.48E-05 18.114E+01ILRT Dose Rate from 3.81E-02 1.27E-01 1.90E-013a and 3bDelta From 3 yr --- 8.29E-02 1.42E-01TotalDose From 10 yr --- 5.94E-02DoseRate*1)3b Frequency (LERF) 3.36E-09 1.12E-08 1.68E-08Delta 3b From 3 yr --- 7.84E-09 1.34E-08LERF IFrom 10 yr ....5.60E-09CCFP % 23.84% 23.89% 23.93%Delta From 3 yr --- 0.05% 0.09%CCFP %From 10 yr --.--- 0.04%( The overall difference in total dose rate is less than the difference of only the 3a and 3bcategories between two testing intervals. This is due to the fact that the Class 1 person-rem/yr decreases when extending the ILRT frequency.P0247130002-47226-7 Risk Impact Assessment of Extending the Indian Point ILRT Intervals

7.0 CONCLUSION

SBased on the results from Section 5 and the sensitivity calculations presented in Section 6, thefollowing conclusions regarding the assessment of the plant risk are associated withpermanently extending the Type A ILRT test frequency to fifteen years:* Reg. Guide 1.174 [4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Reg. Guide 1.174 defines "very small"changes in risk as resulting in increases of CDF below 1.OE-06/yr and increases inLERF below 1.OE-07/yr. "Small" changes in risk are defined as increases in CDFbelow 1.OE-05/yr and increases in LERF below 1.OE-06/yr. Since the ILRT extensionwas demonstrated to have no impact on CDF for IP2 and IP3, the relevant criterion isLERF. The increase in internal events LERF resulting from a change in the Type AILRT test interval for the base case with corrosion included for IP2 is 9.84E-08/yr(see Table 5.6-1a). In using the EPRI Expert Elicitation methodology, the change isestimated as 1.05E-08/yr (see Table 6.2-2a). Both of these values fall within thevery small change region of the acceptance guidelines in Reg. Guide 1.174. For IP3,the increase is estimated at 1.26E-07/yr (see Table 5.6-1b), which is within thesmall change region of the acceptance guidelines in Reg. Guide 1.174. In using theEPRI Expert Elicitation methodology, the change is estimated as 1.34E-08/yr (seeTable 6.2-2b), which is within the very small change region of the acceptanceguidelines in Reg. Guide 1.174.* The change in dose risk for changing the Type A test frequency from three-per-tenyears to once-per-fifteen-years, measured as an increase to the total integrated doserisk for all internal events accident sequences for IP2, is 0.584 person-rem/yr(0.62%) using the EPRI guidance with the base case corrosion case (Table 5.6-1a).The change in dose risk drops to 1.11E-01 person-rem/yr when using the EPRIExpert Elicitation methodology (Table 6.2-2a). For IP3, it is 0.751 person-rem/yr(0.93%) using the EPRI guidance with the base case corrosion case (Table 5.6-1b).The change in dose risk drops to 1.42E-01 person-rem/yr when using the EPRIExpert Elicitation methodology (Table 6.2-2b). The values calculated per the EPRIguidance are all lower than the acceptance criteria of 51.0 person-rem/yr or <1.0%person-rem/yr defined in Section 1.3.* The increase in the conditional containment failure frequency from the three in tenyear interval to one in fifteen years including corrosion effects using the EPRIguidance (see Section 5.5) is 0.84% for IP2 and 0.85% for IP3. This value drops toless that 0.10% for IP2 and IP3 using the EPRI Expert Elicitation methodology (seeTable 6.2-2a and Table 6.2-2b, respectively). This is below the acceptance criteria ofless than 1.5% defined in Section 1.3.* To determine the potential impact from external events, a bounding assessmentfrom the risk associated with external events utilizing information from the IP2 andIP3 IPEEEs similar to the approach used in the License Renewal SAMA analysis wasperformed. As shown in Table 5.7-2a for IP2, the total increase in LERF due tointernal events and the bounding external events assessment is 5.20E-07/yr. Asshown in Table 5.7-2b for IP3, the total increase in LERF due to internal events andthe bounding external events assessment is 5.70E-07/yr. Both of these values are inRegion II of the Reg. Guide 1.174 acceptance guidelines.P0247130002-47227-1 Risk Impact Assessment of Extending the Indian Point ILRT Intervals* As shown in Table 5.7-4, the same bounding analysis indicates that the total LERFfrom both internal and external risks is 6.78E-06/yr for IP2 and 6.34E-06/yr for IP3,which are less than the Reg. Guide 1.174 limit of 1.OE-05/yr given that the ALERF isin Region II (small change in risk)." Finally, since the external events assessment led to exceeding one of the twoalternative acceptance criteria (i.e. greater than 1.0 person-rem/yr, an alternativedetailed bounding external events assessment was also performed to demonstratethat the alternate 1.0% person-rem/yr criterion and the other acceptance criteriacould still be met. In this case, as shown in Table 5.7-7 for IP2, the total change inLERF from both internal and external events was 5.52E-7/yr, the change in person-rem/yr was 3.28/yr representing 0.59% of the total, and the change in the CCFP was0.89%. For IP3, the total change in LERF from both internal and external events was5.97E-7/yr, the change in person-rem/yr was 3.55/yr representing 0.65% of thetotal, and the change in the CCFP was 0.89%. All of these calculated changes meetthe acceptance criteria. As shown in Table 5.7-8, this assessment indicates that thetotal LERF from both internal and external risks is 2.65E-06/yr for IP2 and 2.83E-06/yr for IP3, which are less than the Reg. Guide 1.174 limit of 1.OE-05/yr given thatthe ALERF is in Region II (small change in risk).* Including age-adjusted steel liner corrosion effects in the ILRT assessment wasdemonstrated to be a small contributor to the impact of extending the ILRT intervalfor IP2 and IP3.Therefore, increasing the ILRT interval on a permanent basis to a one-in-fifteen year frequencyis not considered to be significant since it represents only a small change in the IP2 and IP3risk profiles.Previous AssessmentsThe NRC in NUREG-1493 [6] has previously concluded the following:* Reducing the frequency of Type A tests (ILRTs) from three per 10 years to one per20 years was found to lead to an imperceptible increase in risk. The estimatedincrease in risk is very small because ILRTs identify only a few potential containmentleakage paths that cannot be identified by Type B and C testing, and the leaks thathave been found by Type A tests have been only marginally above existingrequirements.* Given the insensitivity of risk to containment leakage rate and the small fraction ofleakage paths detected solely by Type A testing, increasing the interval betweenintegrated leakage-rate tests is possible with minimal impact on public risk. Theimpact of relaxing the ILRT frequency beyond one in 20 years has not beenevaluated. Beyond testing the performance of containment penetrations, ILRTs alsotest the integrity of the containment structure.The findings for IP2 and IP3 confirm these general findings on a plant specific basis consideringthe severe accidents evaluated, the containment failure modes, and the local populationsurrounding IP2 and IP3.P0247130002-47227-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy

8.0 REFERENCES

[1] Nuclear Energy Institute, Industry Guideline for Implementing Performance-BasedOption of 10 CFR Part 50, Appendix J, NEI 94-01, Revision 3-A, July 2012.[2] Electric Power Research Institute, Risk Impact Assessment of Revised ContainmentLeak Rate Testing Intervals, EPRI TR-104285, August 1994.[3] Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals:Revision 2-A of 1009325. EPRI, Palo Alto, CA: October 2008. 1018243.[4] U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic RiskAssessment in Risk-Informed Decisions on Plant-Specific Changes to the LicensingBasis, Regulatory Guide 1.174, Revision 2, May 2011.[5] Letter from Mr. C. H. Cruse (Constellation Nuclear, Calvert Cliffs Nuclear PowerPlant) to U.S. Nuclear Regulatory Commission, Response to Request for AdditionalInformation Concerning the License Amendment Request for a One-Time IntegratedLeakage Rate Test Extension, Accession Number ML020920100, March 27, 2002.[6] U.S. Nuclear Regulatory Commission, Performance-Based Containment Leak-TestProgram, NUREG-1493, September 1995.[7] U.S. Nuclear Regulatory Commission, Final Safety Evaluation for Nuclear EnergyInstitute (NEI) Topical Report (TR) 94-01, Revision 2, "Industry Guideline forImplementing Performance-Based Option Of 10 CFR Part 50, Appendix J" and ElectricPower Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007,"Risk Impact Assessment Of Extended Integrated Leak Rate Testing Intervals" (TACNo. MC9663), Accession Number ML081140105, June 25, 2008.[8] Consolidated Edison Company of New York, Individual Plant Examination for ExternalEvents for Indian Point Unit 2 Nuclear Generating Station, Revision 0, December1995.[9] New York Power Authority, Indian Point Three Nuclear Power Plant Individual PlantExamination for External Events, IP3-RPT-UNSPEC-02182, Revision 0, September1997.[10] Entergy Nuclear, Re-analysis of MACCS2 Models for IPEC, Calculation IP-CALC-09-00265, December 2009.[11] Entergy Nuclear, MAAP/MACCS2 Computer Codes Calculated Dose for IPECContainment Structure Based on Allowable Leakage From an Intact Containment,Calculation IP-CALC-13-00042, September 2013.[12] ERIN Engineering and Research, Shutdown Risk Impact Assessment for ExtendedContainment Leakage Testing Intervals Utilizing ORAMTM, EPRI TR-105189, FinalReport, May 1995.[13] Oak Ridge National Laboratory, Impact of Containment Building Leakage on LWRAccident Risk, NUREG/CR-3539, ORNL/TM-8964, April 1984.[14] Pacific Northwest Laboratory, Reliability Analysis of Containment Isolation Systems,NUREG/CR-4220, PNL-5432, June 1985.P0247130002-47228-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy[15] U.S. Nuclear Regulatory Commission, Technical Findings and Regulatory Analysis forGeneric Safety Issue II.E.4.3 (Containment Integrity Check), NUREG-1273, April1988.[16] Pacific Northwest Laboratory, Review of Light Water Reactor RegulatoryRequirements, NUREG/CR-4330, PNL-5809, Vol. 2, June 1986.[17] U.S. Nuclear Regulatory Commission, Severe Accident Risks: An Assessment for FiveU.S. Nuclear Power Plants, NUREG-1150, December 1990.[18] Entergy Nuclear, Indian Point Unit 2 Probabilistic Safety Assessment (PSA),Calculation IP-RPT-09-00026, Revision 0, November 2011.[19] Entergy Nuclear, Indian Point Unit 3 Probabilistic Safety Assessment (PSA),Calculation IP-RPT-10-00023, Revision 0, November 2012.[20] Entergy Nuclear, Indian Point Units 2 & 3, License Renewal Application, Appendix E,Applicant's Environmental Report, Accession Number ML071210530, April 23, 2007.[21] Letter from P. B. Cowan (Exelon Nuclear, Peach Bottom) to U.S. Nuclear RegulatoryCommission, Response to Request for Additional Information -License AmendmentRequest for Type A Test Extension, Accession Number ML100560433, February 25,2010.[22] Letter from P. B. Cowan (Exelon Nuclear, Peach Bottom) to U.S. Nuclear RegulatoryCommission, License Amendment Request -Type A Test Extension, AccessionNumber ML092440053, August 28, 2009.[23] Letter from Dave Morey (Southern Company, Farley Project) to U.S. NuclearRegulatory Commission, Joseph M. Farley Nuclear Plant Technical SpecificationRevision Request Integrated Leakage Rate Testing Interval Extension, NEL-02-0001,Accession Number ML020990040, April 4, 2002.[24] Letter from D.E. Young (Florida Power, Crystal River) to U.S. Nuclear RegulatoryCommission, License Amendment Request #267, Revision 1, Supplemental Risk-Informed Information in Support of License Amendment Request #267, Revision 0,3F0401-11, Accession Number ML011210207, April 25, 2001.[25] Letter from John A. Ventosa (Entergy, Indian Point Energy Center) to U.S. NuclearRegulatory Commission, Indian Point Nuclear Power Plant Units 2 and 3Reassessment of the Seismic Core Damage Frequency, NL-13-084, AccessionNumber ML13183A279, June 26, 2013.[26] Letter from Thomas P. Kirwin (Entergy, Palisades Nuclear Plant) to U.S. NuclearRegulatory Commission, License Amendment Request to Extend the ContainmentType A Leak Rate Test Frequency to 15 Years, Accession Number ML110970616,April 6, 2011.[27] U.S. Nuclear Regulatory Commission, Palisades Nuclear Plant -Issuance ofAmendment to Extend the Containment Type A Leak Rate Test Frequency to 15Years (TAC No. ME5997), Accession Number ML120740081, April 23, 2012.[28] Westinghouse, Indian Point Unit 2 SI Recirculation (LHSI and HHSI) Performance forthe Containment Sump Program, Entergy Calculation IP-CALC-06-00231, Revision 1,April 2010.P0247130002-47228-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacy[29] Westinghouse, Indian Point Unit 3 SI Recirculation (LHSI and HHSI) Performance forthe Containment Sump Program, Entergy Calculation IP-CALC-07-00054, Revision 2,June 2010.[30] E-Mail from D. Gaynor (Entergy) to D. Vanover (ERIN), FW: Inputs for NPSH Calcs,July 24, 2013.[31] U.S. Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October1975.P0247130002-47228-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyAppendix APRA Technical AdequacyP0247130002-4722 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyNote that the information provided in this appendix was provided by Entergy personnel.A. 1 OVERVIEWA technical Probabilistic Risk Assessment (PRA) analysis is presented in this report to helpsupport an extension of the IP2 and IP3 containment Type A test integrated leak rate test(ILRT) interval to fifteen years.The analysis follows the guidance provided in Regulatory Guide 1.200, Revision 2 [A.1], "AnApproach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results forRisk-Informed Activities." The guidance in RG-1.200 indicates that the following steps shouldbe followed to perform this study:1. Identify the parts of the PRA used to support the application" SSCs, operational characteristics affected by the application and how these areimplemented in the PRA model." A definition of the acceptance criteria used for the application.2. Identify the scope of risk contributors addressed by the PRA model* If not full scope (i.e. internal and external), identify appropriate compensatorymeasures or provide bounding arguments to address the risk contributors notaddressed by the model.3. Summarize the risk assessment methodology used to assess the risk of theapplication* Include how the PRA model was modified to appropriately model the risk impact ofthe change request.4. Demonstrate the Technical Adequacy of the PRA" Identify plant changes (design or operational practices) that have been incorporatedat the site, but are not yet in the PRA model and justify why the change does notimpact the PRA results used to support the application." Document peer review findings and observations that are applicable to the parts ofthe PRA required for the application, and for those that have not yet beenaddressed justify why the significant contributors would not be impacted." Document that the parts of the PRA used in the decision are consistent withapplicable standards endorsed by the Regulatory Guide. Provide justification toshow that where specific requirements in the standard are not met, it will notunduly impact the results." Identify key assumptions and approximations relevant to the results used in thedecision-making process.Items 1 through 3 are covered in the main body of this report. The purpose of this appendix isto address the requirements identified in item 4 above. Each of these items (plant changesP0247130002-4722A-1 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacynot yet incorporated into the PRA model, relevant peer review findings, consistency withapplicable PRA standards and the identification of key assumptions) are discussed in thefollowing sections.The risk assessment performed for the ILRT extension request is based on the current Level 1and Level 2 PRA models of record. Information developed for the license renewal effort tosupport the Level 2 release categories is also used in this analysis supplemented by additionalcalculations to more appropriately represent the intact containment case in the ILRT extensionrisk assessment.Note that for this application, the accepted methodology involves a bounding approach toestimate the change in the LERF from extending the ILRT interval. Rather than exercising thePRA model itself, it involves the establishment of separate evaluations that are linearly relatedto the plant CDF contribution. Consequently, a reasonable representation of the plant CDFthat does not result in a LERF does not require that Capability Category II be met in everyaspect of the modeling if the Category I treatment is conservative or otherwise does notsignificantly impact the results.As further discussed below, the PRA models used for this application are the latest models,which were released in November 2011 (for IP2) and November 2012 (for IP3). There are nosignificant plant changes (design or operational practices) that have not yet been incorporatedin those PRA models.A discussion of the Entergy model update process, the peer reviews performed on the IP2 andIP3 models, the results of those peer reviews and the potential impact of peer review findingson the ILRT extension risk assessment are provided in Section A.2. Section A.3 provides anassessment of key assumptions and approximations used in this assessment and Section A.4briefly summarizes the results of the PRA technical adequacy assessment with respect to thisapplication.A.2 PRA UPDATE PROCESS AND PEER REVIEW RESULTSA.2.1 IntroductionThe Indian Point Unit 2 (IP2) and Unit 3 (IP3) Probabilistic Risk Assessment (PRA) models usedfor this application [A.2 and A.3] are the most recent evaluations of the IP2 and IP3 riskprofiles for internal event challenges. The IP2 and IP3 PRA modeling is highly detailed,including a wide variety of initiating events, modeled systems, operator actions, and commonP0247130002-4722A-2 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacycause failure events. The PRA model quantification process is based on the event tree andfault tree methodology, which is a well-known methodology in the industry.Entergy employs a multi-faceted approach to establishing and maintaining the technicaladequacy 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 theuse of self-assessments and independent peer reviews. The following information describesthis approach as it applies to the IP2 and IP3 PRA models.A.2.2 PRA Maintenance and UpdateThe Entergy risk management process ensures that the applicable PRA model is an accuratereflection of the as-built and as-operated plant. This process is defined in the Entergy fleetprocedure EN-DC-151, "PSA Maintenance and Update" [A.4]. This procedure delineates theresponsibilities and guidelines for updating the full power internal events PRA models at alloperating Entergy nuclear power plants. In addition, the procedure also defines the processfor implementing regularly scheduled and interim PRA model updates, and for tracking issuesidentified as potentially affecting the PRA models (e.g., due to changes in the plant, industryoperating experience, etc.). To ensure that the current PRA model remains an accuratereflection 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 PRAmodel. Potential PRA model changes resulting from these reviews are entered intothe Model Change Request (MCR) database, and a determination is made regardingthe significance of the change with respect to current PRA model." New engineering calculations and revisions to existing calculations are reviewed fortheir impact on the PRA model.* Plant specific initiating event frequencies, failure rates, and maintenanceunavailabilities are updated approximately every four years, and* Industry standards, experience, and technologies are periodically reviewed to ensurethat any changes are appropriately incorporated into the models.In addition, following each periodic PRA model update, Entergy performs a self-assessment toassure that the PRA quality and expectations for all current applications are met. The EntergyPRA maintenance and update procedure requires updating of all risk informed applications thatmay have been impacted by the update.P0247130002-4722A-3 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyA.2.3 Regulatory Guide 1.200 PWROG Peer Review of the IP2 and IP3 Internal EventsPRA ModelsBoth the IP2 and IP3 internal events models went through a Regulatory Guide 1.200 PWROwners Group peer review using the NEI 05-04 process.The IP2 PRA internal events model peer review was performed in December 2009, and usedthe American Society of Mechanical Engineers PRA Standard RA-Sb-2005, and RegulatoryGuide 1.200 Revision 1. The IP3 PRA internal events model peer review was performed inDecember 2010. Since the IP3 peer review was later, it used RA-Sa-2009 (the AmericanSociety of Mechanical Engineers / American Nuclear Society Combined PRA Standard) andRegulatory Guide 1.200 Revision 2. As noted in the forward to the combined standard, theprimary purpose, in addition to combining internal and external events into a single standard,was to ensure consistency in format, organization, language, and level of detail. It was alsonoted that, among the criteria observed in assembling the component Standards were:(a) the requirements in the Standards would not be revised or modified(b) no new requirements would be includedAn internal comparison of the ASME standard to the combined ASME / ANS standard confirmedthat there were few substantive changes to the internal events portion of the standard,although the expected level of documentation was increased in some cases.The IP2 and IP3 PRA peer reviews addressed 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)P0247130002-4722A-4 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyDuring the IP2 and IP3 PRA model peer reviews, the technical elements identified above wereassessed with respect to Capability Category II criteria to better focus the SupportingRequirement assessments.A.2.4 Peer Review ResultsThe ASME PRA standards used for the IP2 and IP3 peer reviews each contained a total of 326numbered supporting requirements. A number of the supporting requirements weredetermined to be not applicable to the IP2 or IP3 PRA (e.g., BWR related, multi-site related).Of the applicable supporting requirements, 95% were satisfied at Capability Category II orgreater for IP2, and 97% were satisfied at Capability Category II criteria or greater for IP3.The Facts and Observations (F&Os) for the IP2 PRA peer review are provided in the report,entitled, "RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for theIndian Point 2 Nuclear Power Plant Probabilistic Risk Assessment" [A.5]. Of the 41 Facts andObservations (F&Os) generated by the Peer Review Team, 21 were considered Findings.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 theIndian Point 3 Probabilistic Risk Assessment" [A.6]. Of the 68 Facts and Observations (F&Os)generated by the Peer Review Team, 11 were considered Findings.As a result of the Regulatory Guide 1.200 PWROG peer reviews, all the F&Os (other than bestpractices) were identified as potential improvements to the IP2 and IP3 PRA models ordocumentation and were entered into the Entergy Model Change Request (MCR) database.Tables A.2-1 and A.2-2 contain the findings resulting from the peer review of each unit, thestatus of the resolution for each finding and the potential impact of each finding on thisapplication. In summary, a majority of the findings were related to documentation and have nomaterial impact. As shown, almost all findings have been resolved and incorporated into theupdated model and/or documentation. Resolution of the few open peer review findings isexpected to have, at most, a minor impact on the model and its quantitative results and nosignificant impact on the conclusions of this application.In resolving the IP3 peer review findings, several additional internal flooding sources wereidentified as not being addressed in the original internal flooding analysis report. Most of thosesources involved fire protection piping, but they also included auxiliary component coolingwater (ACCW) piping in the fan house and short sections of component cooling water (CCW)P0247130002-4722A-5 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical Adequacypiping in a pipe chase in the foyer outside the charging pump rooms. These additional sourceswere included in the final model used for this application.A.2.5 External EventsAlthough EPRI report 1018243 [A.7] recommends a quantitative assessment of thecontribution of external events (for example, fire and seismic) where a model of sufficientquality exists, it also recognizes that the external events assessment can be taken fromexisting, previously submitted and approved analyses or another alternate method ofassessing an order of magnitude estimate for contribution of the external event to the impactof the changed interval. Since the most current external events models for IP2 and IP3 arethose embodied in the IPEEE, a multiplier was applied to the internal events results based onthe IPEEE, similar to that used in the SAMA analysis [A.8 and A.9]. This is further discussed inSection 5.7 of the risk assessment.A.2.6 SummaryThe IP2 and IP3 PRA technical capability evaluations and the maintenance and updateprocesses described above provide a robust basis for concluding that these PRA models aresuitable for use in the risk-informed process used for this application.P0247130002-4722A-6 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-3 Appendix Al, Section 3.4, "Other IE-A8 Appendix Al, Section 3.4, Document the interviews OPEN No ImpactInitiating Events" states 'Other "Other Initiating Events" states This is a documentation issue. Although This is aplant-specific initiators and event 'Other plant-specific initiators discussions were held with plant personnel, no documentationprecursors were also investigated and event precursors were formal interview form or format was used. This enhancement issue.using an FMEA of plant systems as also Investigated using an remains open as a documentation improvementdiscussed below and this was FMEA of plant systems as item for the next update.reviewed with plant personnel to discussed below and this wasverify expected plant response.' It reviewed with plant personnelis not clear that interviews were to verify expected plantconducted, response.' It is not clear thatinterviews were conducted.1-7 Not met since the frequencies were IE-C5 The SR requires that the IE Weight the initiating event OPEN No significantnot weighted by the fraction of frequencies be weighted by frequency time by the While we agree that the wording in the SR itself impacttime the plant was at power. the plant availability. This has fraction of time the plant indicates that weighting should be done, the The current approachnot been done for IP2 initiating was at power. ASME standard acknowledges that the SR provides a slightlyevents, wording is somewhat unclear and provides a conservative result,detailed note of explanation (Note 1 of the and use of theSR). Entergy believes that using the annual stipulated weightingaverage model, which Note 1 acknowledges approach would haveshould not include the weighting factors, is the no significant impactappropriate baseline model in the absence of an on this application.all modes model. We do agree, as the standardstates, that an all modes model should accountfor the time in each operating state. Entergydoes not have an all modes model at this time.We believe that tying risk values to plantavailability without an all modes model canpotentially provide inappropriate risk insights tonon-PSA personnel. It does not apply any risk toother operating states. Therefore, we believethat at the least, our current model meets theSR, when taken in concert with the associatedNote 1.P0247130002-4722 A-7 Risk Impact Assessment of Extending the Indian Point tLRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTnON ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-8 While the documentation of the SC-Cl The current documentation Provide basis for Resolved No ImpactSuccess Criteria is detaiied with poses a potential problem in parameters, limits, Additional references/basis for parameters, Documentation issuesufficient information to support facilitating PRA applications, setpoints, etc. limits, and setpoints were added to Section -incorporated in finalthe model development, the lack of upgrades, and peer review due 01.3.2, "Level 1 Assumptions" and other project file for thereferences to supporting to the significant amount of pertinent sections of the success criteria analysis model used for thisdocuments for a variety of information included that is notebook, application.assumptions and sections makes not traceable.the review difficult and the abilityto maintain the model based uponplant changes and analysisrevisions very difficult to track andchange.Examples are:1) RCS peak pressure within 120seconds of an ATWS2) The normal relief flow througheach PORV valve is 179,000 lb/hr;the maximum flow is 210,000 lb/hrNote that these are simply a coupleof examples of a more prevalentissue.1-t1 Attachment E summarizes the tE-C4 Attachment E summarizes the Produce a table which Resolved No Impactcalculation of initiating event IE-C5 calculation of initiating event shows the actual Added a table showing a sample calculation to Documentation issuefrequencies but there must be a frequencies but there must be calculations using generic, enhance Appendix At of the update report. The -incorporated in finaltable that shows the actual a table that shows the actual plant-specific, and calculations used to develop the IE frequencies project file for thecalculations using generic, plant- calculations using generic, Bayesian updating are contained in the EXCEL files that are part of model used for thisspecific, and Bayesian updating. It plant-specific, and Bayesian the IP2 model update project files and are application.would be helpful to include this updating. It would be helpful retained for future reviews, updates ortable, to include this table, applications. This issue is only a matter of theextent and the details of the calculationsextracted and made part of the written report.Also note that the methodology used for thesecalculations was discussed in Appendix At,Section 11 and the results were summarized inAttachment E.P0247130002-4722ý_a Risk Impact Assessment of Extending the Indian Point ILRT intervalsAppendi, A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-13 No definition or criteria for the DA-A2 The criteria to establish the Provide documentation Resolved No Impactdefinitions of failure modes, and definitions of SSC boundaries, regarding the failure This is a documentation issue. The Current Documentation issuesuccess criteria were identified in failure modes, and success modes to consider for process satisfies the requirements of this SR. The -incorporated in finalthe review of the Data analysis criteria In a manner consistent evaluation of the data boundaries, failure modes, and success criteria project file for thepackage. with corresponding basic event analysis and the associated considered in the Data Analysis are consistent model used for thisdefinitions in Systems Analysis success criteria. (It Is with those used for each system to match the application.are required per the SR. In noted that Attachment 2 of failure modes, common cause and boundaries ofthis case SSC boundaries were Appendix DO, identifies unavailability events. The data analysis notebookdiscussed and examples many of the issues for discusses this (for example, see Appendix D1,provided. However, there was consideration in relation to sections 1.4 and 3.1 thru 3.3 and 4.1, 4.3 andno similar documentation for this SR.) 4.6) and shows that these are all addressed inthe failure modes and success the updated plant model. App. D1, Attachment Acriteria includes discussions and definitions of componentboundaries related to component failure modesand how this was considered in the data analysis.This is consistent with Appendix E, Table E0.1-3which lists the failure modes and associatedcodes that are used in the model. All modeledbasic events are captured in the fault trees andthe associated model data base with codescorresponding to this table and the Data Analysisis shown to match the failure modes andboundaries of these events. In the associatedSystem Notebook, each fault tree is discussedand the overall system success criteria In themodel are summarized.1-14 Accident sequences that reach and AS-A8 DEFINE the end state of the Rewrite the statement to Resolved No Impactremain in a stable state for 24 accident progression as indicate that the accident The statement referred to in the finding, which Documentation issuehours are assumed to be occurring when either a core sequence is mitigated exists in Section 4 of the main report and in -incorporated in finalsuccessfully mitigated. This can be damage state or a steady state when a stable state without Appendix F1.0, has been revised to read: project file for theinterpreted to mean that the condition has been reached core damage has been model used for thismission time is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reached. The mission time "Accident sequences that reach a stable state application.reaching a stable state. This for this is usually 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and remain in that state for thestatement should indicate that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time after the initiating eventaccident sequence is considered are assumed to be successfully mitigated. It Ismitigated when a stable state assumed that sufficient additional resources existwithout core damage is reached. and sufficient time is available by that time torespond to any additional challenges."Ptla7130002-4722 9~

Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-16 SR is MET, however, three system SY-B8 Walkdowns were documented Provide conclusion of Resolved No Impactpackages in which the section as required for this SR. walkdown in all systems The walhdown records for the systems noted in Documentation issuerelating to spatial dependencies However, this Is a packages. the finding (Control Building HVAC, Primary -incorporated in fSnalhad no conclusion as to whether a documentation issue. Water and AFW Building Ventilation systems) project file for thespatial dependency exists (e.g. have been reviewed and no spatial dependencies model used for thisControl Building HVAC, Primary have been Identified. The conclusion has been application.Water, AFWP Building Ventilation) added to each of those system notebooks underSection 1.5 "LOCATION AND SPATIALDEPENDENCIES". The remaining systemnotebooks already contain this conclusion.1-18 Not Met CC II/III due to the lack of DA-D4 A review of the Update Evaluate the posterior data Resolved No Impactdiscussion and documentation Spreadsheet in support of the in relation to the Revised App. Dt and Data Analysis spreadsheet No change wasrelating to examination of Bayesian analysis reflects a uncertainty bounds of the to follow the same approach used for IP3 and required to theinconsistencies between the prior single failure in which the posterior and prior clarify that the requirement in SR DA-D4 to posterior data set.distribution and the plant-specific posterior mean fell outside the uncertainties to address "check that the posterior distribution isevidence to confirm that they are uncertainty bound of the prior discrepancies and reasonable given the relative weight of evidenceappropriate distribution. document the issue such provided by the prior and the plant-specific data"that the discrepancies (if was performed. The discrepancies between thethey exist) can be generic and the updated means were identifiedexplained or resolved, and evaluated and all were found to bereasonable based on the nature of the Bayesianupdate algorithm, the number of failures and theavailable plant data. Appendix D1, Section 3.6was revised to discuss the approach. Thesestatistical tests satisfy the requirements of DA-D4.1-19 There is no evidence that HR-C2 INCLUDE those modes of Analyze miscalihbration of Resolved No Impactmiscalibration of equipment that unavailability that, following equipment that provided Comment incorporated. Additional pre-initiator Change incorporatedprovided initiation signals for completion of each unscreened initiation signals for hunman failure events (HFEs) were added to the in model used for thisstandby pumps were analyzed. activity, result from failure to standby pumps. model to represent miscalibration errors. See application.restore (b) initiation signal or SAS system notebook, Table 1.2 Pre-tnitiatorSection Ht.0 states: 'This review set point for equipment start- Human Failure Events (HFEs) Screening.did not identify any Human Failure up or realignmentEvents (HFEs) that are not alreadyaccounted for as possible failuremodes in the Human Reliabilityanalysis (HRA).'P0247130002-47122utA-10 Risk Impact AsssseSn et of Extending tire Indian P01W ILRT IntervalsAppendhix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-20 A review of the CCF in the System SY-B4 Naming convention should Correct the naming Resolved No ImpactWork Packages (i.e. AFW) reveals match in all references. This convention in the System The common cause basic event names in the Documentation issuethat the Common Cause names issue does not affect results Packages to match the AFW System Work Packages have been corrected -incorporated in finallisted do not match the common since the model names and model. and now match the basic event names used in project file for thecause names in the model and data the data analysis names are the AFW system fault tree model and data model used for thisanalysis package. consistent. analysis, application.(Example: FW406, FW-CCFS-AFWPM, etc.)1-23 In the Scope of Analysis it is IFSO-A4 For each potential source of Include maintenance Resolved No Impactstated: 'In this analysis, all causes flooding, IDENTIFY the induced flooding in the A search of the IP2 condition reporting system No changes to theof flooding were considered except flooding mechanisms that flood initiator frequencies was performed for a period of 15 years for the flooding frequencyplant-specific maintenance would result in a release. Internal Flooding Analysis. No significant Internal values were required.activities-the contribution of INCLUDE: .flooding events (including maintenance Induced),normal maintenance to flooding is (a) Failure modes of were identified which would significantly alter theincluded in the rupture frequency components such as pipes, generic data.data used.' The flood frequencies in tanks, gaskets, enpansionthe EPRI flood guideline do not joints, fittings, seals, etc.include maintenance. (b) Human-inducedmechanisms that could lead tooverfilling tanks, diversion offlow-through openings\created to performmaintenance; inadvertentactuation of fire-suppressionsystem0c) Other eventsresulting In a release into theflood areaFnlu7t3nnIl.a722 u-tiP0247130002-4722A-11 Risk Impnact Assessment of Extending the Inidan Point ILRT Inte-Is~Appendix A P5.A 7ecthsaI AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-24 IDENTIFY the characteristic of IFSO-A5 There is no documentation Identify the pressure and Open No Impactrelease and the capacity of the IFSO-A6 that identifies the pressure temperature of the source. This is a documentation issue. While Appendix C This is asource. INCLUDE the pressure and and temperature of the source, does not specifically identify the pressure and documentation issue.temperature of the source. temperature of the sources, the analysis did The description indocument that the maximum flow rate resulting Appendix C will befrom a guillotine rupture was determined as well enhanced during theas lesser calculated release rates. A range of next update.release sizes consistent with the available EPRIpipe rupture frequency data were, in fact,considered and a flow rate and frequency ofoccurrence derived for each. By this means, thesize and frequency of possible releases werematched as required for the quantitativedetermination of the consequences of internalflooding. This remains an open finding, pendingenhancement of the documentation regarding thepressures and temperatures of the rupturedsystems to meet the letter of the SR.P0247130002-4722A-12 Risk Impact Assessment of E'tending the Indian Point ILRT interoaisAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-26 Capability categories met. Latest DA-Ct It would be helpful to indicate Provide documentation Resolved No significantversions of recognized generic data instances in which the generic regarding the failure Appendix D1 was revised to clarify that any impactsources were used. Generic data data and the model do not modes to consider for mismatches are due to discrepancies in the At most, this mayfor unavailability were not used. match. As currently evaluation of the data generic data sources. Added the following result in a slightdocumented, it is not clear analysis and the associated wording to section 1.4 to address boundaries and conservatism asNote: The analysts ensured, to the how often this occurs or how success criteria. (It is other Issues; "Consistent with System Analysis noted in theextent possible, that the parameter significant mismatches of this noted that Attachment 2 of requirements, the failure rates, common cause disposition.definitions and component type might be. Note: the EDG Appendix DO, identifies failure events and unavailability events wereboundaries were consistent load output breakers are many of the issues for identified from the system fault trees to bebetween the model and the data identified specifically in the consideration in relation to consistent with corresponding systems analysissource. Appendix D notes that text as being one area of this SR.) definitions, success criteria and boundaries (tomismatches may be present, but mismatch. If this is the only the extent practical considering the differences inthat any such Instances would be instance, then this should be the boundary definitions in the generic andconservative because the generic clarified. common cause databases). Component failuredata would include subcomponents data was matched to corresponding events inthat are treated separately in the system fault trees. Failure modes that are in themodel. system models were mapped to correspondingbasic event Type Codes and other events used inNote: The opening paragraph In CAFTA (common cause failure and maintenanceAttachment 0 indicates: 'The unavailability events)." Also revised Attachmentboundary definitions used in the A, section 1.0, item 2 to add; "Note that themodel may need to be modified boundaries provided below are consistent withdepending on the generic database those used in NUREG/CR-6928, however they areand should be clearly defined so not defined in the same manner or to the samethat the failure modes in the model level of detail as they are in the NRC CCFmatch those in the generic database which may result in overlaps in thedatabases.' Apparently, this was boundaries that could lead to conservativenot done in all cases -as noted estimates for the CCF failures". No additionalabove. documentation or evaluation of the data analysisis required to satisfy this requirement.roianisnoti.t,22 u-tIP0247130002-4722A-13 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-27 Met for CCI but not CCII; Section DA-C13 Appendix D1, section 3.7 says Document the interviews Resolved No Impact3.7 System Unavailability Due to 'If no Maintenance Rule or used to meet this As demonstrated in the EXCEL file used for the This was aTesting and Maintenance discusses plant records were available requirement data update, the population of components for documentation issuethat 5 years of unavailability data for a particular component, which Maintenance Rule (MR) unavailability data since there were nowas collected via the Maintenance generic data from NUREG/CR- did not exist was limited to the Appendix R Diesel additional insightsRule program. If no Maintenance 6928 were used to estimate Generator and a few MR non-risk significant available from plantRule or plant records were unavailability.' systems. The Appendix R diesel has only been in personnel.available for a particular service a limited time and the System Engineercomponent, generic data from confirmed that there were no unavailable hoursNUREG/CR-6928 were used to that could be applied for the update. Theestimate unavailability. Maintenance Rule Coordinator and/or theappropriate System Engineers were queriedregarding the other systems for which MRavailability was not monitored but were unable toprovide reliable estimates due to the lack ofmonitoring data. As a result, generic data wasapplied to these system components.Since the discussions with plant personnel did notyield useful information and could not be used "togenerate estimates" for unavailability, additionaldocumentation of those discussions would be oflittle additional value and was not generated.2-2 .Capability Category I met. DA-C1O Discussion in Appendix D was Add discussion to further Resolved No ImpactDocumentation in Appendix Dl was not explicit enough to know explain whether this SR Appendix Dl, Section 3.4 was enhanced to clarify Documentation issuenot sufficient to determine if It was whether Cat II was met. was met at Cat I1. that failure modes were not decomposed into -incorporated in finalnecessary to decompose sub-elements. Therefore, Appendix D does not project file for thesurveillance test data Into sub- address decomposition of failure modes and it model used for thiselements and whether this was was not necessary to perform additional reviews application.done. of surveillance tests to address sub-elementspecific data.ro 247 1 3000 2.47 22 u-anP02,17130002-4722A-14 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION3-2 Each System Notebook contains SY-B14 TAKE CREDIT for system or Provide analysis that the Resolved No ImpactTable B-2a Supporting SY-A22 component operability only if equipment can function The model only takes credit for component Documentation issueRequirements for HLR-SY-A that an analysis exists to beyond design basis operability based on design or rated capability -incorporated in finalstates under SY-A20 something demonstrate that rated or environment. and does not assume or take credit for operation project file for the-such as this for CCW: 'The design capabilities are not beyond design basis capability unless specific model used for thisComponent Cooling Water System exceeded. calculations and evaluations were available, as application.by its design function removes heat noted in the system notebooks for AFV, CBfrom containment. Therefore, the HVAC, EDGV. Clarification was provided in theComponent Cooling Water System system notebooks, as required, to revisedis fully capable of providing heat wording of "Harsh Environments" under sectionremoval. Therefore, no further t.S and in Table B-2a for how SY-A20 is met (seeanalysis is required to support this the other various system notebooks includingfunction.' CCW, CVCS, HHSI, LHSI, IAS, EDG, SWS).However it is not clear thatanalyses were done to take creditfor equipment associated withrecirc inside containment.3-4 There Is no problem with the DA-D1 Issue centers on the Calculate realistic Resolved No Impactgeneric data or the Bayesian calculation of 'realistic parameter estimates using Revised failure identification to include plant No changes to theupdating process used. The issue parameter estimates' using plant specific data. failures not included in EPIX data as explained in data analysis wereis the calculation of 'realistic plant specific data since only revised Appendix D1, Section 3.5. Entergy fleet required.parameter estimates' using plant EPIX / Maintenance Rule procedures and fleet standards address EPIXspecific data since only EPIX / information was used. reporting and confirm that all Maintenance RuleMaintenance Rule information was (MR) functional failures require an EPIX report.used. They also require all failures of high criticalcomponents to be included in EPIX reporting,which includes failures that may cause a trip orimpact plant operation, even of non-risksignificant operating systems within MR scopethat might be monitored under plant criteria andmight not otherwise be captured. Theserequirements ensure that failures of all modeledcomponents are captured in the EPIX data usedfor the PSA model. The only exceptions arefailures of high critical components that occurredprior to 2007, when these procedures wereimplemented. Those failures were obtained fromspecific plant records and included in the update.No further action is required to satisfy thisrequirement.P024713tOD-4722 -tA-15 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION4-1 Met CC It/ItI Based upon a DA-D4 A review of the Update Evaluate the posterior data Resolved No Impactthorough analysis of the generic Spreadsheet in support of the in relation to the The associated data analysis spreadsheet was No change wasdata using plant specific data for Bayesian analysis reflects a uncertainty bounds of the ievised to allow the discrepancies between the required to theBayesian updating. However, there single failure in which the posterior and prior generic and the updated means to be identified posterior data set.is a lack of discussion and posterior mean fell outside the uncertainties to address and evaluated and all were found to bedocumentation relating to uncertainty bound of the prior discrepancies and reasonable based on the nature of the Bayesianexamination of inconsistencies distribution, document the Issue such update algorithm, the number of failures and thebetween the prior distribution and that the discrepancies can available plant data. Appendix D1, Section 3.6the plant-specific evidence to be explained or resolved. revised to clarify that the requirement in SR DA-confirm these inconsistencies are D4 to "check that the posterior distribution isappropriate reasonable given the relative weight of evidenceprovided by the prior and the plant-specific data"was performed. These statistical tests satisfy therequirements of DA-D4.4-2 This SR is Not MET. The use of DA-D1 It is not apparent that all plant Perform a more extensive Resolved No ImpactEPIX as the basis for plant related DA-D4 specific failures associated review of the plant specific See disposition for finding 3-4. No changes to thefailures associated with PRA with PRA related components failures to ensure that the data analysis weremodeled components is insufficient have been captured in the data is complete. (Note: required.to ensure that all failures are data review for this model should it be determinedcaptured. EPIX captures update. that the Indian Point EPIXMaintenance Rule Functional database does actuallyFailures and Critical component Include all PRA modeledfailures (post 2007). Therefore, component failures, thisthis database is limited in scope. FAO can be dispositionedas such).Also it should be considered thatthe Maintenance Rule will notcapture all failures associated withnon-risk significant systems.Therefore, this data is also notinclusive.4-3 Documentation of the data analysis DA-El Supporting files were provided Incorporate the Resolved No ImpactIs not complete due to the lack of during the review that spreadsheet into the Revised Appendix D1, Section 3.6 to include Documentation issueany reference to the basis for the contained critical information document or as a reference reference to the applicable spreadsheets along -incorporated in finaldata results. It was noted during relating to the data analysis. in order to ensure with discussion of how they are the basis for the project file for thethe review that the data analysis is This Information in the form of traceability of the analysis results. The spreadsheets are also retained in the model used for thisactually calculated using an Excel Spreadsheet is not and inputs for the analysis. project files that are maintained available for PRA application.spreadsheets; however, those Included in the Data Analysis Also include guidance on applications, upgrades, and future reviews. Anspreadsheets are not part of the package and is not referenced the use of the information example of the calculations in the Excel fies wasdata analysis package. by the package. contained In within the added to Appendix D. No further action isspreadsheet, required to satisfy the requirements of this SR.P0247130002-4722A-16 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendi.x A PRA Technical AdequacyTABLE A.2-1SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP2 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION4-4 The model uses a single value for QU-B1O The modularnzation of RPS in Incorporate the RPS fault Resolved No ImpactRPS in relation to the ATWS tree the ATWS logic precludes the tree system into the ATWS The RPS is a somewhat unique system, and while Use of a single valueand certain initiating Events. This ability for risk significant logic in a manner that we agree that the modeling of RPS Is not fully for RPS unavailabilityRPS module for the ATWS logic is determinations of support allows results consistent with this SR, we disagree that this has no Impact on thisquantified using the RPS fault tree. systems and components interpretation of individual finding warrants the SR not being met. In application.Although modularization of within RPS. events, particular, the RPS is a fail-safe system. As such,initiating events allows for the loss of a support system does not materiallydetermination of risk significance of impact the reliability of the RPS. Although thethe Initiator, the use of this module shunt trip function does rely on 125V dc power,restricts the usability of the model the increase In unreliability of the RPS associatedfor risk significance determination with unavailability of dc power is negligible. Infor those components associated addition, regarding the modeling of transmitterswith RPS. and trip relays, it should be noted that the RPSfault tree, which is consistent with NUREG/CR-5500 (Volume 2), Is conservative in that it onlycredits two trip signals (overpower delta T andpressurizer high pressure). tndividual testsimpacting the RPS are addressed for onlinemaintenance by adjusting the top event for RPSunreliahility accordingly. Furthermore, thelimited applicability of the Finding should notpreclude the SR from being met.PtlC7t3000Z.47Z2 u-tnP0247130002-4722A 17 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC, BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-11 Appendix C1 of IP-RPT-I0-00023, IFSN-BI Analysis details available to Provide required Resolved No impactRev. A provides a high to medium IFSN-Bi2 the peer review team such as documentation The backup spreadsheets have been obtained as The backup requiredlevel summary of the flood flooding calculations, were not well as the software used for flood level to support futurescenarios, and provides greater sufficient to support upgrades calculations, instructions for use of this software model updates anddepth in some areas. Analysis and would have to be obtained and material that supports its application. This applications is now Indetails available to the peer review or reproduced for future model additional documentation was included in the the projectteam such as flooding calculations, changes. The documentation final model documentation package. Initiator documentation.were not sufficient to support also lacks in reference to specific flag files are contained in the electronicupgrades and would have to be quantification input files Included in the model update documentationobtained or reproduced for future documentation (initiator package. A list of flag files was also added to themodel changes. The documentation specific flag files) internal flooding notebook.also lacks in reference toquantification input documentation(initiator specific flag files)1-12 The walkdown notes in Appendix A tFSN-A5 There is no specific physical For SSCs susceptible to Resolved No impactof IP-RPT-10-00023, Rev. 0, location information found in spray failure (also see FAO Additional discussion was added to the walkdown AdditionalAppendix C.A note the general the documentation for SSCs 2-3), ensure sufficient Appendix to support the spray impacts included information has beenlocation of each SSC with respect to other than flood area and relational location in the model. This includes reference to included in theIts room and elevation as well as its elevation. Therefore, it cannot information between the environmental qualification documents where updated modelsubmergence height. Some be determined which SSCs in target SSC and spray these were used as a basis for stating that documentation.additional general locational any area are susceptible to sources are provided so equipment would not be vulnerable to sprayinformation is sometimes identified spray from any specific spray that a determination can be damage. A conservative separation criterion ofin Section 4.2 of IP-RPT-10-h0f23, source. In the scenario made as to whether the 30 feet was used in examining the potential forRev. 0, Appendix C.t. For example, development it identifies SSCs can be damaged by spray impacts in the analysis. The compositeit may state that a flood source may which equipment is impacted each potential spray piping and general arrangement drawings wereimpact one but not both trains of by spray, but it cannot be source, scrutinized to ascertain whether equipment couldequipment; specifics are not given determined how that be sprayed should a line or other piece ofas to why both cannot be impacted information was obtained or if equipment rupture. The text of the report has(e.g., shielding, curbs, etc.), but the It is correct, been changed to note this. Providing additionalinformation implies the impact of specific location information within the modelspatial information. documentation will be considered to supportfuture updates but is considered a documentationThere is no specific physical location enhancement issue with no expected impact onInformation related to spray type the analysis.failures found in the documentation.SSCs are only identified locationallyby their flood area and elevation. Itcannot be determined which SSCsin any area are susceptible to sprayfrom any specific spray source.P00247 13000 2.47 22A-18 Risk Impact Assessment of Estending the Indian Point ILRT IntervalsAppendix A PRA Techncal AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION1-15 The initiating event frequencies are IE-C5 The initiating event Include the plant Open No significantnot weighted by the fraction of time frequencies are not weighted availability factor in the While we agree that the wording in the SR itself impactthe plant is at power, by the fraction of time the calculation of initiating indicates that weighting should be done, the The current approachplant is at power. event frequencies. ASME standard acknowledges that the SR provides, at most, aSection 10.9 of Appendix AO wording is somewhat unclear and provides a slightly conservativeprovides guidance to account for detailed note of explanation (Note 1 of the result in comparisonplant availability in initiating event SR). Entergy believes that using the annual to use of thecalculations. Section 4.0 of average model, which Note 1 acknowledges stipulated weightingAppendix At states that the should not include the weighting factors, is the approach and wouldavailability factor for the data appropriate baseline model in the absence of an have no significantupdate period was calculated, all modes model. We do agree, as the standard impact on thisHowever, the calculated value is not states, that an all modes model should account application.incorporated into the initiating event for the time in each operating state. Entergyor final CDF results, does not have an all modes model at this time.We believe that tying risk values to plantavailability without an all modes model canpotentially provide inappropriate risk insights tonon-PSA personnel. It does not apply any risk toother operating states. Therefore, we believethat at the least, our current model meets theSR, when taken in concert with the associatedNote 1.3-7 The effects of the flood on PSFs IFQU-A6 Limited flooding-related Discuss flood effects on Resolved No impactwere not specifically addressed in human actions are included in PSFs and make No short term isolation actions were credited in As discussed in thethe HRA analysis. the HRA discussion in adjustments to the HRA the flooding analysis. The only significant field disposition, the onlyAppendix H, but there is no analysis if needed, action credited in the internal events model that potential for amention of any effects of the could be impacted by the plant conditions flooding impact onflood on PSFs. associated with flooding was alignment of the modeledalternate cooling to the charging pumps on loss operator actions hasof CCW for certain specific CCW failure locations, been addressed inThe model has been updated to address that the updated modelconcern, and assumes that operator action is used for thisprecluded by a break in the location that would application.impact that action.P0207130002-4722A-19 Risk impact Assessment of Eyctending the Indian Point ILRT IntervalsAppendi, A PRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION4-14 Failure modes and success criteria DA-A2 Based on the documents As described in Sections Resolved No significantdefined in Systems Analysis are DA-D6 reviewed and the Issues 5.10 and 6.3.11 of This was a documentation issue. The level of impactconsistent with the Data Analysis. identified, component Appendix DO, assure modeling in the IP3 update required use of As noted, anyThis SR also asks for establishing boundaries are not consistent component boundaries various databases since not all databases differences Inconsistent SSC boundaries between among failure rate, CCF and defined in failure rate and provided data for the components included in the boundary definitionsthe system level analysis and the unavailability data. Plant- CCF data match the PSA model. tn some cases, the databases do not have would at most resultdata analysis, specific features need to be model. Assure the sufficient information to clearly delineate the in a very minorReviewed Appendix E6 and E27 of considered for boundary boundaries used In the test applicable boundaries. The system models and conservatism andthe systems notebooks and definitions. and maintenance data is generic databases were reviewed to confirm that would have noAppendix D for the Data Analysis. It is possible to ensure that consistent with the PSA either there was agreement between the model significant impact onBelow is a list of issues identified: the inconsistent boundary model. Make adjustments and generic database boundaries, or component this application.1. System notebooks do not define definitions result in or provide justification for boundaries In the current model conservativelythe component boundaries. The conservative results, but any mismatch identified. overlap the boundaries shown in the genericcomponent boundaries are defined realistic rather than Review plant-specific CCF databases used for the update. The failure ratesby the generic failure rate data conservative results is Ideal. experience for consistency for these additional components were found to besource with limited discussions on CCF events tend to dominate to meet SY DA-D6 small and inclusion in the model results in, atplant-specific SSC features and system level cutsets and requirements, most, a very minor conservatism in the results.modeling considerations, conservative CCF basic event The model documentation was enhanced to2. The guidance document Appendix values may mask other provide additional detail to clarify the issues withDO Section 5. ce states 'Assure the important components in a the generic database boundaries and the slightlycomponent boundaries established system. conservative modeling approach.in the generic data match thosedefined in the PSA model. Make Regarding the example given of the batteryadjustments or justify differences', chargers, the input and output breakers areAlso, Attachment 4, Section 3.0 of included In the generic database boundarythe same document states that CCF definition for common cause failures whereas theboundaries are dictated by the fault input breakers are not clearly identified to betree modeling. However, the included In the generic independent failure rate.component boundaries defined for The PSA model does not include common causefailure rate and CCF data do not failure of the input or output breakers. Thematch. The justification for using model does conservatively include independentthe data that way is that it is the failure of the input breakers due to specificconservative to do so. It Is true that modeling considerations. This approach isthis approach is conservative for considered appropriate to satisfy the SREmergency Diesel Generators, but it requirement.may not be conservative for othercases like batteries and batterychargers where CCF of outputbreakers are not modeled.P024713aa02-4722A-20 Risk Impact Assessment of Extending the Indlan Point ILRT IntervalsAppendix A PRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION4-14 (continued) Regarding the test and maintenance boundaries,3. Sections 1.2 and 1.4 of Appendix the tP3 Maintenance Rule Basis documents forDI state that the data analysis each system, which define the functions thepackage is consistent with the system must meet and the interfacing boundariessystem analysis. However, as between systems, were compared to thediscussed in Item number 1 above, maintenance unavailability terms in the updatedsystems analysis only defines the model. The system functions are consistent withsystem boundary and not the the system models. The unavailable hourscomponent boundaries within the monitored under the Maintenance Rule weresystem. assigned to the same major components in the4. Boundaries of the test and model so that the model boundaries agree withmaintenance unavailability events or conservatively overlap the maintenanceare not specifically discussed, but unavailability boundaries.seem to be same as the boundariesfor the failure rates. Data from theMaintenance Rule program is usedfor this case, but It is not clear if thesystem and component boundariesconsidered In this program isconsistent with the PSA modelboundaries. Section 6.3.11 ofAppendix DO discusses this issue,but there Is no evidence that theanalysis done In Appendix Dlconsidered boundaries applies toroutine test and maintenancepractices at IP3.POZO,1301i02-tZ2 u.2P0247130D02-4722A-22 Risk Impact Assessnent of Extending the Indian Point ILRT IntervalsAppendix A PPA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION6-i The justification/statement that the SC-Bi The justification/ statement Perform rigorous Resolved No impactCST inventory is sufficient for AFW SY-B11 that the CST inventory is evaluation/justification of Plant design documentation supports the 24 Documentation issuefor 24 hrs should be enhanced, sufficient for AFW for 24 hrs the CST inventory to mission time for the CST. The Appendix B write- -incorporated inshould be enhanced. IP-RPT- support 24-hour AFW up was revised to reference a June 2004 final project file for10-00023, Rev. 0, Appendix B, operation. Westinghouse calculation in support of IP3 power the model used forSection B1.3.1.3.2 states early uprate project. The results of this calculation this application.that CST inventory is sufficient (along with initial calculation boundaryfor 24 hrm while later reveals conditions) are used to document adequate CSTthat the MAAP analysis shows water inventory supply to support AFW operationinsufficient CST inventory with for secondary-side cooling for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Instatement that alignment to addition, as noted, CST inventory is typicallythe city water supply may be maintained above the minimum inventory level,required. An informal providing additional margin. Final modelcalculation with the minimum documentation was modified to remove theflow requirement in EOP apparent discrepancies.concludes that "it would seemthat there is enough inventoryin the CST to allow the AFWsystem to operate for 24hours". Then in IP-RPT-10-0023, Section Insights statesthat 'As the normal CSTinventory is sufficient tosupply the AFW pumps for the24-hour mission time in thePSA', no credit is taken for thealternate suction path fromcity water supply.P0247130002-4722A-22 Risk Impact Assessment of ES-tendlng the Indian Point ILP7 IntervalsAppendix A PPRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION6-6 Supporting requirement IFSO-A4 is IFSO-A4 This supporting requirement is Identify the flooding Resolved No impactintentionally not met as stated In intentionally not met as mechanisms that would The intent of the statement in the report was to As noted in thetP-RPT-10-00023, Rev. 0, Appendix stated in IP-RPT-10-OOO23, result in a release for each acknowledge that the EPRI data used for the disposition, plantC1, Section 3.3: 'The one Rev. 0, Appendix CI, Section potential source of flooding analysis included all rupture mechanisms that specific conditionsupporting requirement of the ASME 3.3: 'The one supporting to meet the SR. contribute to piping system failures and to note reports werestandard that we have made no requirement of the ASME there are no readily available data that would reviewed forattempt to meet is IF-B2: "for each standard that we have made allow us to distinguish between different release applicable eventspotential source of flooding, identify no attempt to meet is IF-B2: mechanisms. The identification of specific causes involving humanthe mechanisms that would result in "for each potential source of of failure is therefore a documentation issue. The induced floodinga flooding release". In this analysis, flooding, identify the only contributor not included in the EPRI data is events, which wereno distinction was made between mechanisms that would result human induced flooding events. Since no the only events notthe various causes of floods because in a flooding release". In this applicable generic data exists related to human covered by the EPRIthe rupture frequencies used analysis, no distinction was Induced events, plant specific condition reports data. No suchincluded all floods." made between the various were reviewed for applicable events (none were events were foundcauses of floods because the identified) and discussions were held with plant and the frequenciesrupture frequencies used operations personnel. Based on those used remain valid.included all floods." discussions, activities that could challenge The modelsystem integrity such as large scale movements documentation hasof water and plant modifications are typically been modified toperformed during outages and would not specifically discussconstitute significant contributors to flooding risk. both failureNonetheless, the model documentation has been mechanisms and themodified to specifically discuss both failure conclusions of thesemechanisms and the conclusions of these human human inducedinduced failure evaluations. failure evaluations.6-7 As stated in IP-RPT-10-'OO23, Rev. tFSO-A5 As stated in IP-RPT-10-'OO23, Identify the characteristic Resolved No impact0, Appendix C1, Table 3.3.1.1 for Rev. 0, Appendix Cl, Table of release for each source We consider this a documentation issue. While Documentation issueIFSO-A5, maximum flow rate 3.3.1.1 for IFSO-AS, and its identified failure the table mentioned in the finding did state that a -incorporated Inresulting from a guillotine rupture is maximum flow rate resulting mechanism. maximum flow rate resulting from a guillotine final project file fordetermined and used, instead of from a guillotine rupture is rupture was determined, it also noted that the the model used foridentifying the characteristic of determined and used, instead frequency of this and lesser releases were this application.release for different failure of identifying the characteristic calculated. A range of release sizes consistentmechanism, of release for different failure with the available EPRI pipe rupture frequencymechanism. This is in contrary data were, in fact, considered and a flow rate andto the SR. frequency of occurrence derived for each. By thismeans, the size and frequency of possiblereleases were matched as required for thequantitative determination of the consequencesof Internal flooding. The text in the report hasbeen modified to clarify this matter. Additionalinformation regarding the pressures andtemperatures of the ruptured systems has alsoSeen added to the documentation.P0247130002-4722A-23 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Techncal AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION6-8 IP-RPT-1l-00023, Rev. 0, Appendix IFSO-Al IP-RPT-10-00023, Rev. 0, Identify the potential Resolved No impactCt, Section 4.1.3 states that the IFSO-BS Appendix C1, Section 4.1.3 sources of flooding for each All accessible flood areas were included In the Since as noted in thepotential flood sources were IFSO12 states that the potential flood flood area per the plant walkdowns. Appendix A has been revised disposition, all areasidentified by walkdowns and the sources were identified by standard. to include the areas that were previously omitted were, in fact, walkedexamination of drawings, and listed IFSO-A3 walkdowns and the Perform and document from the documentation, including those areas down, this was ain Appendix A, Plant Walkdown. IFSO-A6 examination of drawings, and walkdowns for missed flood mentioned in the finding, documentation issueHowever, Appendix A does not listed in Appendix A, Plant areas. If these areas The statement in the introduction to the and wasprovide adequate information on Walkdown. However, Appendix cannot be walked down for walkdown notes was intended only to Incorporated in finalflood source as (1) some flood areas A does not provide adequate operational or health acknowledge that there might be small bore, field project file for theare not included in the walkdown information on flood source as reasons, other methods of run piping (less than 1 inch diameter) that were model used for thissuch as 3PAB41-1A,43-60A, 46- (1) some flood areas are not obtaining this data (e.g., not shown on system drawings and would not application.73A,55-63A, 3FH72-B, 3FH80-A, included in the walkdown such plant drawings, operator have been confirmed by the waikdown. Suchetc.; (2) Appendix A has stressed as 3PAB41-1A,43-60A, 46- interviews, etc.) should be small bore pipes were not considered to bethat the walkdown notes do NOT 73A,55-63A, 3FH72-B, employed and documented. signifhcant flood sources.provide a definitive listing of all 3FH80-A, etc.; (2) Appendix A Prepare an integrated list ofequipment and lines or other flood has stressed that the the internal flood sources.sources. Also other fluid sources walkdown notes do NOThave not been considered in the provlde a definitive listing ofanalysis. all equipment and lines orother flood sources. Also otherfluid sources have not beenconsidered in the analysis.6-11 IP-RPT-1O-fiOO23, Rev. 0, Appendix IFSO-B1 There is no list of the internal Prepare an integrated list of Resolved No impactC, Section 4.1.3, which is the flood sources in the analysis the internal flood sources. This is documentation issue. A list of internal Documentation issuesection in the main report for flood that may facilitate PRA flooding sources has been developed and was -incorporated insources, just refers Appendix A, applications, upgrades, and included in a new Table 4.2.1.1 in the final final project file forPlant Walkdown for the information. peer review. update report. This table identifies all the the model used forThere is no list of the internal flood It could facilitate applications, flooding sources in each area and identifies this application.sources in the analysis that may update and review if sources adjacent or lower areas through which floodwaterfacilitate PRA applications, were identified in the main might propagate.upgrades, and peer review. report.50247135000-4722A-24 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix 4 PRA Technical AdequacyTABLE A.2-2SUMMARY OF INDUSTRY PEER REVIEW FINDINGS FOR THE IP3 INTERNAL EVENTS PRA MODEL UPDATEFINDING FINDING DESCRIPTION ASSOC. BASIS FOR PEER REVIEW REVIEW TEAM DISPOSITION IMPACT ON ILRTSR FINDING SUGGESTED APPLICATIONRESOLUTION6-12 tP-RPT-1t-01023, Rev. 0, Appendix IFSO-B2 IP-RPT-10-00023, Rev. 0, Provide adequate Resolved No impactC identifies applicable flood sources Appendix C identifies documentation on the Although Section 3.1.2 previously described the Documentation issuein its Appendix A, Plant Walkdown, applicable flood sources in its process used to identify process for identifying flooding sources, -incorporated inwhich is not adequate for process Appendix A, Plant Walkdown, applicable flood sources additional description has been added to that final project file fordocumentation purpose. For which is not adequate for section and an additional table (Table 4.2.1.1) the model used forexample, the walkdown notes process documentation has been added, which provides additional detail this application.stressed that they do NOT provide a purpose. For example, the describing the sources in each flood zone.definitive listing of all equipment walkdown notes stressed that The statement in the introduction to theand lines or other flood sources; they do NOT provide a walkdown notes was intended only tothere is no list of sources to be definitive listing of all acknowles was tere only toexamined. equipment and lines or other achnowledge that there might be small bore, fieldflood sources; there is no list run piping (less than 1 inch diameter) that wereflood sources; there imno, lnot shown on system drawings and would notof sources to be enamined have been confirmed by the walkdown. Suchsmall bore pipes were not considered to besignificant flood sources.P0247130002-4722u-25 Risk Impact Assessment of Extending the Indian Point ILRT IntervalsAppendix A PRA Technical AdequacyA.3 IDENTIFICATION OF KEY ASSUMPTIONSThe methodology employed in this risk assessment followed the NEI guidance. The analysisincluded the incorporation of several sensitivity studies and factored in the potential impactsfrom external events in a bounding fashion. None of the sensitivity studies or boundinganalysis indicated any source of uncertainty or modeling assumption that would have resultedin exceeding the acceptance guidelines. Since the accepted process utilizes a bounding analysisapproach which is mostly driven by that CDF contribution which does not already lead to LERF,there are no identified key assumptions or sources of uncertainty for this application (i.e. thosewhich would change the conclusions from the risk assessment results presented here).A.4 SUMMARYA PRA technical adequacy evaluation was performed consistent with the requirements of RG-1.200, Revision 2. This evaluation combined with the details of the results of this analysisdemonstrates with reasonable assurance that the proposed extension to the ILRT interval forIP2 and IP3 to fifteen years satisfies the risk acceptance guidelines in RG 1.174.A.5 REFERENCES[A.1] Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy ofProbabilistic Risk Assessment Results for Risk Informed Activities, Revision 2, March2009.[A.2] Engineering Report, IP2-RPT-09-00026, Rev.0, "Indian Point Unit 2 ProbabilisticSafety Assessment (PSA)", November 2011.[A.13] Engineering Report, IP3-RPT-10-00023, Rev.0, "Indian Point Unit 3 ProbabilisticSafety Assessment (PSA)", November 2012.[A.4] Entergy Fleet Procedure EN-DC-151, Revision 2, "PSA Maintenance and Update",January 2011.[A.5] PWR Owners Group LTR-RAM-II-09-092, "RG 1.200 PRA Peer Review Against theASME PRA Standard Requirements for the Indian Point 2 Nuclear Power PlantProbabilistic Risk Assessment," March 2010.[A.6] PWR Owners Group LTR-RAM-I-11-055, "RG 1.200 PRA Peer Review Against theASME PRA Standard Requirements for the Indian Point 3 Probabilistic RiskAssessment," October 2011.[A.7] "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals:Revision 2-A of 1009325", EPRI, Palo Alto, CA: 2008. 1018243.[A.8] Entergy Engineering Report, IP-RPT-07-00007, "IP2 Cost Benefit Analysis of SevereAccident Mitigation Alternatives", Revision 0.[A.9] Entergy Engineering Report, IP-RPT-07-00008, "IP3 Cost Benefit Analysis of SevereAccident Mitigation Alternatives", Revision 0.P0247130002-4722A-26