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{{Adams
#REDIRECT [[W3F1-2015-0021, Response to Request for Additional Information Regarding the Request to Permanently Extend the Integrated Leak Rate Test Frequency to 15 Years]]
| number = ML15124A946
| issue date = 05/04/2015
| title = Response to Request for Additional Information Regarding the Request to Permanently Extend the Integrated Leak Rate Test Frequency to 15 Years
| author name = Chisum M R
| author affiliation = Entergy Operations, Inc
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000382
| license number = NPF-038
| contact person =
| case reference number = W3F1-2015-0021
| document report number = TAC MF4727
| document type = Calculation, Letter
| page count = 133
| project =
| stage = Response to RAI
}}
 
=Text=
{{#Wiki_filter:10 CFR 50.90W3F1-2015-0021May 4, 2015U.S. Nuclear Regulatory CommissionAttn:  Document Control Desk11555 Rockville PikeRockville, MD 20852
 
==Subject:==
Waterford Steam Electric Station, Unit 3 Response to Request for AdditionalInformation Regarding the Request to Permanently Extend the Integrated LeakRate Test Frequency to 15 YearsWaterford Steam Electric Station, Unit 3 (Waterford 3)Docket No. 50-382License No. NPF-38
 
==REFERENCES:==
: 1. Entergy Letter W3F1-2014-0052, License Amendment Request toChange Technical Specifications to Extend the Type A Test Frequency to15 Years, dated August 28, 2014. (ADAMS Accession No.ML14241A305)2. Letter from NRC, Request for Additional Information Regarding theRequest to Permanently Extend the Integrated Leak Rate Test Frequencyto 15 Years (TAC No. MF4727), dated February 18, 2015.  (ADAMSAccession No. ML15033A422)
 
==Dear Sir or Madam:==
In letter dated August 28, 2014 (Reference 1), Entergy Operations, Inc. (Entergy) submitted alicense amendment request to change the Waterford 3 Technical Specifications to permanentlyextend the Integrated Leak Rate Test (ILRT) frequency to 15 years.In letter dated February 18, 2015 (Reference 2), NRC requested Entergy to provide additionalinformation to support review of the license amendment request to extend the ILRT frequency.This letter provides the response to that request for additional information.This correspondence contains no new commitments.If you have any questions or require additional information, please contact the RegulatoryAssurance Manager, John Jarrell, at 504-739-6685.Entergy Operations, Inc.17265 River RoadKillona, LA 70057-3093Tel 504-739-6660Fax 504-739-6678mchisum@entergy.comMichael R. ChisumSite Vice PresidentWaterford 3 W3F1-2015-0021Page 2I declare under penalty of perjury that the foregoing is true and correct. Executed on May 4,2015.MRC/LEMAttachments: 1. Waterford 3 Response to Request for Additional Information(TAC No.MF4727)2. Internal Events PRA Peer Review - Facts and Observations (Findings Only)3. Calculation, Waterford 3 Evaluation of Risk Significance of an ILRT Extension4. Revised Section 4.5.3 of License Amendment Request W3F1-2015-0021Page 3cc: Mr. Marc L. Dapas, Regional AdministratorU.S. NRC, Region IVRidsRgn4MailCenter@nrc.govU.S. NRC Project Manager for Waterford 3Michael.Orenak@nrc.govU.S. NRC Senior Resident Inspector for Waterford 3Frances.Ramirez@nrc.govChris.Speer@nrc.govLouisiana Department of Environmental QualityOffice of Environmental ComplianceSurveillance DivisionJi.Wiley@LA.gov  toW3F1-2015-0021Waterford 3 Response to Request for Additional Information dated February 18, 2015.(TAC No.MF4727)  to W3F1-2015-0021Page 1 of 23By letter dated August 28, 2014 (Agencywide Documents Access and Management System(ADAMS) Accession No. ML14241A305), Entergy Operations, Inc., submitted a licenseamendment request (LAR) to change the Waterford Steam Electric Station, Unit 3 (WF3)Technical Specification 6.15, "Containment Leakage Rate Testing Program," to allow apermanent extension of the Type A primary containment integrated leak rate test frequencyfrom 10 years to 15 years.The U.S. Nuclear Regulatory Commission staff has reviewed the LAR and the followinginformation is needed to complete the review
.RAI #1Regulatory Issue Summary 2007-06 states that the NRC staff expects that licensees fullyaddress all scope elements with Revision 2 of Regulatory Guide (RG) 1.200, "An Approach forDetermining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," by the end of its implementation period (i.e., one year after the issuance ofRevision 2 of RG 1.200). Revision 2 of RG 1.200 endorses, with exceptions and clarifications,the combined American Society of Mechanical Engineers (ASME)/American Nuclear Society(ANS) PRA standard ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early ReleaseFrequency Probabilistic Risk Assessment for Nuclear Power Plant Applications."Given that the implementation date of RG 1.200, Revision 2, was April 2010, and the LAR wassubmitted in September 2014, identify any gaps between the WF3 internal events PRA modelused in this application and RG 1.200, Revision 2, requirements that are relevant to this LAR.Additionally, address the technical adequacy requirements of RG 1.200, Revision 2, that areapplicable to this LAR, or explain why addressing the requirements would have no impact onthis application.RAI #1 ResponseThe internal events PRA model used in the baseline analysis was the Revision 4 Internal EventsPRA model which is the model that underwent a RG 1.200 Rev. 1 Peer Review. The Revision 5model was not ready for use in this application because the Level 2 portion (damage statesother than LERF) was not complete at the time of the LAR submittal. Since then, the Level 2portion of the Rev. 5 model has been completed and a sensitivity analysis was performed toaddress the impact of using the updated analysis . The results of this sensitivity show that,although some risk increase occurs with the update, all risk metrics still meet the acceptancecriteria for acceptable risk thresholds. Since this is the case, the technical adequacy of theinternal events PRA as it is applicable to this application is based on the Revision 5 model.The Waterford 3 PRA (Revision 4) has undergone a RG 1.200 Rev. 1 Peer Review against theASME PRA Supporting Requirements by a team of knowledgeable industry (vendor and utility)personnel. The review was conducted by the Westinghouse Owners Group in August of 2009.The conclusion of the review was that the Waterford 3 PRA model substantially meets theASME PRA Standard and can be used to support risk-informed applications.The findings and conclusions of this review are contained in LTR-RAM-II-09-039, "RG 1.200PRA Peer Review Against the ASME PRA Standard Requirements for the Waterford SteamElectric Station, Unit 3 Probabilistic Risk Assessment."  The overall conclusion found that the  to W3F1-2015-0021Page 2 of 23Waterford 3 PRA meets the ASME PRA Standard at Capability Category II or better for 81% ofthe applicable Supporting Requirements, with 90% met at Capability Category I or better. Thisreview resulted in ninety-six new Facts and Observations (F&Os), forty-nine "Suggestions",forty-five "Findings" and two "Best Practices". Overall, the Waterford 3 PRA was found tosubstantially meet the ASME PRA Standard at Capability Category II and can be used tosupport risk-informed applications.Since the completion of this Peer Review, Reg. Guide 1.200 was revised to Revision 2 whichendorses the ASME/ANS PRA Standard RA-Sa-2009. Because of this revision, a GapAssessment was performed to determine if the results of the Peer Review would have beenaltered if the later issue of the Reg. Guide were to have been used (PSA-WF3-08-01). Theresult of this Gap Assessment shows that no additional Findings would have been issuedhowever two Suggestion level F&Os could potentially have been considered as Findings.Moreover, since the Peer Review was performed in 2009, the Waterford Internal Events PRAmodel has been updated (to Revision 5) in support of efforts to transition to a risk-informedlicensing basis under NFPA-805. While no changes in methods were associated with thisupdate, most of the open Findings were addressed. Although this model was not used inperformance of the original RI-ILRT, a sensitivity study has been performed to see how theresults presented in this License Amendment Request are sensitive to the updated model(ECS14-010, Rev. 1, "Waterford 3 Evaluation of Risk Significance of an ILRT Extension",contained in Attachment 3 of this letter). As can be seen in Table 1 below, usage of the updatedmodel causes a slight increase in the resulting risk metrics, but the change in LERF is still withinthe Reg. Guide 1.174 guidelines for a "very small" change and the percent change in CCFP isstill below the 1.5% criterion.Table 1: ILRT Extension Risk Changes-Baseline and Revision 5 ModelChanges due to extension from 10 years (current)BaselineW/ Rev. 5 Risk from current (Person-rem/yr)2.01E-022.74E-02% Increase from current0.006%0.007%( Risk / Total Risk) LERF from current (per year)2.35E-083.20E-08 CCFP from current3.53E-033.79E-03 CCFP from current (% Change)0.46%0.57%Changes due to extension from 3 years (baseline) Risk from baseline4.82E-026.57E-02(Person-rem/yr)% Increase from baseline0.014%0.017%( Risk / Total Risk) LERF from baseline5.64E-087.68E-08(per year) CCFP from baseline8.47E-039.08E-03 CCFP from baseline (% Change)1.10%1.39%  to W3F1-2015-0021Page 3 of 23The variations displayed in the above table show the impact that the resolution of the Findingshad on the risk calculation performed for this LAR. As is evident from the table the populationdose, delta LERF, and percentage CCFP change metrics still meet the acceptance criteriadescribed in the LAR. However, since not every Finding was completely addressed in the modelupdate, potential impact to this LAR could exist. Therefore, all Findings are discussed inAttachment 2 to this letter to show their related Supporting Requirement and disposition relatedto this LAR. The two Suggestion level F&Os identified in the Gap Analysis are also included inAttachment 2.It should be noted that the Internal Floods Hazard Group was not included in the originalassessment. In order to address the findings related to IF, it is necessary to gain insights intothe impact this hazard group has on the results. Waterford has an Internal Floods model thatwas peer reviewed along with the Internal Events model and has a total CDF of 2.48E-6;however, it does not assess all Level 2 end states. The risk calculation supporting the LAR wasrevised to include this contribution and is attached as Attachment 3 to this letter. Because thesechanges affected statements made in the LAR, Attachment 4 to this letter includes theapplicable changes to the LAR resulting from this calculation revision. (Since no conclusionsmade in the LAR were changed by including the Internal Floods contribution, only Section 4.5.3required revision.)The results provided in Table 1 above already include the contribution from Internal Floods forboth the baseline and the updated model case. With the inclusion of the Internal Floodcontribution, the Reg. Guide 1.200 guidance is met for inclusion of necessary hazard groups. Allapplicable hazards groups have been addressed by the analysis including Internal Events,Internal Floods, Internal Fires, and Seismic.As described by the dispositions to the F&Os in Attachment 2, most of the peer review findingshave already been addressed in the Rev. 5 model. For those that were not addressed, theimpact to this LAR would not besignificant had the items been addressed in the model. Anadditional sensitivity case was performed to determine the impact to the results by doubling theInternal Floods contribution. The results of this sensitivity case showed that the risk thresholdsfor population dose, delta LERF, and percentage CCFP change were not exceeded. Thisprovides further confidence that addressing open findings would not cause an adverse impacton this application. Resolution of Peer Review gaps is discussed in more detail in the responseto RAI #3.The primary piece of the PRA used to support this application is the Internal Events Model alongwith the contribution from Internal Floods. At-Power operation is the only operational modeneeded for consideration since containment is opened for the majority of shutdown operationsand therefore, leakage would be of minimal concern. Also, the resolution of most of the PeerReview Findings and the respective minimal impact as described in Attachment 2 demonstratesthe technical adequacy of the PRA used in this application.The key assumptions and key sources of model uncertainty as they relate to this applicationoriginate mainly from the Level 2 analysis and from the EPRI methodology used to determinethe risk increase associated with the test frequency extension. A couple specific items should bementioned:  to W3F1-2015-0021Page 4 of 231. One of the key assumptions from the EPRI methodology is that all Class 3b releaseswould be categorized as LERF based on the NEI guidance. Since LERF is used directlyas the risk metric, and because not all leakage related Class 3b releases would beLERF, this is considered as a conservative assumption (ECS14-010, R1).2. Because the Internal Floods analysis did not calculate Level 2 damage states after coredamage, it was assumed that half the CDF contribution would bin to Intact and halfwould bin to LERF. This is consistent with the contribution from the sequence results inthe revised Level 2 analysis (PSA-WF3-01-LE Rev. 1, "WF3 Large Early ReleaseFrequency Model") for transients which is the only initiator used in the Internal Floods analysis.3. Another key assumption and source of model uncertainty is related to the simplifiedLevel 2 analysis used in the Revision 4 PRA model which follows the Westinghouseguidance from WCAP 16341 P, "Simplified Level 2 Modeling Guidelines." This guidanceresults in a larger contribution to LERF because of simplifications. The updated InternalEvents model (Revision 5) did a more detailed analysis to address some of thesesimplifications and the result was a lower LERF contribution, but a higher INTACT whichleads to higher Class 3b leakage LERF contribution. The impact of this assumption isseen in the sensitivity analysis using the Revision 5 model and the slight increase in riskresults because of this higher INTACT contribution.Based on the discussion above, it can be concluded that the PRA model used to support thisapplication is of sufficient quality per the guidance contained in Regulatory Guide 1.200Revision 2. The sensitivity case using the Revision 5 model shows the impact of addressing themajority of the peer review findings while the sensitivity case to increase the Internal Floodingcontribution shows that any findings that remain open would not impact the conclusionsincluded in the original LAR.RAI #2Section 4.5.2 of the LAR states that, "The WF3 Fire PRA (FPRA) model has undergone a Reg.Guide 1.200 Peer Review against Sections 2 and 3 of the ASME PRA Standard."  The ASMEPRA Standard RA-Sa- 2009 contains 10 parts, each with several sections. Clarify whether theabove statement from the LAR refers to Sections 2 and 3 of Part 4, "Requirements for Fire At-Power PRA."  If the Fire PRA has not been peer-reviewed against ASME/ANS RA-Sa-2009,clarify how the fire PRA was determined to be of sufficient quality for this application.RAI #2 ResponseThe correct verbiage in the LAR for statement in question above should have been to refer toSection 4 of the 2009 ASME PRA Standard. The PRA quality of the Internal Events PRA model(which is used as an input into the Fire PRA model) is shown by the Peer Review againstSections 2 and 3 of the ASME PRA Standard. However, the technical elements in Section 4 ofthe ASME Standard cover the full breadth of the Fire PRA. The Waterford Fire PRA has beenpeer-reviewed against Section 4 (Part 4) of ASME/ANS RA-Sa-2009. Specifically, the Fire PRApeer review used the Supporting  to W3F1-2015-0021Page 5 of 23Requirements (SRs)in Section 4 of ASME/ANS PRA Standard along with any associated NRCclarifications or qualifications for the individual SRs as contained in Revision 2 to RG 1.200.RAI #3Section 4.5.2 of the LAR states that, "The industry peer review of the updated PRA model hasbeen performed. The updated PRA model meets ASME Capability Category II requirements byaddressing gaps identified by the peer review."  Provide a list of all supporting requirementsfrom the peer-review relevant to this LAR for which the PRA did not meet the ASME/ANS RA-Sa-2009 capability category 1 supporting requirements. Explain why these gaps would notimpact this specific application. For gaps that did not impact another application (e.g., NFPA-805) describe why the finding does not impact this LAR.RAI #3 ResponseAs discussed in RAI Response #1, the original baseline analysis used in this application utilizedthe Waterford 3 Internal Events PRA model (Revision 4) that underwent the Peer Review, andnot the updated model (Revision 5) which resolved most of the Findings from the Peer Review.The Fire PRA model used for the external events analysis did utilize the updated Internal EventsPRA model (Revision 5). However, as also discussed in RAI #1, a sensitivity analysis wasperformed (ECS14-010 Rev. 1, "Waterford 3 Evaluation of Risk Significance of an ILRT") usingthe updated PRA model (Revision 5) and found that, although the risk results showed someincrease, all risk criteria are still met. Since these criteria are met and the Rev. 5 Level 2 modelis now the model of record (it was completed shortly after submittal of this LAR), the remaininggaps that are relevant to this LAR are only those related to the updated model.The Internal Events PRA model underwent a peer review against the ASME/ANS PRA StandardRA-Sb-2005 as clarified by RG 1.200, Rev. 1. Based on the gap analysis done in PSA-WF3-08-01 ("Waterford 3 PRA Peer Review Gap Assessment to the 2009 PRA Standard"), no additionalgaps were found between the Internal Events PRA model and the ASME/ANS PRA StandardRA-Sa-2009 as endorsed by RG 1.200, Rev 2. However, this report did note that two Facts andObservations (F&Os) that were originally given as "Suggestions" would probably be consideredas "Findings" if using Revision 2 to RG 1.200.The Internal Events Peer Review report (LTR-RAM-II-09-039, "RG 1.200 PRA Peer ReviewAgainst the ASME PRA Standard Requirements for the Waterford Steam Electric Station, Unit 3Probabilistic Risk Assessment") lists the assessment of Supporting Requirement CapabilityCategories (CCs). Of all the SRs, 31 did not meet CC-I, while the remaining SRs wereevaluated as meeting CC-I or greater. Each SR that did not meet CC-I has corresponding F&Osrelated to the finding. The updateto the Internal Events model was done to incorporate plantchanges and to address the F&Os given from the Peer Review. Attachment 2 lists the F&Osclassified as Findings as well as the two Suggestion F&Os identified in the Gap Analysis. Adisposition for each of these F&Os is also given to describe either how the finding was resolvedor its applicability to this application. This table excludes the MU (Configuration Control) SRs asthey have no impact on this application.Based on the dispositions of these F&Os, only 9 Internal Events specific F&Os would still beconsidered as not fully addressed and 8 Internal Floods specific F&Os would still be considered  to W3F1-2015-0021Page 6 of 23as not addressed. These F&Os relate to SRs that did not meet CC-I. Based on the dispositionsof the Internal Events F&Os in Attachment 2, only two Findings (AS-A7-01 and SC-B3-01)would have potential non-negligible impact on this application if they were resolved. For theF&Os related to Internal Flooding, based on the dispositions in Attachment 2, resolution of twoof these Findings (IF-B2-01 and IF-D7-01) would have potential non-negligible impact on thisapplication.Due to other conservatisms in the Internal Flooding model such as conservative treatment offlood mitigating operator actions and a bounding duration of flooding release, any potentialimpact because of these findings would be greatly reduced if not negated. Qualitativelyspeaking, more detailed operator recovery action credit itself would provide reduction in CCDPfor many of the flooding scenarios.For the two F&Os related specifically to Internal Events, only two have the potential to impactthis LAR. However, though the potential impact would not be negligible, it would be bounded bythe sensitivity analyses performed. The sensitivity case showing the impact of doubling theinternal floods contribution shows that nearly a 40% additional increase in CDF with respect tothe Internal Events CDF (the CDF in relation to the risk calculation is equivalent to the sum ofLevel 2 Plant Damage states) would not cause the risk thresholds associated with thisapplication to be exceeded. The binning of this contribution was half INTACT and half LATEwhich is more conservative than the SBO (both internal events gaps are related to SBOsequences) contribution to Level 2 damage states of less than 25% INTACT.Therefore, as discussed in the dispositions in Attachment 2, the bounding impact related to theSC-B3-01 Finding would be a 25% increase in CDF while the AS-A7-01 Finding would be muchless than that. So, it can be concluded that the resolution of these Findings would not have anyadverse impact on this application and that additional margin exists for any potential impactfrom Internal Flooding gaps. Also, it should be noted that reference to other applications is notincluded in Attachment 2, and each gap is discussed with respect to this LAR.RAI #4In the LAR, the licensee proposed to revise Section 6.15 of WF3 TS, as follows:A program shall be established to implement the leakage rate testing of thecontainment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordance with theguidelines contained in NEI 94-01, Revision 2-A, "Industry Guideline for ImplementingPerformance-Based Option of 10 CFR 50, Appendix J," dated October, 2008, exceptthat the next Type A test performed after the May 21, 2005 Type A test shall beperformed no later than May 20, 2020.The term "except that" in the above proposed TS wording gives the appearance that theextension of the next Type A test is an exception to the guidelines contained in NEI 94-01,Revision 2A. Provide clarification for the term "except that."  to W3F1-2015-0021Page 7 of 23The NRC staff notes that this was identified for similar applications previously submitted for theNRC review and Entergy had provided clarification in letters dated January 20, 2011, forArkansas Nuclear One, Unit 2, and March 11, 2014, for Arkansas Nuclear One, Unit 1.RAI #4 ResponseEntergy is not requesting any exceptions to the guidelines contained in NEI 94-01, Revision 2-A.The term "except that" in the proposed TS wording of the revision are removed. The proposedrevision to Section 6.15 of W3 TS is as follows:A program shall be established to implement the leakage rate testing of thecontainment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, asmodified by approved exemptions. This program shall be in accordance with theguidelines contained in NEI 94-01, Revision 2-A, "Industry Guideline for ImplementingPerformance-Based Option of 10 CFR 50, Appendix J," dated October, 2008. Thenext Type A test performed after the May 21, 2005 Type A test shall be performed nolater than May 20, 2020.RAI #5Sections 4.0 and 4.3 of the LAR state that the ASME Boiler and Pressure Vessel (BPV) Code,Section XI, Subsection IWL, does not apply to WF3.As described in Section 3.8 of the WF3 final safety analysis report, both the shield building andthe containment vessel are supported on a common reinforced concrete foundation mat. Thecontainment vessel is supported on the concrete fill, which transfers the loads by bearing to thefoundation mat below.Subsection IWL provides the examination requirements for reinforced concrete Class CCcomponents. Considering that the containment vessel is supported on a concrete fill and areinforced concrete foundation mat, provide clarification regarding the LAR's statement ofSubsection IWL not being applicable to WF3.RAI #5 ResponseSubsection IWL provides the examination requirements for reinforced concrete Class CCcomponents. Although the containment vessel is supported on a concrete fill and a reinforcedconcrete mat, it is not part of the containment system.Per ASME Section XI 2001- 2003 Addenda, Subsection IWL-1210 Examination Requirements,"The examination requirements of this Subsection shall apply to concrete containments."Per WF3 FSAR, Section 3.8.1, Concrete Containment, "The Containment System does notutilize a concrete containment. The primary containment is a free standing steel pressurevessel which is surrounded by a reinforced concrete Shield Building. The Shield Building isdesigned as a seismic Category I structure and is discussed under Subsection 3.8.4. The steelcontainment and the Reactor Building internal structure are described in Subsection 3.8.2 and3.8.3, respectively."  to W3F1-2015-0021Page 8 of 23"The Steel Containment Vessel (SCV) is a low leakage rate free standing steel pressure shell,completely enclosed by the concrete shield structure, with an annular space provided betweenthe walls and domes of each structure to permit construction, operations, and in-serviceinspection. The SCV consists of a vertical upright cylinder, all welded steel pressure vessel,with hemispherical top head and an ASME ellipsoidal bottom head. The steel vessel is rigidlysupported on a concrete base that was placed after the cylindrical shell and the ellipsoidalbottom had been constructed and post weld heat treated. The containment vessel, shieldbuilding, reactor auxiliary building, and fuel handling building are supported on a commonfoundation mat. Concrete floor fill was placed above the ellipsoidal shell bottom of the SCVafter the vessel had been post weld heat treated, to anchor the vessel. All components andframing inside the SCV are supported on the concrete floor fill."Per ASME Section XI 2001 - 2003 Addenda, Subsection IWL-1220(b), portions of the concretesurface that are covered by the liner, foundation material, or backfill, or are otherwise obstructedby adjacent structures, components, parts, or appurtenances, are exempt from the examinationrequirements of IWL-2000. Per ASME Section XI 2001 - 2003 Addenda, Subsection IWE-1220(b), embedded or inaccessible portions of containment vessels, parts, and appurtenancesthat met the requirements of the original Construction Code are exempted from the examinationrequirements of IWE-2000. Since the common concrete foundation slab and the bottom steelplate are inaccessible, they are exempt from examination per ASME Section XI 2001 - 2003Addenda, Subsection IWL-1220(b) and IWE-1220(b) respectively.RAI #6Please provide information of instances, during implementation of the WF3 containmentin-service inspection program, where existence of or potential for degraded conditions ininaccessible areas were identified and evaluated based on conditions found in accessible areas,as required by 10 CFR 50.55a(b)(2)(viii)(E) and 10 CFR 50.55a(b)(2)(ix)(A). If there were anyinstances of such conditions, discuss the findings and corrective actions taken to disposition thefindings.RAI #6 ResponseA condition report dated 10/20/2000 documents an instance, during implementation of the WF3containment in-service inspection program, where existence of or potential for degradedconditions in inaccessible areas were identified and evaluated based on conditions found inaccessible areas. The condition description states that:"VT-3 Examinations of the interior moisture barrier (located between the containment vessel andthe concrete floor on the ledge at elevation - 1.5') revealed 22 locations where the moisturebarrier has failed by various mechanisms. The moisture barrier is intended to provide long termcorrosion protection to the containment vessel. No immediate/short term challenges tocontainment integrity were noted during the examinations. The NDE visual examination reportprovides detail on the location and conditions noted. Additionally, the affected areas have beenmarked on the containment vessel. to W3F1-2015-0021Page 9 of 23One of the affected locations is located immediately below penetration #21. This location isbeing wetted by condensation from the CCW pipe. The containment vessel at location #21 isexperiencing general corrosion. The corrosion noted is not sufficient to affect either thestructural integrity or the leak tightness of containment; however, the corrosion does indicate thepotential for degradation below the moisture barrier and requires further investigation.None of the remaining locations exhibited signs of either wetting or corrosion of the containmentvessel."The Responsible Engineer's (RE) Evaluation of Inaccessible Areas was documented in theresponse to the corrective action dated 1/30/2001 and is listed below:Scope:This evaluation covers the evaluations required by CEP-CII-002 paragraph 1.7.3.5 andby 10 CFR 50.55a(b)(2)(ix)(A). Evaluations that are required by CEP-CII-002paragraph 1.7.3.3 are documented in attachment 2 to this corrective action (CA).Results of Evaluation:1) During examination of the moisture barrier two areas were identified which couldindicate the presence of degradation in inaccessible areas.2) Investigation of the first area revealed only limited areas of surface corrosion withno significant wall loss or pitting. All surface areas of the containment vessel at thislocation were determined to be acceptable by examination in accordance with IWE-3122.1.3) Investigation of the second area revealed excessive corrosion in the region belowthe moisture barrier in the annulus. A condition report dated 10/27/2000 wasprepared to document corrective actions associated with this corrosion.Discussion:Paragraph 1.7.3.5 of CEP-CII-002 requires the RE (or designee) to prepare a conditionreport when the RE determines that conditions exist in accessible areas which couldindicate the presence of or result in degradation of inaccessible areas. The purpose ofthis evaluation is to evaluate the acceptability of the inaccessible area in question.Additionally the RE is to prepare inputs to the OAR-1 which include the following:1) A description of the type and estimated extent of degradation, and the conditionsthat led to the degradation;2) An evaluation of each area, and the results of the evaluation, and;3) A description of necessary corrective actions.CA #7 addresses the need for the RE to provide inputs to the OAR-1The flaws identified by the NDE VT-3 reports revealed two areas that indicatedpotential degradation of the containment vessel in the inaccessible areas below themoisture barrier. to W3F1-2015-0021Page 10 of 23One area is located immediately below penetration #21 and has been wetted due tocondensation from the CCW pipe using penetration #21. General corrosion of thecontainment vessel was noted in the vicinity of the moisture barrier in this location.After removal of the moisture barrier, a small area of general corrosion was noted toexist below the moisture barrier at this location. This area of corrosion did not extendbelow the area that could be accessed by removal of the moisture barrier. At thislocation, the corrosion consisted of only a light surface corrosion with no pitting orcracking. Additionally, there was no discernable thinning of the containment vesseldue to the corrosion. As a result the corrosion was determined to be acceptablewithout engineering evaluation (other than the evaluation required due to theindications of degradation in inaccessible areas - the areas subsequently examinedfollowing removal of the moisture barrier). The surface areas were accepted byexamination in accordance with the provisions of IWE-3122.1. After determination thatthe areas were acceptable by examination, the areas of general corrosion werecleaned and the vessel was re-coated. The moisture barrier in this area was replacedon the same MAI. The NDE VT-3 report documents the re-inspection of the moisturebarrier.One area is located almost directly below the maintenance access hatch.Investigation of the area revealed that the corrosion was more extensive than originallyanticipated and condition report dated 10/27/2000 was prepared to document thecorrective actions associated with the corrosion on the containment vessel below themoisture barrier within the annulus region.The Responsible Engineer (RE) provided inputs to the Owner's Activity Report (OAR-1) inresponse to a corrective action dated 11/12/2001. These inputs are provided in Tables 2through 4 below:  to W3F1-2015-0021Page 11 of 23Table 2 - Conditions in accessible areas which indicate the potential for degradation in inaccessible areas (Per 10 CFR 50.55a(b)(2)(ix)(A):Type and Extentof DegradationConditions that led todegradationEvaluationResults of EvaluationNecessary Corrective ActionMechanicalDamage to theinner and outermoisture barrierswith somecorrosion noted in2 locations.Wear and Tear due totraffic and work aroundthe moisture barrier.CR-W3-2000-1275CA 4, Attachment 3.During examination of the moisturebarrier two areas were identified whichcould indicate the presence ofdegradation in inaccessible areas.1) Investigation of the first area, area#13 on NDEN 200-151, revealed onlylimited areas of surface corrosion withno significant wall loss or pitting. Allsurface areas of the containmentvessel at this location were determinedto be acceptable by examination inaccordance with IWE-3122.1.2) Investigation of the second area,area #15 on NDEN-155, revealedmore serious corrosion in the regionbelow the moisture barrier in theannulus. CR-W3-200-1375 wasprepared to document correctiveactions associated with this corrosion.All surface areas examined weredetermined to be acceptable byexamination in accordance with IWE3122.1 following UT measurementsand determination that the corrosionmechanism was not active.1) The inner and outer moisturebarriers were repaired on MAI #
421737.2) QA NDE inspections of theseareas are noted in inspectionreports NDEN 2000-483 andNDEN 2000-484.3) 100% of the moisture barriershall be examined each refuelingoutage until sufficient data isobtained to allow re-evaluationby the RE to determine theoptimum examination schedule.4) Corrosion noted below themoisture barrier on thecontainment vessel within theannulus is considered in CR-W3-2000-1375. Area determined tobe acceptable by examination inaccordance with IWE 3122.1. to W3F1-2015-0021Page 12 of 23Table 3 - Areas with Flaws or Other Relevant Conditions Requiring Evaluation for Continued Service:Examination CategoryItem NumberItem DescriptionFlaw CharacterizationFlaw or Relevant ConditionFound During ScheduledSection XI Examination orTest? (Yes/No)No Areas requiredevaluation for continuedservice.N/AN/AN/AN/ATable 4 - Areas Requiring Repair, Replacement or Corrective Measures for Continued Service:
CodeClassRepair,Replacement orCorrective MeasureItem DescriptionDescription ofWorkFlaw or RelevantCondition FoundDuring ScheduledSection XIExamination or Test?(Yes/No)Date CompletedRepair/Replacement PlanNumber MCRepairMoisture Barrier MB-02Mechanical Damage intwo locations.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-04Mechanical Damage intwo locations.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-05Mechanical Damage inone location.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-06Mechanical Damage intwo locations.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111  to W3F1-2015-0021Page 13 of 23 CodeClassRepair,Replacement orCorrective MeasureItem DescriptionDescription ofWorkFlaw or RelevantCondition FoundDuring ScheduledSection XIExamination or Test?(Yes/No)Date CompletedRepair/Replacement PlanNumber MCRepairMoisture Barrier MB-07Mechanical Damage intwo locations.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-08Mechanical Damage in 6locations.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-09Mechanical Damage inone location.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-10Mechanical Damage in 3locations.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-11Mechanical Damage in 3locations.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-13Mechanical Damage in 2locations that overlap withMB-14.Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111  to W3F1-2015-0021Page 14 of 23 CodeClassRepair,Replacement orCorrective MeasureItem DescriptionDescription ofWorkFlaw or RelevantCondition FoundDuring ScheduledSection XIExamination or Test?(Yes/No)Date CompletedRepair/Replacement PlanNumber MCRepairMoisture Barrier MB-14Mechanical Damage in10 locations. (2 overlapwith MB-13, 3 overlapwith MB-15)Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111 MCRepairMoisture Barrier MB-15Mechanical Damage in14 locations. (3 overlapwith MB-14)Repairmoisturebarrier 1.Yes11/6/00MAI 421737Exempt fromrepair/replacement rules ofIWA 4000 by IWA 4111Note 1: Repair of moisture barriers consisted of removal of damaged areas of the moisture barrier seal  to W3F1-2015-0021Page 15 of 23RAI #7Section 9.2.3.2 of NEI 94-01, Revision 2-A, "Industry Guideline for ImplementingPerformanceBased Option of 10 CFR Part 50, Appendix J," and Condition 2 in Section 4.1 ofthe NRC safety evaluation for NEI 94-01, Revision 2, require supplemental general visualinspections of accessible interior and exterior surfaces of the containment for structuraldeterioration that may affect the containment leak-tight integrity. These inspections must beconducted prior to each Type A test and during at least three other outages before the nextType A test if the interval for the Type A test has been extended to 15 years.Provide a schedule for a typical 15-year interval (between the last Type A test in 2005 and theproposed next Type A test in 2020), in a tabular format, of in-service inspections that were andwill be performed on the containment vessel, and explain how it meets the requirements inSection 9.2.3.2 of NEI 94-01, Revision 2-A, and Condition 2 in Section 4.1 of the NRC safetyevaluation NEI 94-01, Revision 2. Please include the in-service inspection intervals with thestart date and end date of each inspection period, and the corresponding refueling outages.RAI #7 ResponsePreventative maintenance tasks exist to perform periodic general inspections of the accessibleinterior and exterior surfaces of the containment vessel. The table below provides a schedulefor a typical 15-year interval (between the last Type A test in 2005 and proposed next Type Atest in 2020) with the in-service inspection intervals with the start date and end date of eachinspection period and the corresponding refueling outages.Containment Examination ScheduleExamination TypeISI InspectionIntervalISI Inspection PeriodRefuel Outage/ DateILRT Type A Test 2 nd Interval 3 rd PeriodRF13 / 2005IWE ContainmentSurface AreaInspections 2 nd Interval 3 rd PeriodRF13 / 2005IWE Inner/OuterMoisture BarrierInspection 2 nd Interval 3 rd PeriodRF13 / 2005IWE Inner/OuterMoisture BarrierInspection 2 nd Interval 3 rd PeriodRF14 / 2006IWE Inner/OuterMoisture BarrierInspection 2 nd Interval 3 rd PeriodRF15 / 2008IWE Inner/OuterMoisture BarrierInspection 3 rd Interval 1 st  PeriodRF16 / 2009IWE ContainmentSurface AreaInspections 3 rd Interval 1 st  PeriodRF16 / 2009IWE ContainmentBolted Connections 3 rd Interval 1 st  PeriodRF17 / 2011  to W3F1-2015-0021Page 16 of 23IWE ContainmentSurface AreaInspections 3 rd Interval 2 nd PeriodRF18 / 2012-2013IWE Inner/OuterMoisture BarrierInspection 3 rd Interval 2 nd PeriodRF18 / 2012-2013IWE Inner/OuterMoisture BarrierInspection 3 rd Interval 2 nd PeriodRF19 / 2014IWE ContainmentBolted Connections 3 rd Interval 2 nd PeriodRF19 / 2014IWE ContainmentSurface AreaInspections 3 rd Interval 3 rd PeriodRF20 / 2015IWE Inner/OuterMoisture BarrierInspection 3 rd Interval 3 rd PeriodRF20 / 2015IWE Inner/OuterMoisture BarrierInspection 3 rd Interval 3 rd PeriodRF21 / 2017IWE ContainmentBolted Connections 3 rd Interval 3 rd PeriodRF21 / 2017IWE Inner/OuterMoisture BarrierInspection 4 th Interval 1 st PeriodRF23 / 2020IWE ContainmentBolted Connections 4 th Interval 1 st PeriodRF23 / 2020ILRT Type A Test 4 th Interval 1 st PeriodRF23 / 2020RAI #8The LAR states that WF3 has three periods during each 10-year in-service inspection interval.Table 4-2 of the LAR presents the ASME BPV Code, Section XI, Subsection IWE, inspectionresults from 2003 to 2014. Please provide the following:a. The edition of the ASME BPV Code associated with each WF3 in-service inspectioninterval.b. It is not clear from the review of Table 4-2 of the LAR that 100 percent of thecontainment vessel accessible surface areas and the interior and exterior moisturebarriers have been inspected since 2005. Please clarify or supplement the informationin Table 4-2 to demonstrate that the requirements of Table IWE-2500-1 of the ASMEBPV Code have been satisfied.RAI #8a ResponseInitial Interval - Containment ISI Code of Record: ASME BPV Code, Section XI, 1992 Editionwith 1992 Addenda. to W3F1-2015-0021Page 17 of 23Second Interval - Containment ISI Code of Record: ASME BPV Code, Section XI, 1992 Editionwith 1992 Addenda and ASME Section XI, 1998 Edition with 1999 and 2000 Addenda. Wheresubsection IWA is referenced, the 1992 Edition with the 1992 Addenda apply. Those portions ofthe program affected by request CEP-IWE/IWL-001 are developed in accordance with therequirements of ASME Section XI, 1998 Edition with 1999 and 2000 Addenda.Third Interval - Containment ISI Code of Record: ASME BPV Code, Section XI, 2001 Editionwith 2003 Addenda.RAI#8b ResponseThe following supplemental information is added to Table 4-2 to demonstrate that therequirements of Table IWE-2500-1 of the ASME BPV Code have been satisfied.May 2005A general visual inspection of the inside liner plate was performed inaccordance with ASME Section XI Subsection IWE. The examination of theliner plate met the screening criteria or was accepted by the responsibleEngineer. The general visual inspection results reflect compliance with thebuilding structural integrity requirements.All accessible areas of the outer liner plate were examined from the annulusarea. The steel liner plate was inspected in all accessible areas and nodiscrepancies were found.The inner and outer moisture barrier inspections were performed in RF13.Inner moisture barrier sections MB-01 thru MB-12 were inspected. Six (6)areas were found to be unsatisfactory and were repaired and re-inspected withsatisfactory results. Outer moisture barrier sections MB-13 through MB-15were inspected with pitting noted in the NDE visual inspection report. Thecondition was accepted by the Responsible Engineer (RE) since it was a pre-existing condition which was previously identified and evaluated under aprevious condition report dated 10/27/2000 and subsequently rediscovered. Allareas were greater than design except one which was within design allowabletolerances. The areas were repaired and re-inspected with satisfactory results.Fall 2006Eleven (11) bolted connection inspections were performed in RF14 withsatisfactory results. The inner and outer moisture barrier sections MB-01through MB-15 were inspected in RF14. Inner moisture barrier sections MB-01through MB-12 were satisfactory with no reportable damage. Pitting was notedon outer moisture barrier sections MB-13, MB-14, and MB-15 on the NDEvisual examination report. The condition was accepted by the ResponsibleEngineer (RE) since it was a pre-existing condition which was previouslyidentified and evaluated under a previous condition report dated 10/27/2000and subsequently rediscovered.Spring 2008The inner and outer moisture barrier sections MB-01 through MB-15 wereinspected in RF15. All sections were satisfactory with the exception of sectionsMB-02, -03, -05, and -06 which revealed signs of age related degradation andmechanical damage which required repair. The repair was performed and thecondition was captured in condition report dated 5/1/2008. to W3F1-2015-0021Page 18 of 23The inspections performed in May 2005 (RF13), Fall 2006 (RF14), and Spring 2008 (RF15)satisfy the requirements of Table IWE-2500-1 of the ASME BPV Code for the 3 rd period of the 2 nd Interval.November 2009 The inner and outer moisture barrier sections MB-01 through MB-15 as well ascontainment surface area inspections of dome quadrants 1 through 9, plates 1through 162 (with the exception of 71), and the area around the fuel transfertube were performed in RF 16 with satisfactory results.Inside Liner Plate:  In accordance with ASME Section XI Subsection IWE, ageneral visual inspection was performed. The examination of the liner platemet the screening criteria or was accepted by the responsible Engineer. Thevisual inspection was performed in accordance with the program plan andunder the RE's direction. The results of this General Visual inspection reflectcompliance with the building structural integrity requirements.Annulus:  All accessible areas of the outer liner plate and inner shield buildingwere examined from the annulus area, 360° from the -1.50 ft. elevation andaccessible areas from the three permanent ladders located at AZ-310, AZ-196,and AZ-133. The permanent ladder at AZ-310 goes from elevation +20 to thetop of the dome. The steel liner plate was inspected in all areas - nodiscrepancies were found.Note:  A general inspection was performed on the liner plate surfaces requiredby ASME Section XI, Subsection IWE.April 2011Twenty-seven (27) program bolted connections were examined in RF17 withsatisfactory results.The inspections performed in November 2009 (RF16) and April 2011 (RF17) satisfy therequirements of Table IWE-2500-1 of the ASME BPV Code for the 1 st period of the 3 rd Interval.December 2012 Containment Surface area inspections were performed on sections MB-01through MB-15 in RF18 as well as the moisture barrier inside the annulus from0° to 138° azimuth. Results of the liner inspections were satisfactory. As aresult of the steam generator replacement activities, hydroblasting wasperformed and water was found standing on the moisture barrier between the30° and 70° azimuth location. Three 18"x18" moisture barrier sections wereremoved and the liner examined at the 30°, 42°, and 70° locations to assure noactive degradation was present. After replacement of these sections of themoisture barrier, an examination of the repaired moisture barrier areas wereperformed; the examination results were satisfactory.May 2014The inner moisture barrier was inspected in RF19 of items MB-02 through MB-11 with satisfactory results. The outer moisture barrier was inspected in RF18.Twenty-seven (27) program bolted connections were examined in RF19 withsatisfactory results.The inspections performed in December 2012 (RF18) and May 2014 (RF19) satisfy therequirements of Table IWE-2500-1 of the ASME BPV Code for the 2 rd period of the 3 nd Interval. to W3F1-2015-0021Page 19 of 23RAI #9Attachment 4 of the LAR states that Table 4-1 presents summaries of the results from the WF3shield building interior and exterior structural inspections which were performed during eachrefueling shutdown and prior to any integrated leak test. Contrary to this statement, Section 4.3of the LAR states that Table 4-1 presents summaries of the results from the WF3 containmentbuilding interior and exterior structural inspections which were performed every three years andthe shield building inspection was performed prior to any integrated leak test. Also, the datesincluded in Table 4-1 do not appear to support the statement in Attachment 4 that the WF3shield building was inspected during each refueling outage. Please provide clarification.RAI #9 ResponseThe following clarification is provided. The statement in Section 4.3 of the LAR that, Table 4-1presents summaries of the results from the WF3 containment building interior and exteriorstructural inspections which were performed every three years and the shield building inspectionwas performed prior to any integrated leak rate test, is correct. Attachment 4 of the LAR isrevised to reflect the clarification.  "Table 4-1 presents summaries of the results from the WF3containment building interior and exterior structural inspection surveillances. Thesesurveillances were performed every three years and prior to any integrated leak rate test."The following information is added to Table 4-1:September 1995 The following interior and exterior areas of the shield building were inspectedwith no deficiencies noted: shield building roof, exterior shield building walls tothe roof, exterior surfaces in areas of the DCT-A, DCT-B, B Switchgear, +35penetrations rooms, MSIV A, MSIV B, MSIV passage way, -4 RAB wing area,-35 RAB wing area, and +21 RAB. All accessible penetrations, CAP valves,and the top of the containment vessel were inspected inside the annulus withno structural problems observed. Interior inspections were performed onpenetrations from elevations -4, +21, electrical penetrations at +35 and +46,and the containment ring header with no structural deficiencies.March 1999The interior and exterior portions the steel containment vessel was performedin RF9. No indications were noted which would impair the structural integrityof the containment vessel.RAI #10Table 4-2 of the LAR includes the results of the inspection of the containment vessel interiorcoating performed in 2003. Please discuss the highlights of findings from WF3 recentinspections of the containment vessel coating and actions taken to disposition them. to W3F1-2015-0021Page 20 of 23RAI #10 ResponseThe highlights of findings from recent WF3 inspections in RF18 and RF19 of the containmentvessel and actions taken to disposition them are provided below.RF18 Inspections:Recent containment liner plate inspections performed in RF18 were documented in NDE visualexamination reports and are summarized in the table below:ComponentDescriptionResultsDS-05Containment Dome OuterSurfaceOne 10"x8", one 3"x3", and three 1" areasof rust at 96' platform also 4"x12" area ofrust at 85' platform. No pitting or wall lossat any of these areasConstruction HatchSurface area associated withthe construction hatchHatch had been removed and washanging in storage rack at the time of theexamination. Removed for SGRP;Satisfactory; No indications notedDS-01Containment Dome InnerSurface 0°-90° AzSatisfactory; No indications noted.DS-03Containment Dome InnerSurface 180°-270° AzSatisfactory; No indications noted.DS-04Containment Dome InnerSurface 270°-360° AzSatisfactory; No indications noted.DS-02Containment Dome InnerSurface 90°-180° Az.Satisfactory; No indications noted.Maintenance HatchSurface Area Associatedwith the Maintenance HatchSatisfactory; No indications noted.MPAL-SASurface Areas of thePersonnel Airlock (CBMPAL0001)Satisfactory; No indications noted.MPEAL-SASurface Areas of thePersonnel EmergencyEscape AirlockSatisfactory; No indications noted.WS-01Containment Liner InnerSurface 0°-90° Az.@ -4 ElSatisfactory; No indications noted.WS-02Containment Liner InnerSurface 90°-180° Az.@ -4 ElSatisfactory; No indications noted.WS-03Containment Liner InnerSurface 180°-270° Az.@ -4 ElSatisfactory; No indications noted.WS-04Containment Liner InnerSurface 270°-360° Az.@ -4 ElSatisfactory; No indications noted.WS-05Containment Liner InnerSurface 0°-90° Az.@+21 ElSatisfactory; No indications noted.WS-06Containment Liner InnerSurface 90°-180° Az.@+21 ElSatisfactory; No indications noted. to W3F1-2015-0021Page 21 of 23WS-07Containment Liner InnerSurface 180°-270° Az.@+21 ElSatisfactory; No indications noted.WS-08Containment Liner InnerSurface 270°- 360°Az.@+21' ElSatisfactory; No indications noted.WS-09Containment Liner InnerSurface 0°-90° Az.@+46 ElSatisfactory; No indications noted.WS-10Containment Liner InnerSurface 90°-180° Az.@+46' ElSatisfactory; No indications noted.WS-11Containment Liner InnerSurface 180°-270° Az.@+46' ElSatisfactory; No indications noted.WS-12Containment Liner InnerSurface 270°-360° Az.@+46' ElSatisfactory; No indications noted.WS-13Containment Liner InnerSurface 352.8°-138°Az.@+46' ElSatisfactory; No indications noted.WS-14Containment Liner InnerSurface 138°-207° Az.@+46' El.Satisfactory; No indications noted.WS-15Containment Liner InnerSurface 207°-352.8° Az.Active corrosion noted at lug weld tocontainment liner adjacent to Pen.36. Nopitting or wall loss in this area. Corrosionappears to be the result of condensationdripping from chill water line locatedabove this lug.RF19 Inspections:Recent inspections of the containment vessel coatings were performed in May 2014 (RF19).The findings from these inspections are discussed below."Coating failures were only found on the vessel liner plates, dome, and polar crane ring girder.Mechanical damage was observed on all the components, other than the dome. Rusting of thesubstrate was not observed in the areas where damage of the coatings (either from coatingfailure or mechanical damage) was observed. The coating system that was observed failing isCarboline Carbo Zinc 11 (CZ11) primer top coated with Carboline Phenoline 305. The type offailure is splitting of the CZ11 primer, i.e. the primer splits leaving CZ11 on the substrate. Thisis typical failure of this coating system.The areas of failures are shown on the attached plate identification sheets for the liner, dome,and ring girder. A breakdown of the coating failure areas are shown below in the table. to W3F1-2015-0021Page 22 of 23LocationTotal AreaPlates 1 to 36525.118 ft.
2Plates 37 to 162, includingConstruction Hatch,Maintenance Hatch,Personnel Hatch, and Escape Hatch605.504 ft.
2Dome204.9 ft.2Ring Girder97.7 ft.2Total Area: 1,433.22 ft.
2These failures are acceptable based on the allowable failures used for the design of the safetyinjection sump screen per the design calculation for Debris Generation Due to LOCA withinContainment for Resolution of GS-191. According to this calculation, the allowable amount ofcoating failures is as follows.LocationAllowable Failed AreaContainment Vessel Dome3,082 ft.2Containment Liner Between Elevation 112 ft.and 138 ft.1,144 ft.2Total Area: 4,226 ft.
2The design input records document that:
"No extra square footage will be used for failed steel coatings. The amount of failed coatingsalready included for the containment liner and containment dome is conservative
.""Every refueling outage 10% of service level 1 coatings on structural steel are inspected inaccordance with Procedure NOECP-451, Conduct Engineering Inspection of ReactorContainment Building protective coatings and commitment A8350. In addition, 100% of thecontainment liner plates are inspected for failed coatings. These inspections usually do notidentify any failed paint on the structural steel, and limited amounts on the containment linerplates. However, any failed paint is either repaired, or added into the total of already identifiedfailed coatings. Therefore, the total of 4,226 ft 2 is considered a conservative amount."Since the total found failed area of 1,433.22 ft.
2 is less than the allowable failed area, it isacceptable to have 1,433.22 ft.
2 of failed coatings inside containment. A condition report wasinitiated to document the coating failures."  to W3F1-2015-0021Page 23 of 23RAI #11Please discuss NRC Information Notice 2014-07, "Degradation of Leak-Chase ChannelSystems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner,"as it may apply to WF3. If applicable, discuss the operating experience, inspection results, andany corrective actions taken.RAI #11 ResponseNRC Information Notice 2014-07, "Degradation of Leak-Chase Channel Systems for FloorWelds of Metal Containment Shell and Concrete Containment Metallic Liner," was addressed incondition report dated 5/27/2014 and found that WF3's containment liner is not designed withthe channel system described in the Information Notice and there are no additional actionsrequired for WF3. Specifically, "WF3 does not have any components that should be added tothe Containment Inservice Inspection Program equivalent to the items discussed in IN 2014-07.There are no channels installed to encompass the welds in the ellipsoidal bottom head of thesteel containment vessel with associated pressurization lines/tubing/valves. There are noadditional actions to take as part of this Information Notice. This conclusion is based on areview of design basis documents and associated controlled drawings. Additionally, a walk-down was performed in RF18 to specifically look for any covers similar to the ones identifieddue to in this Information Notice.RAI #12Please provide the following information:a. Percent of the total number of Type B tested components that are on 120-monthextended performance-based test interval.b. Percent of the total number of Type C tested components that are on 60-monthextended performance-based test interval.RAI #12a ResponseEighty-five percent (85%) of the total number of Type B tested components are on a 120-monthextended performance-based test interval.RAI #12b ResponseForty-eight percent (48%) of the total number of Type C tested components are on a 60-monthextended performance-based test interval. toW3F1-2015-0021Internal Events PRA Peer Review - Facts and Observations (Findings Only)  to W3F1-2015-0021Page 1 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionAS-A7-01Accident Sequence Modeling Open forInternal EventsMinimalImpact on RI-ILRTBased on review of the WF-3 event trees and toplogic model, accident sequences are notdelineated for all possible scenarios - particularlyin cases where a mitigating function may havesucceeded. Specifically: station blackoutsequences after successful power recovery, andtransient sequences with successful operation ofRCS pressure control. In each of these casesadditional mitigating systems must be questionedto determine that the sequence terminates in asafe state.Partially ADDRESSEDA review of the event trees concluded thattransient sequences with successful RCSpressure control are correctly modeled. Theappropriate systems required following thesuccessful RCS pressure control have beenconfirmed to ensure a safe end state.Station blackout scenarios require success ofEmergency Feedwater for secondary heatremoval. During the most recent revision, creditfor offsite power recovery was removed forscenarios involving hardware failure of all 3 EFWpumps (PSA-WF3-01-QU). Also, many of thecutsets that would be added by additionalmodeling in this area would be non-minimal.Resolution of this F&O however, could have asmall, but potentially noticeable impact on this application. to W3F1-2015-0021Page 2 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionAS-A7-02Modeling of ADVs for SGTROpen forInternal EventsNegligibleImpact on RI-ILRTIn the Accident Sequence Notebook, assumption2.20 reads: ...for SGTRs, failure of ADV to closeafter opening is not included due to block valvesupstream of the ADV that could be closed by theoperator. It is not clear if after not modeling thefailure to ADV to close, if the closure of the blockvalves by the operator has been modeled. If it isnot modeled, the review team believes that itshould be, so as to not lose the dependency thatthis operator action might have on other operator actions.PARTIALLY ADDRESSEDThis finding has been evaluated though it has notbeen explicitly closed out. The WaterfordAccident Sequence analysis (PSA-WF3-01-AS)has been revised. Modeling of the atmosphericdump valves (ADVs) to close post SGTR hasbeen added into the top logic, and theassumptions associated with not modeling thisfailure mode have been removed/revised. Nocredit is taken for an operator action to reclosethe ADV.This finding has no impact on the risk impact forthis LAR since credit for this operator actionwould only serve to reduce impact to CDF, evenwith the consideration of dependencies.AS-B3-01Environmental Effects onContainment EquipmentClosedThe AS report (PRA-W3-01-001S01 Revision 1)includes discussion of the phenomenologicalimpacts of heating of the containment sump water(failure of HPSI recirculation due to loss ofrequired NPSH and pump cavitation) and largecontainment rupture (loss of safety injection dueto the rapid depressurization, flashing of hot waterin the sump, and loss of net positive suction headto the HPSI pumps) that can occur due toinadequate containment heat removal. However,some events such as steamline breaks andfeedwater line breaks can result in harshenvironments (especially steam and hightemperature) where mitigating equipment arelocated.ADDRESSEDWCAP-16679-P - 'Accident SequencePhenomena' was reviewed to determine if anyphenomena other than the SLB and FLB impactwere not addressed in the current Waterford ASanalysis. All other phenomena have beenaddressed in the accident sequence and thesystem analyses. The effects of steam line andfeed line breaks are evaluated in the initiatingevent document. to W3F1-2015-0021Page 3 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition DA-C2-01Use of Condition Reports forData CollectionClosedThe method used for collecting failure dataappears to be valid, but the method for collectingunavailability data is not. If unavailability data isnot tracked directly by a Maintenance RuleFunction, the use of the Condition Report processto identify unavailability is not valid sinceplanned/scheduled maintenance activities and/ortesting procedures that make a pump/systemunavailable will not be tracked in ConditionReports unless something goes wrong during thescheduled activity.
ADDRESSEDAn update to the Data Report has beenperformed (PSA-WF3-01-DA-01). Included in theupdate was a review of Operator logs to identifyunavailability probabilities for those systems nottracked by the System Engineers or included inthe Maintenance rule Database. The updateddata is included in the internal events PRA model.DA-C6-01Demand Based DataAssumptionsClosedAssumptions 8, 10, and 12 violate therequirements for calculating demands based onthe ASME standard. Specifically, Assumption 8does provide a method to ensure that Post-Maintenance demands are excluded fromconsideration - which is a requirement of theASME Standard, and Assumptions 10 and 12count changing fan speeds as demands which isinconsistent with how the fans are modeled andtreated in the PRA.
ADDRESSEDThe internal events PRA data has been updated.The assumptions listed in the finding associatedwith PI data collection are no longer relevant orincluded in the model. The update included areview of amp hours and operator logs to capturemultiple starts that could be due to post-maintenance testing and exclude them.
DA-C7-01DocumentationClosedNo review of surveillance tests or plannedmaintenance activities is documented. Theidentification of these tests and maintenanceactivities, and the estimation of their frequencies(based on TS requirements of "frequency ofperformance" requirements) is an ASME Standardrequirement for DA-C7. Review or estimation ofsurveillance test practices is required forrequirement DA-C9.
ADDRESSEDThe internal events PRA data has been updated.The update utilized both MR data and operatorlogs to collect both plant specific failure andunavailability data (PSA-WF3-01-DA-01). to W3F1-2015-0021Page 4 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition DA-C8-01Modeling of Normally RunningEquipment in StandbyClosedThe model assumes a base, normal alignment.No consideration when base operating SSC isactually in standby, or standby SSC is operating.This may have an adverse impact on supportsand dependencies. Provide rational andscreening as to why the time that components arein their standby (or operating) status are notincluded in the model, or include the standby timein the model.
ADDRESSEDPlant-specific operational records were used todetermine the time that components wereconfigured in standby status. The model hasbeen revised to use conditional probabilities, asappropriate, for systems that have both runningand standby equipment associated with them.
DA-C10-01Plant Specific DataClosedAn assumption in the Data Report (PRA-W3-01-001S05, Rev. 1) notes that surveillance data wasembedded in the PI system and the failure data.Since failure decomposition is not employed in theWF3 PRA model, surveillance tests were notseparately reviewed. The component exposure isaccomplished by considering the possibleopportunities for component operation. The majorsource of raw data on equipment operation isfrom the WF3 PI database. The PI database usesthe information from the plant computer todetermine the start and stop information on agiven piece of equipment. From the start and stopinformation, the duration or the running hours canbe determined for the piece of equipment. Areview of surveillance tests was not performed todetermine whether all of the exposure and failuredata collected was applicable to the componentfailure modes.
ADDRESSEDThe internal events PRA data has been updated.The update includes surveillance data (viaoperator logs) and no longer uses the PI system(PSA-WF3-01-DA-01). The current data effortfully meets the DA-C10 SR requirements. to W3F1-2015-0021Page 5 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition DA-C12-01Collection of UnavailabilityDataClosedAssumption 7 states that system unavailabilitydata was only available on a monthly basis(should be able to refine this based on operatorlogs) and that over-estimation was potential if thesystem outage occurred during the month theplant was in an outage, and that thisoverestimation was assumed to be acceptable.Some Maintenance Rule (MR) data may becollected during lower modes, but should not beincluded in the at-power model data.
ADDRESSEDThe internal events PRA data has been updated(including plant specific unavailability). Theupdate utilized both MR data and operator logs tocollect unavailability data. The current data effortfully meets the DA-C12 SR requirements (PSA-WF3-01-DA-01).Outage time was removed from the update to themaintenance unavailability.
DA-C12-02Plant Specific Data CollectionClosedSection 3.3.6 states that for systems that are nottracked for Unavailability by the SystemEngineers, the unavailability probabilities were notupdated due to a lack of data. This is notacceptable since, although the data is not trackedby the System Engineers, the data does exist.
ADDRESSEDThe internal events PRA data has been updated(including component unavailability). Operatorlogs along with maintenance records werereviewed for years 2002 through 2012 to identifyunavailability of major safety systems andassessed at a train level (PSA-WF3-01-DA-01).Plant specific data was unavailable for a fewspecific components. For these NUREG/CR-6928data was used.HR-A1-01Systematic Review for Pre-InitiatorsClosedPre-initiators are identified in SY notebook.However, there is no related test or maintenanceprocedure listed. There is no evidence to showthat the systematic review of procedures andpractices has been done.
ADDRESSEDWaterford has an extensive number of pre-initiator HRAs modeled. These events cover allstandby systems and trains.The updated Waterford HRA analysis includes asystematic review of procedures and practices inevaluating pre-initiators. All pre-initiators havethe relevant procedures included in thedevelopment documentation (Appendix A of PSA-WF3-01-HR). to W3F1-2015-0021Page 6 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionHR-B1-01Pre-InitiatorsClosedThere is no pre-initiator identified for CCW,because of the CCW is a running system.However, the CCW system may support thesafety related standby system. The path of theCCW to support this system may be failed due topre-initiator HFE.
ADDRESSEDRestoration errors of CCW to a standby systemare included in the restoration of the associatedstandby system (PSA-WF3-01-HR). For instanceCCW to the Containment Spray pumps isincluded in the Containment Spray restorationlogic (YHF3PMPATA and YHF3PMPATB) not theCCW logic. Therefore, these restoration errorshave been identified and evaluated in the current model.HR-D1-01Common MiscalibrationModelingClosedASEP is used for both misalignment andmiscalibration. Status check may take as a creditfor misalignment pre-initiator. However, it is not acredit for miscalibration. The same tool is not afactor to fail the alignment, but it is an importantcommon factor to fail multiple miscalibration.
ADDRESSEDCommon calibration tools were accounted for inthe common miscalibration events in thestaggering of the tests. Therefore, the values areconsidered acceptable. Additionally, a sensitivityanalysis was conducted by increasing all pre-initiator HFEs by a factor of 10 (PRA-W3-01-001S13). This sensitivity showed that theWaterford CDF only increases by 28.37% withthe increase in pre-initiator events of whichmiscalibration events are only a fraction.HR-F2-01Human Failure Event CuesClosedThe cue of each HFE is not clearly addressed.
ADDRESSEDFor the base PRA model, the cues (i.e.,annunciators, EOP/AOP entry conditions) areexplicitly discussed in each operator action in themodel and are documented in the operatorinterview sheets. All of this is clearly documentedin the updated PRA HRA analysis - PRA-WF3-01-HR. to W3F1-2015-0021Page 7 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionHR-G4-01Human Reliability AnalysisTimingClosedAs seen in the HRA spreadsheet (hfe_cp.xls), thetime available to complete actions is based on arange of references including plant-specificcalculations. However, in some cases unjustifiedand/or inaccurate assumptions were used as abasis. The event timelines in the HRAspreadsheets also do not consistently identify thespecific point in time relevant indications arereceived. For example, the success criteria forsump recirculation require the operators to closethe minimum flow valves to the RWSP within onehour (per the HPSI and SC notebooks). Theoperating procedure for recirculation requiresthese valves to be closed within 2 minutes. Nojustification for the one hour timing is providedand this timing is inconsistent with the 1.82 hrused in the HRA. The justification should includeconsideration of the quantity of water that wouldbe diverted to the RWSP during the time frame,the habitability impacts of containment sumpwater being sent to the RWSP resulting in higherradiation levels in the RAB areas that aretraversed by the piping to the RWSP, and theimpact this has on operator recovery actions.
ADDRESSEDThe recent model update included an update tothe event referenced in the finding (PSA-WF3-01-HR). The new time window for the action is 1hour (60 minutes). The new value used in theinternal model reflects this reduced (from 1.8 hrs)time window. Operator interviews indicate thatthe action will take 2 minutes (not that it isrequired in 2 minutes). to W3F1-2015-0021Page 8 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionHR-G6-01Human Reliability AnalysisDocumentationClosedA review of the summary HEP list did not indicateany issues with inconsistencies between theHEPs. There was no documentation that WSEShad performed an internal consistency evaluationof their post-initiator HEPs. A discussion with theanalyst indicates that they did perform an internalconsistency analysis considering the scenariocontext, plant history, procedures, operationalpractices, experience, and the relative difficultiesof the actions and the timing. However, becauseno issues were found, they did not document thereview. WSES does need to document that theyhad performed the review, describe the generalprocess and indicate that no discrepancies hadbeen found. A table of HEP by HEP comparisonsis not needed.
ADDRESSEDSection 4.1.4.1 of the updated HRA notebook(PSA-WF3-01-HR) describes the consistencyreview of the modeled events. to W3F1-2015-0021Page 9 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionHR-H2-01Modeling Non-Proceduralized ActionsClosedAs documented in the HRA report (PRA-W3-01-001S03, Rev. 1), the recovery actions included inthe WF-3 PSA are not explicitly directed byprocedures. Although there is no procedure ortraining for some of these actions, discussionswith operators/TSC have evidently indicated theseactions would likely be pursued (although nodocumentation of these discussions was includedin the operator interview sheets). In fact, theworksheet for one action notes the operators donot have enough training or practice to credit theaction, although it is given and HEP of 0.1. Thereare 9 non-proceduralized operator actionsmodeled. A review of these non-proceduralizedactions shows that:* the time available is short (EHFMANTNR)* the action is not trained or not practiced(EHFMANTNR, MHFSAIABYR)* the working environment is poor(OHFMSSGAGR, QHEFEFWSBOR)* the decision to implement is complicated(OHFMSSGAGR)* the action is complex (QHEFEFWSBOR)All of the above items contribute to there notbeing sufficient justification for their credit in the model.ADDRESSEDThe Waterford PRA does not credit any non-proceduralized actions (actions modeled in thepast are no longer in the model). This isdocumented in the updated HRA report, PSA-WF3-01-HR. to W3F1-2015-0021Page 10 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionIE-A6-01Operator Insights on IEDevelopmentClosedThere is no evidence that interviews of operatorsor engineers were conducted to determine scopeof initiating events or if current list of initiatingevents is correct. MREP information is includedbut it is not directly applicable to cover all of theaspects that need to be considered in the initiatingevent scoping and if they overlooked any initiatingevents.ADDRESSEDWaterford evaluated each plant system withOperations Personnel to determine if a loss of asystem or train would cause a plant scram or not.Since identification of initiators reviewed allgeneric sources and plant specific systems (PSA-WF3-01-IE), additional initiating events types arehighly unlikely to be identified. This finding wasproperly addressed, but insufficientlydocumented.IE-C6-01Initiating Event Fault TreeModelingClosedThe IE report (PRA-W3-01-001S06 Rev. 2)documents IE fault tree modeling for T9 (loss ofCCW), T9RCP (loss of CCW to RCPs), TIA (lossof IA), and TTCW (loss of turbine cooling water).The initiating event fault tree modeling for thesesystems considers multiple failures, CCF eventsand routine system alignments. These IE FTsexclude many failures that are included in thesystems analysis (failures of valves, breakers, etc.in redundant paths to transfer open or transferclosed; component failure rates less than 1% ofthe pump active failures such as sensors andtransmitters; and flow diversion paths).
ADDRESSEDThe current IE fault tree logic is more thoroughthan past models. The current model isdocumented in PSA-WF3-01-IE. The current IEfault trees include items in redundant paths(including valves and breakers). to W3F1-2015-0021Page 11 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionIE-C12-01Interfacing System LOCAOpen forInternal EventsNo Impact onRI-ILRTISLOCA - low pressure LPSI and HPSI linecontain two check valves in series. The failurerate of the check valves need to be treated asconditional, rather than independent.Additionally need to address small ruptures in theLPSI MOVs. At present only large leakage isconsidered.NOT ADDRESSEDThis finding is associated with the inclusion ofState of Knowledge Correlation. The increase inprobabilities due to SOKC would be minor perWCAP-17154-P. Additionally, a review ofNUREG/CR-6928 shows that small ruptures aredefined as 1 to 50 gpm. Leaks of this size arenot considered sufficient to meet the classificationfor ISLOCA.This finding has a negligible impact on theinternal events and would also have negligibleimpact on the risk calculation for ILRT. If anyincreases in ISLOCA sequences occurredbecause of this finding, the impact would benegligible since the Class 3a and 3b EPRIrelease categories used to calculate the riskmetrics are determined by subtracting thecontribution of Class 2 and Class 8 releases(which include ISLOCAs) from the overall CDF. to W3F1-2015-0021Page 12 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionIF-B2-01Internal FloodOpen forInternal EventsMinimalImpact on RI-ILRTAlthough required by this SR, no evaluation ofindividual component failure modes, human-induced mechanisms, or other events that couldrelease water into the area were identified. Theevaluation assumed that using a guillotine rupturewas adequate to not require any specific failuresor human-induced mechanisms. This does notmeet the intent or specifics of this requirement.Other SRs are also potentially not met when only theuse of a guillotine rupture is used. These include:(IF-B3)  Waterford 3 basically characterized allflood sources as catastrophic ruptures but wherethere are potential spray targets they do evaluatespray impacts. Waterford characterizes the floodin terms of gpm for larger sources or as total floodcapacity for smaller flood sources. Waterforddoes consider pressure of the flood source to alimited extent, primarily when evaluating thepotential for spray impacts. However, there is noevidence that Waterford considered thetemperature of the flood source beyond statingthat HELB is treated elsewhere. Waterford shouldinclude some discussion of temperature in PRA-W3-01-002.(IF-D6) Section 2.0 of the Internal Floodinganalysis specifically states "all causes of floodingwere considered except plant-specificmaintenance activities. No mention ofinclusion/exclusion of generic maintenanceactivities was found. While Waterford discussesoperator error contributions to flooding at a veryhigh level in section 3.1.2, basically the onlyfloods considered were catastrophic failures. Theflood scenario frequencies were then quantifiedusing generic pipe rupture data and plant-specificpipe length. The resulting low frequencies oftenlead to scenarios being subsumed. While theoperator induced floods may be less severe, thefrequencies will be higher, so they should beconsidered explicitly.NOT ADDRESSED(IF-B2) Though the intent of the SR is not met,the potential increase is most likely to bebounded by the conservative assumption listed.The frequency of human-induced failures mightbe higher, but the situations in which they couldoccur is limited. Also, operator action to mitigatethese events would be more reliable since themaintenance activity would bring heightenedawareness to the system. However, the exclusionof such scenarios does introduce a potential non-conservatism to the flooding analysis andpotential impact to this application.(IF-B3) The majority of piping in the Waterfordplant that isn't addressed by High Energy LineBreaks are of a lower temperature. This isprimarily a documentation issue which affectsonly a small contribution to the overall results.Resolution of this Finding would have minimalimpact on the results of the Flooding Analysisand would also have minimal impact on this application.(IF-D6) The analysis assumed that usingguillotine rupture would bound any additionalcontribution from human-induced failures. Also,the frequency of human-induced failures might behigher, but the situations in which they couldoccur is limited. The exclusion of such scenariosdoes introduce a potential non-conservatism tothe flooding analysis and potential impact to this application. to W3F1-2015-0021Page 13 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionIF-C3c-01Internal Flood - Lack ofEngineering DocumentationOpen forInternal EventsNo Impact onRI-ILRTThere do not appear to be any Engineeringcalculations available to support some of thestatements or inherent assumptions made in theInternal Flooding Analysis. In particular, roomdimensions and flood rates are not available tojustify flood depths stated for various rooms,some zones credit "air tight" doors as beingstructurally sound up to a depth of 6 inches withno justification of door integrity against a staticwater load of this depth,  "air tight" doors appearto be treated as "flood doors" with no justificationas to how this was determined (normally air tightdoor seals are not designed to prevent waterintrusion or extrusion), timing related calculations(time for flood to reach susceptible equipment,flood rates, etc.) were not included or referenced,etc. If these calculations exist, they should beeither provided in appendices to the report orreferenced in the appropriate sections of thereport.If the calculations do not exist, they should beperformed, and the statements and inherentassumptions in the analysis re-verified to ensurethey reflect the results of the calculations.On page 89 of the Internal Flooding Report, withinthe 2nd paragraph, a statement is made that aparticular door is assumed to open out, and thatthe flood propagation pathway will go through thatdoor. No discussion, or calculation, is provided tojustify why that particular door will open versusanother of the doors from the room (there aremultiple doors associated with the room). If thereis no basis behind that particular door failing priorto the other doors, then an evaluation of theflooding impacts from other doors opening shouldbe performed.On page 215, there is an un-supportedassumption that drain failures have a failureprobability of 0.1. Need to provide basis for thisassumption.PARTIALLY ADDRESSEDThe supporting calculations referenced in theinternal flooding analysis (PRA-W3-01-002) wereperformed by a vendor for each pipe breakscenario to determine the impacts. Thissupporting documentation was not transmitted bythe vendor along with the primary calculation andthus were not available during the Peer Review.Since the Peer Review, the supportingdocuments have been received from the vendor.This is a documentation only issue and would notimpact this application.As for the door opening assumption, the drawing(G764) referenced in the report shows that this isthe only door that opens "out" of the room otherthan one that lead outside the building. This isonly a documentation issue.In general credit for drains less than 24" indiameter was not given for flood mitigation. Also,the factor of 0.1 was only applied in one room(RAB21-221). The scenarios involved with thisroom have minimal impact to the results (<1.0E-11). In general this treatment is conservative andwould not impact this application. to W3F1-2015-0021Page 14 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionIF-C7-01Internal Flood - Pump HouseFloodOpen forInternal EventsNo Impact onRI-ILRTThe Fire Water pump house has been excludedfrom evaluation on the basis that the failure of thefire pumps will not precipitate a reactor trip andthe fire protection itself is not used to mitigate anyaccident scenario that might lead to core damageother than those occasioned by fire. Thisexclusion needs to be re-visited to determine if aninternal flood in the fire water pump house has thepotential to initiate a flood/spray event elsewherein the plant due to spurious fire water valveactuations (e.g. look at potential forspray/submergence on a fire water control panelto determine if it could cause spurious signals tofire water equipment in the plant resulting in aplant spray/flood event.), and if this inadvertentactuation could result in the need for a plantshutdown. If this impact has been evaluated,document it.NOT ADDRESSEDThe Fire Pump control panel cannot affectsuppression system actuation inside the plant asit reacts to system pressure in the main FireProtection Water loop. The Fire Protection MainControl Panel and Local control panels controlthe operation of suppression within the ReactorAuxiliary building and Turbine building. Anymalfunction in the Fire Pump House would notaffect the main control panel and therefore couldnot cause a release to damage any risksignificant equipment or cause a plant trip. ThisFinding has no impact on the quantified results ofthe IF analysis and therefore no impact on theILRT application. to W3F1-2015-0021Page 15 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionIF-D5a-01Internal Flood - Flood InitiationFrequenciesOpen forInternal EventsNo Impact onRI-ILRTAlthough Waterford calculates the initiating eventfrequency for each evaluated flood scenario usinggeneric data, and the specific calculations arepresented in a footnote for each scenario, areduction factor has been inappropriately appliedto component rupture failure rates. The analysisstates that the generic component failure ratesare obtained from EGG-SSRE-9639 (see Table3.2.1.2 in Flood report). However, these failurerates are then reduced by an additional factor toconvert them from "spray" failures to "rupture"failures. (The example provided shows a "1/27th"reduction for a 1000 gpm valve failure) Theapplication of the reduction factor is inappropriatesince the data are "rupture" rates, not "spray"rates, and the EGG-SSRE-9639 source documenthas already applied a 1/25 reduction factor toensure that the rates are applicable as rupturerates. Need to use the "rupture" failure rateswithout applying the additional reduction factor.NOT ADDRESSEDThe analysis considers such "spray" events ofhaving flow rates up to 100 gpm. Greater rupturerates for flood and major flood are calculatedusing this correlation. From EGG-SSRE-9639:"It should be kept in mind that the externalrupture events include any leakage greater than50 gpm. Therefore, most of the external rupturesidentified-do not involve complete pipeseverance or catastrophic failure of a valve orpump body. The frequencies for suchcatastrophic rupture events should be lower thanthose presented in this report."Use of the factor is based on the Prugh reportreferenced and is implemented to adjust for thesize of the release to be consistent with the sizesconsidered for the pipe failures. This Finding hasnot impact to this application.IF-D7-01Internal Flood - ExcludedScenariosOpen forInternal EventsNegligibleImpact on RI-ILRTThe discussion for excluding the condensatepolisher building from consideration based on theassumption that the operators would bypass thecondensate polisher system in the event of arupture/leak within the building is inadequate.NOT ADDRESSEDThe worst case scenario from a flood in thecondensate polisher building would be a loss ofmain feedwater (with a plant trip) and a loss ofboth 480V switchgears in the building. The FirePRA developed a scenario with these impactswhich had a CCDP of 4.68E-5 (PRA-W3-05-007)which bounds the potential effects of floods in thisbuilding. While specific flood scenarios shouldprobably be developed for this building, it isevident that the contribution to CDF would beminor as the flood frequency still needs to beconsidered as well. Therefore, addressing thisfinding has a negligible impact this application. to W3F1-2015-0021Page 16 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionIF-D7-02Internal Flood - IncorrectScreening MethodOpen forInternal EventsNo Impact onRI-ILRTThe Internal Flooding report is inconsistent  /incorrect in its use of "subsume" versus "screen".For example, in Section 4.2.1.3, the report statesthat scenarios are "subsumed" but the justificationfor subsuming the scenarios is based on thejustification for "screening" of scenarios(screening is defined in SR IF-D7).NOT ADDRESSEDThis is a documentation only finding with noimpact on quantified results and therefore noimpact to this applicationIF-E5a-01Internal Flood - HumanReliability AnalysisOpen forInternal EventsMinimalImpact on RI-ILRTFor operator actions, only actions outside of theControl Room appear to have been reviewed.Also, no analysis could be found to determine ifthere were any "unique" (i.e. not credited in thebase PRA) operator actions that should be addedfor internal flooding recoveries, or if the operatoractions credited were modified to account for thestress level/timing differences associated withinternal flooding scenarios. Of the actionscredited in the base PRA model, 4 of the operatoractions appear to be removed by a recovery rulefile as inaccessible. However, no additionalanalysis was found to justify why these 4 actionswere determined to be inappropriate for internalflooding recovery, or why no other human actionswere impacted by the internal flooding scenarios.NOT ADDRESSEDOperator actions outside the control room arereviewed and the operator action is not credited ifthe flood is on the same elevation as thecomponent being operated locally. Threeadditional actions were developed for specific firescenarios (PRA-W3-01-002).The lack of development of operator actions isone of the main sources of conservatism in thisanalysis. Credit for defined operator actionsrather than conservative assumptions related toflood isolation would serve to greatly improve theresults of this analysis. Since resolution of thisfinding would improve the results, it does nothave negative impact to this LAR. to W3F1-2015-0021Page 17 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionIF-E6-01Internal Flood - UncertaintiesOpen forInternal EventsNo Impact onRI-ILRTIn general, WSES3 used the standardquantification processes from section 4.5.8 of thestandard. However, WSES3 did not propagatethe numerical uncertainties as part of thequantification. WSES3 needs to redo the InternalFlooding Quantification and include thepropagation of the numerical uncertainties andprovide the mean and ERF factors for theresultant CDFs.NOT ADDRESSEDThough no formal uncertainty analysis has beenperformed on the internal flooding model, thisanalysis is based on the internal events modelwhich did have an uncertainty analysis performedon it. The unanalyzed uncertainty associated withthis finding would be due to the initiating eventfrequencies in the IF analysis, that is, the pipebreak frequencies. The associated error factorspresented in the pipe break frequency basisdocument (EPRI TR-1013141) are similar tothose in the internal events analysis (PSA-WF3-01-IE). Also, since the IF contribution to the ILRTis included in a conservative manner, this Findingis judged to have negligible impact on this application.LE-F1b-01Large Early ReleaseFrequency - Conservatism InLERF ResultsClosedAlthough the LERF Model Report (PRA-W3-01-001S12, Revision 1) presents the LERFcontributors, there is no discussion or review ofthe results to indicate there was some evaluationof the significance of various conservatisms.Although Appendix F notes that the contributorshave been reviewed for reasonableness andfound to be typical of what might be expected,there is no documented evidence of this review.
ADDRESSEDA review of the results is documented in thequantification notebook (PSA-WF3-01-QU), notthe LERF analysis.Additionally, multiple cutset review meetings havebeen conducted to ensure the PRA model and itsresults reflected the plant with reasonableaccuracy. These reviews looked at the dominant(top 100 cutsets), some middle cutsets, andcutsets near the truncation limit in the combinedcutset file. The sequence level cutsets were alsoreviewed by looking at each of the individualsequence cutset files. The insights and issuesidentified during these reviews are provided inAppendix F of the Quantification Notebook (PSA-WF3-01-QU). to W3F1-2015-0021Page 18 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition LE-F3-01Large Early ReleaseFrequency - Comparison toother plantsOpen forInternal EventsNo Impact onRI-ILRTTables 4.5.8-2 d and e of the ASME Standardinclude requirements such as documenting areview of a sample of the significant accidentsequences/cutsets, comparing the overall LERFand contributors to similar plants, reviewing asample of non-significant cutsets, identifyingsignificant contributors (such as initiating events,equipment failures, CCFs, and HFEs), review ofcomponent importance measures, and evaluatingthe overall LERF uncertainty intervals. Thesignificant LERF contributors are presented inSection 4.3 of the LERF Report (PRA-W3-01-001S12, Revision 1), a comparison to a similarplant is presented in Section 4.5, and parametricuncertainty was performed in Appendix E, but theother requirements have not been documented.NOT ADDRESSEDThe finding has been partially addressed. Everyelement listed in the finding has not beencompleted and documented. No review ofimportance measures is documented. Besidesthe review of importance measures, all listedrequirements are included in the current modeldocumentation.A review of the results is documented in thequantification notebook (PSA-WF3-01-QU), notthe LERF analysis. A quantitative uncertainlyevaluation (using UNCERT - a Monte Carlosampling software) was also completed toevaluate uncertainly intervals.Additionally, multiple cutset review meetings havebeen conducted to ensure the PSA model and itsresults reflected the plant with reasonableaccuracy. These reviews looked at the dominant(top 100 cutsets), some middle cutsets, andcutsets near the truncation limit in the combinedcutest file. The sequence level cutsets were alsoreviewed by looking at each of the individualsequence cutset files. The insights and issuesidentified during these reviews are provided inAppendix F (PSA-WF3-01-QU).The lack of a formal review of componentimportance measures has no impact on the ILRT LAR. to W3F1-2015-0021Page 19 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionQU-E4-01Sources of UncertaintyClosedThe system notebooks identify the sources ofuncertainty. However, the HRA, AS, IE andsuccess criteria notebooks do not include anyqualitative discussion of uncertainty, though someof them do address the quantitative aspects ofuncertainty. This will facilitate risk informedapplication submittals.
ADDRESSEDEPRI TR-1016737, Table A-1 provides a list of 23topics that are issues for sources of modeluncertainty. This table was reviewed todetermine if these issues where addressed in thequantitative sensitivity analysis or if thecharacterization of the event is consistent withthe EPRI report. Waterford has considered all 23of the model uncertainty issues in the base PRAmodel. In addition, quantitative model sensitivityanalyses were performed on several of theseissues. The Waterford PRA model, asconstructed and documented, facilitates riskinformed application submittals.
SC-A5-01Success Criteria - SuccessBeyond 24 HoursClosedSuccess criteria scenarios that are longer than 24hours are not clearly identified as the mission timeextended to a "safe, stable end-state".Documentation needs to be updated to includediscussion identifying those scenarios with alonger mission time.
ADDRESSEDThe Success Criteria document considers theextension of the mission time beyond the nominal24 hours if the plant is not in a safe & stablecondition (PSA-WF3-01-SC). While no scenarioshave extended mission times, the potential wasconsidered.
SC-B1-01Success Criteria - Wet/Dry FansClosedThere is not a clear basis for the number of wetand dry towers required for either LOCA ortransient success criteria. A reference needs tobe provided if available. If a reference is notavailable, a calculation or evaluation should beperformed to ensure that the success criteria arenot overly bounding - especially with regards tothe transient success criteria. For example,currently transient success criteria requires either14 dry fans or 8 wet fans but does not considercombinations of dry and wet. As this is currentlymodeled, the requirement for 14 dry fans may bemore limiting than the success criteria for LOCAs.
ADDRESSEDA calculation was performed to determine thecombinations of wet and dry cooling tower fansrequired for success under both LOCA andtransient scenarios. Success criteria have beenrevised to reflect the results of the new analysis(PSA-WF3-01-SC Waterford Steam ElectricStation DCT/WCT Success CriteriaDetermination). to W3F1-2015-0021Page 20 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition SC-B1-02Success Criteria - HydrogenFiresClosedGOTHIC code was used to determine room heat-up for the various rooms in the plant. For thebattery room calculation, the GOTHIC codedetermined that room cooling was not required forbattery operation. The PRA currently requiresbattery room cooling due to the potential forhydrogen buildup and potential ignition duringnon-SBO sequences. The potential buildup ofhydrogen is a habitability concern in the room butwill not lead to battery failure without a fire orignition occurring. The ignition of the potentialhydrogen buildup should be considered under thefire PRA evaluation, but not as part of the basePRA model.
ADDRESSEDHydrogen fires have been accounted for in theWF3 Fire PRA. The miscellaneous hydrogen firebin (Bin 19) has been evaluated as described inthe accepted methodology (NUREG/CR-6850).Hydrogen accumulation in the battery rooms wasintentionally neglected in following NUREG/CR-6850 guidance. While the specific battery roomscenario in the F&O could increase risk, theamount of risk increase is considered negligiblewhen compared to the hydrogen fires related tohydrogen systems specifically addressed in theguidance. Waterford staff explicitly followed theapproved guidance in analyzing hydrogen fires.This finding therefore is not applicable to eitherthe Internal Events or Fire PRA models. to W3F1-2015-0021Page 21 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition SC-B3-01Success Criteria - LOCAClassificationsClosedThe current success criteria for LOCAs are basedon plant capabilities and system responses.Although the definitions for small, medium andlarge break LOCAs are reasonable based on thiscriteria, the specific break sizes associated withthe transitions between the LOCA definitions havenot been adequately justified. Currently the breaksizes are based on the original IPE criteria and nothermal hydraulic analyses of the break sizeshave been performed. Per the requirement,thermal hydraulic evaluations are required at alevel of detail to support the definitions/breaksizes so that the appropriate initiating eventfrequencies can be determined. Several utilities'PRAs were dramatically impacted when theMAAP code was used to determine actual breaksizes and some utilities determined that anadditional fourth size LOCA was required toadequately model their plant. This has thepotential to dramatically impact the CDF.
ADDRESSEDThe Initiating Event and Success Criterianotebooks have been updated. Updated MAAPruns based on current plant parameters wereused to verify/re-define LOCA break sizes. (PSA-WF3-01-IE and PSA-WF3-01-SC)  to W3F1-2015-0021Page 22 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition SC-B3-02Success Criteria - BatteryDepletionOpen forInternal EventsMinimalImpact on RI-ILRTSuccess criteria for the battery depletion of the Aand B batteries specify that the batteries willsurvive for 4 hours if non-essential loads arestripped within 30 minutes. The success criteriafor the AB battery specify a 6 hour coping evenwithout any load stripping. There is no discussionof the impact or battery capability for the A & Bbatteries if loads are not stripped. Need toprovide additional information and references forthe battery depletion timing. Specific items thatneed to be addressed include:  Impact of operators failing to strip loads within30 minutes  How long will batteries last without stripping  Impact on potential steam generator overfillonce batteries are depleted (EFW AOVs fail fullopen upon battery depletion, but EFW steampump still providing full flow to both steamgenerators)  Separate operator actions to strip loads need tobe included in the PRA model for the AB batteryand the A & B batteries. Currently a singleoperator action (EAFSTRBATP) is used for loadstripping for all the batteries with a probability ofzero failure being assigned to it. This isacceptable for the AB battery since it is notdependent upon stripping, but is not acceptablefor the A & B batteries since they are dependentupon stripping loads within 30 minutes. Provide a reference to the SBO copinganalysis (ECE89-016, Rev. 3) and any otherreferences associated with the battery depletioncalculations.NOT ADDRESSEDThe basis for crediting a 6 hour coping time wasa study calculation developed for PRA (ER-W3-2002-0622) by removing the conservatisms fromthe design basis calculations. This calculationshows that with load shedding more than 6 hoursis available on A, B, and AB battery loads.Without load shedding the calculation indicatesthat 2.5 hours would be the most limiting time forbattery depletion for A and B. Currently credit isgiven for the time to steam generator overfill oncethe EFW AOVs fail open once the batteriesdeplete along with at least an additional hourbefore core damage conditions are met (PSA-WF3-01-SC, PSA-WF3-01-AS). Even if these twotime periods are considered, the 6 hours given foroffsite power recovery is still slightly nonconservative.A human reliability analysis was performed toanalyze the operator action to shed the A and Bbattery loads within 30 minutes and given a valueof 8.4E-2 (PSA-WF3-01-HR). Although this actionhas not been included in the model, its inclusionwould allow for 6 hours to be allowed for offsitepower recovery for over 90% of the SBO cutsets.The remaining 10% however would requirehigher recovery values. Station Blackoutsequences accounts for roughly 45% of thecutsets (PSA-WF3-01-QU). Therefore, only 4.5%of cutsets would require an increased non-recovery factor. Even if this factor was increasedfive fold, the impact would be less than 25%higher CDF. Resolution of this F&O could have asmall, but potentially noticeable impact on this application. to W3F1-2015-0021Page 23 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition SC-B5-02Success Criteria - SupportingAnalysesClosedThe success criteria documentation does notexplicitly discuss the reasonableness andacceptability of the thermal hydraulic andsupporting engineering bases used to support thesuccess criteria. Appendix B references oldindustry peer reviews and past IPE evaluations.These analyses are out of date and a newcomparison to current analyses needs to beconducted.
ADDRESSEDA comparison of the success criteria betweenWF3 and ANO2 was performed to verify thereasonableness and acceptability of the thermalhydraulic analyses and supporting engineeringbases. This comparison is documented inAppendix B of PSA-WF3-01-SC.
SC-C1-02Success Criteria - InadequateReferencesClosedThroughout the document there are a number ofassumptions and statements made that directlyimpact the success criteria but do not have anyreferences identified to justify their bases.Querying the PRA group determined that most ofthe statements were based on valid references,but they were not identified in the success criteriadocumentation. The references need to bespecifically identified and included.
ADDRESSEDThe Success Criteria notebook has been updatedand includes a more thorough application ofreferences.
SC-C3-01Success Criteria - BatteryUnavailabilityClosedThe success criteria notebook specifies thatmaintenance events associated with the batteriesand chargers are included in the PRA model. Areview of the model indicates that the chargershave a reasonable unavailability time modeled forthem, but that the batteries currently show anunavailability of zero. This should be re-evaluatedsince normal practices include isolation of thebatteries for discharge testing at other utilities andit should be verified if the same practice isemployed here.
ADDRESSEDThe current (updated) model includes a value forbattery unavailability. to W3F1-2015-0021Page 24 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionSY-A8-01System Modeling - ComponentBoundariesClosedNeed to reference and verify that the componentboundaries used match the component failuredata in the Data notebook. Pay particularattention to the diesels.
ADDRESSEDThe Waterford Internal Events PRA model wasrecently updated (August 2013). The updatedsystem notebooks reference the Entergydocument PRA-ES-01-003, which definescomponent boundaries. The boundaries used fordiesels are correct.SY-A12b-01System Modeling - FlowDiversion PathwaysClosedNeed to use the exclusion criteria in SY-A14 tojustify excluding flow diversion pathways. Usingthe criteria 2 normally closed valves should beeasily justified using criteria SY-A14(a). Thecriteria for excluding based on a 1 to 3 ratiobetween the primary piping and the potentialdiversion piping needs to be backed up bypressure differentials. This exclusion criteria isvalid if the system pressures between the primaryand potential diversion piping is the same orsimilar. If the pressure differential is high, furtheranalysis is required to justify exclusion.Overall, the assumptions used to exclude specifictypes of failures needs to be reevaluated andjustification provided on how the exclusion criteriais met.ADDRESSEDFlow diversion pathways were reviewed for theWaterford 3 Fire PRA to determine if additionalpathways needed to be included to addresspotential spurious actuation opening poweroperated valves. As part of this review, the flowdiversion pathways excluded due to the 1/3 rulewere reviewed to verify that no pressuredifferential is present and that sufficient margin isbuilt into the system flow. Considerations forextended time (up to 24 hours) for systems thatmeet the 1/3 criteria resulted in additions to themodel. Flow diversion of the CCW and CCWMakeup systems could cause system failure.These failures were added to the model (bothinternal events and FPRA).SY-A12b-02System Modeling - FlowDiversion of HPSIClosedThe flow diversion path to the SIT (CV transfersopen) causing HPSI failure is inappropriate andshould be removed from the model. The checkvalve will have primary system pressure keeping itclosed. At pressures greater than the SITinjection pressure, the check valve cannotphysically open and allow water into the SITunder these conditions.
ADDRESSEDThe logic associated with this finding is no longerin the PRA model. to W3F1-2015-0021Page 25 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionSY-A16-01System Modeling - LogicSequenceClosedOHFRCPTRIP should be ANDed with loss ofCCW to Seals under gate QT05 ADDRESSEDOHFRCPTRIP should not be ANDed with loss ofCCW to Seals under gate QT05 because thatwould create circular logic. QT05 is linked bylogic to the failure of Sequencer A which isneeded for successful restart of the CCW pumpsA and AB. A detailed review of the system logicand system operation revealed that the currentmodel logic is correct.SY-A16-02System Modeling - Missing HRAClosedDocumentation states: If a loss of CCW pumpoccurs, the stand by pump is started. If thesecond CCW pump cannot be started, then theCCW headers must be split in accordance withOP-901-510. In the event that no CCW pumps arerunning or can be started then the following mustbe performed within 3 minutes:This action doesn't appear to be modeled. Textmissing in notebook (after colon).
ADDRESSEDThe documentation has been updated and therelevant information is now included. This actionis modeled in the internal events PRA. to W3F1-2015-0021Page 26 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionSY-A18a-01System Modeling - CoincidentUnavailabilityOpen forInternal EventsNo impact onILRTHPSI system has an installed spare that can bealigned to either system. Coincident unavailabilitydue to maintenance for redundant equipment ispossible (spare pump OOS for extended periodsand could be OOS with another pump). Need tospecifically address this possibility. This may alsobe true for charging pumps.NOT ADDRESSEDSR SY-A18 (changed to SY-A20 in latest versionof the standard) states:  INCLUDE eventsrepresenting the simultaneous unavailability ofredundant equipment when this is a result ofplanned activity. The Plant Specific Failure DataDevelopment analysis (PSA-WF3-01-DA-01)documents the inclusion of all planned concurrentmaintenance (including installed spares). Thisremains 'not addressed' due to documentation.The coincident unavailability is included in themodel, however the documentation does not fullyexplain the process used to consider/modelevents.The lack of documentation has no impact on thequantified results of the ILRT.SY-B4-01System Modeling - MissingCCF CombinationsClosedCommon cause failure modeling of the 2/4 failurecombinations needs to be reevaluated. Forexample, in the HPSI model check valves SI-241-244 are currently modeled with individualcomponent failures and combinations of 3 or 4failures. Combinations of 2 failures are excluded.This is inappropriate since a combination of 2failures on train A combined with the break(LOCA) on train B would fail the system success.This is currently not accounted for in themodeling. Another example in SI is the modelingof the hot leg injection isolation MOVs and CVs donot include 2/4 failures. Although there is anassumption associated with this, the logic behindthe assumption no longer meets the criteria formodeling common cause failures. Non-lethalcommon cause combinations must be included toensure their impact associated with individualcomponent failures is adequately addressed.
ADDRESSEDReviewed CCF modeling of all the systems andfound several CCF modeling conditions whichwould impact fire PRA results. This led to animmediate model update. Several previouslyexcluded non-lethal CCF combinations wereadded to the model. The changes and resultswere summarized in the Excel spreadsheet"CCF-Disposition" and incorporated in the model. to W3F1-2015-0021Page 27 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionSY-B13-01System Modeling - Control Room HVACClosedIn Table A-4 of the success criteria notebook theGOTHIC code determined that control roomventilation was required; however, Section 1.8,Major Assumption specifies that control roomHVAC is not included in the HVAC system model.Although it is possible to perform a plantshutdown from the remote shutdown panel,different actions and equipment are availableunder this scenario and it should be consideredas a recovery action, not as a standard action.Loss of control room HVAC needs to be includedin the PRA model with the recovery actionsassigned based on plant conditions andequipment available at the remote shutdownpanels during the scenarios.
ADDRESSEDThe latest PRA internal events model updateincluded the addition of MCR HVAC. Failure ofcontrol room HVAC is included in the MCR HVACsystem model and notebook, and not as a sub-system in the HVAC notebook. Loss of controlroom HVAC is also included, in the reactor tripinitiating event frequency since it is a shortduration shutdown limiting condition for operation.SY-B16-01System Modeling - OperatorInterface DependenciesClosedNo discussion of operator interface dependenciesacross systems or trains are provided within thesystem notebooks. Need to add this discussion,or state that this information is provided in theHRA documentation (as appropriate).
ADDRESSEDAs part of the system model and notebookupdate effort, a cross-reference between thesystem notebooks and the HRA evaluation wasadded to specify that operator interfacedependencies across systems or trains areaddressed in the HRA documentation.
SY-C2-01System Modeling - Intersystem CCFOpen forInternal EventsNo Impact onRI-ILRTNeed to add a discussion of what the criteria forCCF considerations are (which types ofcomponents were looked at, were inter- and intra-system CCFs considered, etc. If the componenttypes were determined based off of a list from aReference, provide this information and a pointerto the reference document/methodology.NOT ADDRESSEDThis finding is a documentation issue only.During the disposition of the F&O for SY-B4, thecriteria for CCF considerations were reviewed.Inter-system and Intra- system CCFs consideredare documented in the CCF calculation, but notexplicitly in each system notebook.The lack of documentation documented in thisfinding has no quantitative impact on the internalevents model or the results in this LAR. to W3F1-2015-0021Page 28 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDisposition SY-C2-02System Modeling - TempDiesel GeneratorsClosedStatement under Operator Interface says"temporary diesel generators must be manuallyaligned and started as part of the accident.Therefore, -"  This statement implies that theTEDGs are credited via a post-initiator operatorrecovery action - need to clarify that the TEDGsare not credited in the Base Model, and are onlyused for EOOS.
ADDRESSEDAssumption 29 in the updated AC power systemnotebook addresses this finding:"The temporary EDGs (TEDGs) are notcredited in the W3 Internal Events PRA, butare retained for EOOS alignments only."
SY-C3-01System Modeling -AssumptionsClosedNeed to review system notebooks assumptionssection and remove assumptions associated with"circular logic" and how it is handled in the systemfault trees. The handling of circular logic asdiscussed in the system notebooks does notappear to always "mesh" with how it is describedin the circular logic notebook.
ADDRESSEDAssumptions related to circular logic areaccurately documented in the Circular LogicAnalysis (PRA-W3-01-004, Rev. 0). Theassumptions in this document do "mesh" with themodeled logic. A goal of the previous revisionwas to remove the circular logic assumptions,from the System Notebooks. Some assumptionswere inadvertently left in the notebooks. Therecently revised system notebooks no longerhave the circular logic discussions referenced inthe finding. All content in the system notebooksrelated to circular logic reference the CircularLogic Analysis document. to W3F1-2015-0021Page 29 of 29Internal Events PRA Peer Review - Facts and Observations (Findings Only)FindingTopic (& Associated SR)
StatusFinding/ObservationDispositionQU-E2-01(Originally givenas Suggestion)Assumptions and Sources ofUncertaintyOpen forInternal EventsNo Impact onRI-ILRTAssumptions are identified in the systems andother Notebooks. However, there is no discussionof the impact of these assumptions on the results.It is recommended that in the QU Notebook, aqualitative discussion be provided which reviewsall these assumptions and identifies a set ofsensitivity runs to be made to study the impact ofthese assumptions on the results of the PRA.There is no documentation of a systematic reviewof all the PRA assumptions to identify the list ofsensitivity studies to be carried out. However,many sensitivity studies have been conducted.Perform and document a systematic review ofPRA assumptions to identify the list of sensitivitystudies to be carried outNOT ADDRESSEDThe updated PRA Standard requirement waschanged to include documentation of how thePRA model is affected by model uncertaintiesand assumptions. However, this F&O is related tomodel documentation and will not quantifiedresults. Therefore, resolution of this F&O wouldhave no impact on this application.QU-E4-02(Originally givenas Suggestion)Assumptions and Sources ofUncertaintyOpen forInternal EventsNo Impact onRI-ILRTAdditional evaluation is recommended to performa more systematic assessment the uncertaintyassociated with success criteria, modelinguncertainties, degree of completeness in theselection of initiating events, and possible spatialdependencies. The requirement is toDOCUMENT assumptions and sources ofuncertainty, which is met. This suggestion is tosystematically assess and determine the impactof the qualitative uncertainty items.NOT ADDRESSEDThe updated PRA Standard requirement waschanged to include documentation of how thePRA model is affected by model uncertaintiesand assumptions. However, this F&O is related tomodel documentation and will not quantifiedresults. Therefore, resolution of this F&O wouldhave no impact on this application. toW3F1-2015-0021Calculation, Waterford 3 Evaluation of Risk Significance of an ILRT Extension ANO-1 ANO-2 GGNS IP-2 IP-3 PLP JAF PNPS RBS VY W3 NP-GGNS-3 NP-RBS-3CALCULATIONCOVER PAGEEC #  56864Page 1 of 72Design Basis Calc.
YES NO CALCULATION EC MarkupCalculation No:ECS14-010Revision: 1 Title:Waterford 3 Evaluation of Risk Significance of an ILRTExtension Editorial YES NOSystem(s):MISCReview Org (Department):DEMECHSafety Class:Safety / Quality RelatedAugmented Quality ProgramNon-Safety RelatedComponent/Equipment/Structure Type/Number:Document Type: B13.40Keywords (Description/TopicalCodes):PRAREVIEWSName/Signature/DateVendor Prepared (See Page 5)Name/Signature/DateMark Thigpen(See Associated EC)Name/Signature/DateChris Talazac(See Associated EC)Responsible EngineerDesign VerifierSupervisor/ApprovalReviewer  Comments Attached  Comments Attached CALCULATION REFERENCE SHEET CALCULATION NO: ECS14-010REVISION: 1I. EC Markups Incorporated(N/A to NP calculations) 1.N/A 2.3.4.5.II. Relationships
:Sht Rev Input Doc Output DocImpact Y/NTracking No.1.PE-005-001 "ContainmentIntegrated Leak Rate Test" 42.PRA-W3-01-001S12 "WF3Large Early ReleaseFrequency (LERF) Model" 13.ECS04-001 04.Waterford 3 Emergency Plan 455.III. CROSS REFERENCES
:1."Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50,Appendix J," Revision 3-A, Nuclear Energy Institute, NEI 94-01, July 2012 2.3.4.5.IV. SOFTWARE USED
:Title: CAFTAVersion/Release:5.4 Disk/CD No.
V.DISK/CDS INCLUDED
:Title:Version/Release Disk/CD No.
VI.OTHER CHANGES
:
RevisionRecord of Revision 0Initial issue.
1Revised to include contribution from Internal Floods in risk calculation.Additional sensitivity cases performed for updated Level 2 model. (ReferenceCondition Report CR-WF3-2015-2252)
RSC 14-12/ECS14-010 (Rev.1)
Reliability and Safety Consulting Engineers, Inc. 2220 Award Winning Way, Suite 200  Knoxville, TN 37932 USA www.rscsite.com Waterford 3 Evaluation of Risk Significance of an ILRT Extension Revision 1 April 2015 Principal Analyst Vincent Young Developed for Entergy 
 
RSC 14-12/ECS14-010 (Rev.1)
Document Revision History Revision Date Released Principal Author Reviewer Initials Approval Initials Summary of Revision Draft 7/22/2014 S. Pionke JM/RS Not Required 0 8/8/2014 RS RS Original document 1 4/21/2015 V. Young RS RS Updated to include internal flooding PRA results 2      3      4      5 RSC 14-12/ECS14-010 (Rev.1)
Report Quality Assurance Attribute  (comments are located in comment resolution form or electronically noted in text during review)
Attribute Applicable (Yes/No) Attribute Reviewed (Yes/No) Title Page or Calculation Cover - Contains the title, client, originator, reviewer and approver. Provides information related to revision, and level of review.
Yes Yes Review Comment and Resolution Form - This documents the review process and includes the reviewer comments, concurrence and originator resolution.
Yes Yes Table of Contents, including figures and tables - provides a listing of all major sections, drawings, figures, tables, and illustrations.
Yes Yes Introduction - summary description of the purpose, scope, and the principle tasks required to meet the project objectives. Analysis boundaries, where applicable function. What is included or excluded from the analysis.
Yes Yes Methodology - Describe the process and supporting methodology that is sufficient to understand the approach and to support a peer review. Is the method consistent with RSC Engineers and client standards and practices? Does the method document consideration of special issues (e.g., common cause, circular logic, and asymmetry)?
Yes Yes Analysis and Results - Detailed documentation of the implementation of the task steps that may be supported by report appendices, including any intermediate and final results. Does the analysis use appropriate and verified codes and data input? All figures of event tree and fault trees, and sequence cut sets must be reviewed even if not documented in report. Ensure adequate tables to support assessment such as support systems, success criteria, operator actions, systems addressed in the analysis and dependency. Discussion of system fault tree models, success criteria, application, and system operation as required. Listing and discussion of data selection and application as appropriate.
Yes Yes Conclusions and Recommendations - A concise presentation of the results of the analysis that answer the objective of the analysis. It should highlight important aspects and findings of the assessment and also provide information related to important assumptions and any conservatism or analysis uncertainty present in the analysis. Recommendations (if any) should be based on analysis results. Limitations of the analysis should be clearly listed. Listing of both general and specific assumptions for system analysis is required. For any quantification adequate truncation requirements should be mentioned. Importance and sensitivity assessments for important contributors and uncertain issues as appropriate.
Yes Yes List of References - Documents all sources used in the development of the analysis, document, or model that would be necessary to verify or repeat the analysis. References should be included for any non-document files (Visio, Excel, CAFTA, etc.) supporting the report.
Yes Yes RSC 14-12/ECS14-010 (Rev.1)
Report Quality Assurance Appropriate and Necessary Appendices - Provide adequate supporting documentation to be a able to review and draw conclusions from the report.
This would include any applicable appendices such as any "raw" data used in the analysis, any calculations performed to support the analysis that are not documented in a calculation, or appendix containing the analysis cut sets or other results listings as appropriate.
Yes Yes Reviewer Qualification Statement  I certify that I am qualified under the RSC Engineers QA/QC program to perform the review of this document and have examined the above attributes for the most current revision. Signature/ Date R. Summitt 4/21/15 Approver Qualification Statement I certify that I am a qualified approver under the RSC Engineers QA/QC program. I have reviewed the completed documentation and the methods, analysis and documentation meet applicable industry practices for concept and conformity. R. Summitt 4/21/15
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) i Printed: 4/23/2015 Table of Contents Table of Contents
........................................................................................................................... i List of Tables
...............................................................................................................................
.. ii 1.0  PURPOSE ..................................................................................................................
........... 11.1 
 
==SUMMARY==
OF THE ANALYSIS ........................................................................................ 11.2 
 
==SUMMARY==
OF RESULTS/CONCLUSIONS ...................................................................... 22.0  DESIGN INPUTS ............................................................................................................
....... 43.0  ASSUMPTIONS ..............................................................................................................
..... 134.0  CALCULATIONS .............................................................................................................
.... 144.1  CALCULATIONAL STEPS ............................................................................................... 144.2  SUPPORTING CALCULATIONS ..................................................................................... 165.0  SENSITIVITY STUDIES ...................................................................................................... 305.1  LINER CORROSION ....................................................................................................... 305.2  DEFECT SENSITIVITY AND EXPERT ELICIATION SENSITIVITY ................................ 355.3  POTENTIAL IMPACTS FROM EXTERNAL EVENTS ..................................................... 3
 
==66.0  REFERENCES==
...............................................................................................................
..... 40 Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) ii Printed: 4/23/2015 List of Tables Table 1 Summary of Risk Impact on Extending Type A ILRT Test Frequency .................................. 3Table 2 Release Category Frequencies ..........................................................................................
..... 5Table 3 Decomposition of WF3 LERF Frequency and EPRI Classification ........................................ 6Table 4 Reported Person Rem Estimates for Zion Source Term Groups  (summarized from NUREG/CR-4551).................................................................................................................
........ 8Table 5 Assignment of Zion Source Term Groups to EPRI Classes
................................................. 10Table 6 Predicted Dose from Reference 10 ....................................................................................... 11Table 7 Calculation Parameters Solving for the Scaling Constant (C) .............................................. 12Table 8 Calculation Parameters for the Dose at 25 Miles ................................................................. 12Table 9 WF3 Dose for EPRI Accident Classes .................................................................................. 1 3Table 10 EPRI Containment Failure Classifications .......................................................................... 17Table 11 WF3 PRA Release Category Grouping to EPRI Classes .................................................. 18Table 12 Baseline Risk Profile ................................................................................................
............ 22Table 13 Risk Profile for Once in Ten Year Testing ...........................................................................
24Table 14 Risk Profile for Once in Fifteen Year Testing ...................................................................... 2 6Table 15 Impact on LERF due to Extended Type A Testing Intervals .............................................. 28Table 16 Impact on Conditional Containment Failure Probability due to Extended Type A Testing Intervals .....................................................................................................................
.................. 29Table 17 WF3 Liner Corrosion Risk Assessment Results Using CCNP Methodology .................... 31Table 18 Liner Corrosion LERF Adjustment Using CCNP Methodology .......................................... 33Table 19 WF3 Summary of Base Case and Corrosion Sensitivity Cases ........................................ 34Table 20 Comparison of Jefferys Non-Informative Prior and Expert Elicitation Values
.................... 35Table 21 WF3 Summary of ILRT Extension Using Expert Elicitation Values ................................... 36Table 22 WF3 Upper Bound External Event Impact on ILRT LERF Calculation .............................. 39 Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 1 Printed 04/23/15 1.0  PURPOSE The purpose of this report is to provide an estimation of the change in risk associated with extending the Type A integrated leak rate test interval beyond the current 10 years specified by 10 CFR 50, Appendix J, Option B [1] for Waterford Steam Electric Station Unit 3 (WF3). This activity supports a request for an exemption from the performance of the integrated leak rate test (ILRT) during the planned refueling outage number 20. The assessment is consistent with the processes described in the methodology identified in EPRI's guidance document, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals [2]. Some of the values calculated in this analysis involve very small changes. The detailed calculations performed to support this report were of a level of mathematical significance necessary to calculate the results recorded [20]. However, the tables and illustrational calculation steps presented may present rounded values to support readability. 1.1 
 
==SUMMARY==
OF THE ANALYSIS The reactor containment leakage test program consists of three tests (Type A, Type B, and Type C) [1]. These tests periodically verify the leak-tight integrity of the primary reactor containment and the systems (and their components) penetrating the containment. Type A testing is intended to measure the overall integrated leak rate which is the summation of leakage through all potential leakage paths including containment welds, valves, fittings, and components which penetrate containment. The type B test measures leakage across each pressure-containing or leakage-limiting boundary for a magnitude of containment penetration seals (i.e. resilient seals, gaskets, sealant compounds, flexible metal seal assemblies, air lock door seals, etc.). The final type of testing, Type C, measures containment isolation valve leakage rates. This type of testing is applicable for any valves that provide a direct connection between the inside and outside atmospheres of the primary reactor containment under normal operation, are required to close automatically upon receipt of a containment isolation signal, are required to operate intermittently under post-accident conditions, and are in main steam, feedwater, and other system piping which penetrate containment of direct-cycle boiling water
 
power reactors. 10 CFR 50, Appendix J allows individual plants to extend Type A surveillance testing requirements and to provide for performance-based leak testing. This report documents a risk-based evaluation of the proposed change of the ILRT interval for the WF3. The proposed change would impact testing associated with the current surveillance tests for Type A leakage, procedure PE-005-001 [3]. No change to Type B or Type C testing is proposed at this time. This analysis utilizes the guidelines set forth in NEI 94-01 [4], the methodology used in the EPRI Report [2], and considers the submittals generated by other utilities. This calculation evaluates the risk associated with various ILRT intervals as follows:  3 years - Interval based on the original requirements of 3 tests per 10 years. 10 years - This is the current test interval required for WF3. 15 years - Proposed extended test interval.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 2 Printed: 4/23/2015 The analysis utilizes the WF3 PRA results taken from the Level 2 model [5]. The analysis also includes the PRA results taken from the WF3 internal flooding (IF) model [23]. The release category and person-rem information is based on the approach suggested by the EPRI guidance document [2]. 1.2 
 
==SUMMARY==
OF RESULTS/CONCLUSIONS The specific results are summarized in Table 1 below. Type A testing risk is comprised of EPRI Class 3a and Class 3b. Class 3b is defined as the large early release (LERF) contribution to Type A testing. A breakdown of all the EPRI classifications is contained in Tables 9 and 10 of this report.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 3 Printed: 4/23/2015 Table 1 Summary of Risk Impact on Extending Type A ILRT Test Frequency Risk Impact for 3-years (baseline) Risk Impact for 10-years (current requirement) Risk Impact for 15-years Total integrated risk (person-rem/yr) 3.46E+2 3.46E+2 3.46E+2 Type A testing risk (person-rem/yr) 1.25E-2 4.14E-2 6.25E-2 % total risk  (Type A / total) 0.004% 0.012% 0.018% Type A LERF (Class 3b) (per year) 1.41E-8 4.70E-8 7.04E-8 Changes due to extension from 10 years (current)  Risk from current (Person-rem/yr) 2.01E-2 % Increase from current
( Risk / Total Risk) 0.006%  LERF from current (per year) 2.35E-8  CCFP from current 3.53E-3 Changes due to extension from 3 years (baseline)  Risk from baseline (Person-rem/yr) 4.82E-2 % Increase from baseline
( Risk / Total Risk) 0.014%  LERF from baseline (per year) 5.64E-8  CCFP from baseline 8.47E-3  The results are discussed below:  The person-rem/year increase in risk contribution from extending the ILRT test frequency from the current ten (10) year interval to a fifteen (15) year interval is 2.01E-2 person-rem/year. The risk increase in LERF from extending the ILRT test frequency from the current ten (10) year interval to a fifteen (15) year interval is 2.35E-8/yr. The change in conditional containment failure probability (CCFP) from the current ten Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 4 Printed: 4/23/2015 (10) year interval to a fifteen (15) year interval is 3.53E-3/yr. The change in Type A test frequency from once (1) per ten (10) years to once (1) per fifteen (15) years increases the risk impact on the total integrated plant risk by only 0.006 percent. Also, the change in Type A test frequency from the original three (3) per ten (10) years to once (1) per fifteen (15) years increases the risk only 0.014 percent. Therefore, the risk impact when compared to other severe accident risks is negligible. Regulatory Guide 1.174 [6] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10
-6/yr and increases in LERF below 10
-7/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from a once (1) per ten (10) years to once (1) per fifteen (15) years is 2.35E-8/yr. Guidance in Regulatory Guide 1.174 defines very small changes in LERF as below 10
-7/yr, increasing the ILRT interval from ten (10) to fifteen (15) years is therefore considered non-risk significant and the results support this determination. In addition, the change in LERF resulting from a change in the Type A ILRT test interval from a three (3) per ten (10) years to once (1) per fifteen (15) years is 5.64E-8/yr. The delta LERF is also below the guidance classification of a very small change. Regulatory Guide 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy. Consistency with defense-in-depth philosophy is maintained by demonstrating that the balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in conditional containment failure probability was estimated to be 3.53E-3 (0.46 percent increase) for the proposed change and 8.47E-3 (1.10 percent increase) for the cumulative change of going from a test interval of three (3) in ten (10) years to one (1) in fifteen (15) years. Both CCFP changes meet the criterion of less than 1.5 percent increase obtained from the EPRI guidance document [2]. Therefore the changes in CCFP are considered small and demonstrate that the defense-in-depth philosophy is maintained. In reviewing these results, the WF3 analysis demonstrates that the change in plant risk is small as a result of this proposed extension of ILRT testing. The change in LERF defined in the analysis for both the baseline and the current cases is within the acceptance criterion. In addition to the baseline assessment, three sensitivity exercises are included. These analyses are provided in Section 5 and are consistent with the methods outlined in the EPRI guidance
 
document [2]. 2.0  DESIGN INPUTS The WF3 PRA is intended to provide "best estimate" results that can be used as input when making risk informed decisions. The PRA provides the most complete results for the WF3 PRA. The inputs for this calculation come from the information documented in the WF3 PRA Level 2
 
model [5] and the WF3 IF model [23]. The WF3 release states are summarized in Table 2. WF3 Level 2 results are grouped into four accident sequence states that represent the summation of individual accident categories. The internal flooding initiating event model was not propagated through the Level 2 model.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 5 Printed: 4/23/2015 However, a review of the flooding cut sets indicates that the accident sequences are similar in nature to the internal transient events. The transient initiating event contribution to the release categories is approximately fifty (50) percent INTACT and fifty (50) percent LATE. Since the internal flood scenarios are similar and the flooding scenarios would not impact the core melt and containment phenomena, the same split is applied to the CDF contribution from the internal flooding initiating events. The CDF frequency is equally split between the INTACT plant damage state (PDS) category (1.24E-6/yr) and the LATE PDS category (1.24E-6/yr). The number of sequences comprising each sequence state is also presented in Table 2. Table 2 Release Category Frequencies Release Category Contributing WF3 Accident Categories Frequency (/yr)
EPRI Classification INTACT (S) 10 1.57E-6 Class 1 LERF 1 18 5.31E-7 Class 8 SERF 9 1.76E-9 Class 6 LATE 14 4.56E-6 Class 7 Total N/A 6.66E-6 N/A 1. The LERF contribution for WF3 contains early containment failures due to containment phenomenon and by the EPRI guidance these should be collected in Class 7. To accurately classify the contributions, the LERF contribution is separated to be consistent with the EPRI guidance document
[2]. Table 4.3-2 of the WF3 Level 2 model [5] analysis provides the endstate and frequency of the respective endstate. Table 3 shows the classification of each endstate and the totals of each classification. The description of the outcome is used to classify each of the 18 contributing LERF endstates.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 6 Printed: 4/23/2015 Table 3 Decomposition of WF3 LERF Frequency and EPRI Classification Endstate Description of Outcome Frequency (per year)
EPRI Class LERF01 Containment failure following high-pressure (HP) vessel breach (VB) - Non-SBO 5.12E-9 7 LERF02 Containment failure following HP VB - Non-SBO  1 7 LERF03 Containment failure following low pressure (LP) VB - Non-SBO 1.56E-9 7 LERF04 Temperature induced (TI) SGTR - Non-SBO 1.07E-8 8 LERF05 Containment failure following LP VB - Non-SBO 2.06E-9 7 LERF06 Pressure induced (PI) SGTR - Non-SBO 2.98E-9 8 LERF07 Containment failure following LP VB - Non-SBO 3.35E-10 7 LERF08 Loss of isolation - Non-SBO 1.46E-8 2 LERF09 Containment bypass - Non-SBO 4.38E-7 8 LERF10 Containment failure following LP VB - SBO  1 7 LERF11 Containment failure following HP VB - SBO 1.19E-11 7 LERF12 Containment failure following LP VB - SBO 3.55E-9 7 LERF13 TI-SGTR - SBO 2.45E-8 8 LERF14 Containment failure following LP VB - SBO 4.79E-9 7 LERF15 PI-SGTR - SBO 7.26E-9 8 LERF16 Containment failure following LP VB - SBO  1 7 LERF17 Loss of isolation - SBO 1.84E-9 2 LERF18 Containment bypass - SBO 1.41E-8 8 Contribution to EPRI Classification 2 1.64E-8 Contribution to EPRI Classification 7 1.74E-8 Contribution to EPRI Classification 8 4.98E-7 Total LERF 5.31E-7 1. represents a probabilistica lly insignificant value.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 7 Printed: 4/23/2015 In order to develop the person-rem dose associated to the plant damage state it is necessary to associate each release category with an associated release of radionuclides and from this information to calculate the associated dose. The EPRI guidance on leak rate testing [2] indicates that a surrogate can be applied and is acceptable for estimating risk and suggests one surrogate source is the results contained in NUREG-1150 [7]. NUREG-1150 examined both pressurized water reactors (PWRs) and boiling water reactors (BWRs). The results presented for BWRs (i.e., Peach Bottom, Grand Gulf) are not considered appropriate for this analysis since the core melt mechanics and design are substantially different between WF3 PWR design and the BWRs. Therefore, their results are excluded from consideration. NUREG-1150 also analyzed the Zion, Sequoyah, and Surry PWR designs. Sequoyah utilizes an ice condenser design and the presence of ice and restricted flow paths can lead to sequences and conditions that are not found in a large dry containment design such as WF3.
Therefore, Sequoyah is not considered a good PWR design for comparison. Surry is a 3-loop Westinghouse design large dry containment and may be somewhat closer to the WF3 design. However the 3-loop design and power level may influence source term composition. Therefore it is not selected as a surrogate. The remaining assessed design is Zion. It is a Westinghouse 4-loop design and given the power level and other factors, is considered the best surrogate after examination of the NUREG-1150 analyzed plants. NUREG/CR-4551 [8] provides the Level 2 analysis and offsite consequence assessment for Zion. Table 4.3-2 of that document provides a summary of consequence results that includes population dose (exposure) within fifty (50) miles for internal events. The exposure estimates for a range of fifty (50) miles around the Zion site are provided in Table 4 for each reported source term group.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 8 Printed: 4/23/2015 Table 4 Reported Person Rem Estimates for Zion Source Term Groups  (summarized from NUREG/CR-4551) Source Term Grouping Exposure (rem) 1 1.69E+5 2 3.76E+5 31 1.93E+5 33 3.66E+4 61 2.76E+5 64 6.06E+5 65 1.40E+6 66 2.90E+5 67 1.35E+6 68 2.72E+6 69 6.93E+5 70 2.18E+6 71 3.91E+6 72 1.56E+6 100 3.38E+6 101 4.42E+6 103 5.80E+6 104 5.46E+6 105 6.49E+6 106 8.47E+6 107 6.27E+6 136 9.00E+6
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 9 Printed: 4/23/2015 Table 4 (continued) Reported Person Rem Estimates for Zion Source Term Groups (summarized from NUREG/CR-4551) Source Term Grouping Exposure (rem) 137 7.19E+6 139 1.34E+7 140 8.98E+6 142 1.41E+7 143 1.09E+7 172 1.90E+7 173 1.55E+7 175 3.24E+7 176 1.94E+7 178 4.11E+7 179 3.93E+7 301 1.27E+2 302 6.18E+2 303 3.59E+3 In order to utilize this information it is necessary to convert it to the form needed in the ILRT analysis. This involves classification into one of the four EPRI classes and then determining the
 
representative person-rem estimates. Table 3.4-4 in NUREG/CR-4551 [8] provides some guidance with respect to the composition of the source term grouping. The highest contributing release type was credited to the corresponding EPRI class. While multiple release types are contained in Table 3.4-4, only eight of the categories contained the majority of the release. Zion labeled these categories as Isolation Leak, SGTR, LS, LL, EL, Alpha, NoCF, and BMT. Class 1 consists of any source term groups that are dominated by no containment failures (NoCF). EPRI Class 2 is related to isolation faults; therefore, source term groups with Is. Leak as the main contributor are placed into this EPRI class. EPRI class 7 is related to early and late phenomena-induced failures.
Zion categories LS, LL, EL, Alpha, and BMT are all associated with these types of failures.
EPRI Class 8 pertains to containment bypass. The Zion category associated with bypass is SGTR.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 10 Printed: 4/23/2015 For some source term groups, the contributing type of release is not completely dominated by one single category but a mixture of categories all representing the EPRI classes. Occasionally, other contributors (excluding the highest contributor) make up a sizeable portion of the composition. These other contributors occasionally are types of releases that would be classified differently than the highest release contributor. An example is source term group 172, where the highest contributor is Alpha (Class 7), with 52 percent of the release, while the second and third highest are associated with bypass failures (Class 8), combining for 37 percent of the release.. This group was ultimately classified as Class 7 because the Alpha release is considered the more severe type of release and was the highest contributor to the source term group. Using this information the Zion results are grouped to the EPRI classes. The grouping is presented in Table 5. Table 5 Assignment of Zion Source Term Groups to EPRI Classes EPRI Class Zion Source Term Groups Applied Average Exposure (person-rem) Class 1 301, 302 7.45E+2 Class 2 1, 31, 61, 64, 67, 100 5.97E+6 Class 7 33, 66, 69, 70, 72, 103, 105, 106, 136, 139, 142, 172, 175, 178, 303 1.55E+8 Class 8 2, 65, 68, 71, 101, 104,107, 137, 140, 143, 173, 176, 179 1.26E+8  EPRI's ILRT guidance document [2] utilizes a multiplication factor to develop the design basis leakage value (L a) that is based on generic information that provides comparative local population ratios. The WF3 population dose is adjusted for the local plant-specific population using a "population dose factor". The population dose factor is used to adjust the Zion population dose to account for differences in the local populations of the Zion and WF3 sites.
The population dose factor is calculated by dividing the WF3 population [9] by the Zion population information taken from the EPRI ILRT guidance document [2]. Total WF3 Population = 1,998,010 Zion Population = 4,439,288 Population Dose Factor = 0.45 The relationship above implies that the resultant doses are a direct function of population within fifty (50) miles of each site. This does not take into account differences in meteorology, environmental factors, containment designs or other factors but does provide a reasonable first-order approximation of the population dose as would be generated by the Zion accident sequences. While Zion had two release categories that fell into EPRI Class 1, a more accurate estimate for the INTACT dose rate at WF3 is developed using plant-specific data from Reference 10. The Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 11 Printed: 4/23/2015 INTACT dose is the basis for Class 3a and Class 3b doses, which are key in the ILRT delta-dose calculations. Therefore, using plant-specific information to develop the dose associated with INTACT yields results more reflective of the WF3 site. The method for developing the person-rem dose rate for the population within fifty (50) miles of WF3 utilizes a scaling factor. The dose rates for the exclusion area boundary (EAB) and the low population zone (LPZ) are used to define a distance scaling factor. This scaling factor is then used to estimate the dose for distances beyond the LPZ up to the fifty (50) mile radius. An average person-rem dose is predicted assuming a uniform distribution of radionuclides that decreases with increased distance from the origin. A uniform distribution of the surrounding population is then combined to calculate the final total dose. The analysis depends on inputs from the licensing basis analysis [10] to arrive at the EAB does rate, LPZ dose rate, LPZ total person-rem dose and population data [9]. The EAB is defined as the circular area within a radius of 914 meters (~0.57 miles) from the containment. The LPZ extends the radius to 3,300 meters (~2.05 miles) from the containment. Table 6 below presents the predicted dose rates for the EAB and LPZ two (2) hours after an event and the thirty (30) day LPZ dose. Table 6 Predicted Dose from Reference 10 Location Dose (rem)
EAB 2hr 4.11E+0 LPZ 2hr 6.28E-1 LPZ 30d 2.46E+0  The calculation of the necessary scaling factor is based on the relationship of dose rate and distance. The scaling equation is based on a ratio of the LPZ dose to EAB dose. The equation is presented below:          (eq. 1)
Where: Y = LPZ dose  X = EAB dose 
 
d LPZ = Distance for LPZ d EAB = Distance for EAB C = Scaling Constant This equation assumes that the dose rate is decreasing in a constant manner with distance and is consistent with the Comanche Peak ILRT submittal [11]. Solving the equation yields a value for the scaling constant (C). The input data is listed below in Table 7.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 12 Printed: 4/23/2015 Table 7 Calculation Parameters Solving for the Scaling Constant (C)
Parameter Value (units) X 4.11E+0 (rem)
Y 6.28E-1 (rem) dEAB 914 (meters) d LPZ 3300 (meters)
Solving Equation 1 with the inputs listed above yields a value of 1.46 for the scaling constant, C. Now the LPZ total dose data can be extrapolated to the fifty (50) mile radius dose criteria. Equation 1 is utilized again, but instead of solving for the scaling constant the equation is solved for fifty (50) mile radius dose. As the distance from the containment increases the so does the population surrounding the site, but the dose from an event also decreases with distance.
Consistent with Comanche Peak ILRT submittal, a value of twenty five (25) miles is used in the extrapolation to represent the average dose for the fifty (50) mile radius since it is the midpoint between the containment and the dose radius parameter. The values displayed in Table 8 are used in the same formula as Equation 1 to solve for the dose at twenty five (25) miles. Table 8 Calculation Parameters for the Dose at 25 Miles Parameter Value (units) X (LPZ 30d) 2.46E+0 (rem)
C 1.46 d LPZ 2.05 (miles) d 25 25 (miles)
  = 6.33E-2 (eq. 2) Solving Equation 2 with the inputs from Table 8 yields a value for the whole body dose of 6.33E-2 rem. This value represents an average individual dose. Now that the average person-rem dose rate for the fifty (50) mile radius zone is developed, the effect on the surrounding population is determined. The estimated population is 2.00E+6 persons. However, it is usually assumed that ninety five (95) percent of the population will be evacuated prior to the release such that only five (5) percent of the population would be involved [21]. Given a total population estimate of approximately 2.00E+6 people, this equates to an Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 13 Printed: 4/23/2015 exposed population of 9.99E+4 persons. The whole body dose multiplied by the estimated population exposed to a release yields a fifty (50) mile total population whole body dose of
 
6.33E+3 person-rem. Table 9 contains the release category dose information. Class 1 dose information is derived from a scaling factor based on plant specific data. Class 2, Class 7, and Class 8 are developed by multiplying the Zion dose for these classes, contained in Table 5, by the population dose factor. Class 6 applies a decontamination factor of 0.1 to the dose associated with Class 2 based on an assumption that 10 percent of the release would be scrubbed. Table 9 WF3 Dose for EPRI Accident Classes Release Category Frequency (/yr)
EPRI Class WF3 Dose  (person-rem) INTACT 1.57E-6 Class 1 6.33E+3 LERF 1 1.64E-8 Class 2 2.69E+6 SERF 2 1.76E-9 Class 6 2.69E+5 3 LERF + LATE 4 4.57E-6 Class 7 6.95E+7 LERF 5 4.98E-7 Class 8 5.66E+7 1. The EPRI Class 2 category consists of WF3 assigned LERF contribution associated with isolation failures as re-categorized in Table 3. 2. The EPRI Class 6 category consists of WF3 assigned scrubbed isolation failures in SERF. 3. The EPRI Class 6 Does rate is derived from the Class 2 does rate. A decontamination factor of 0.1 is applied with the assumption that 10 percent of the release would be scrubbed. 4. The EPRI Class 7 category consists of the WF3 assigned LERF contribution associated with phenomenological failures as re-categorized in Table 3. Additionally consistent with the EPRI guidance document, LATE failures are classified as Class 7. 5. The EPRI Class 8 category consists of the WF3 assigned LERF contribution associated with bypass or SGTR failures as re-categorized in Table 3.
3.0  ASSUMPTIONS 1. The maximum containment leakage for EPRI Class 1 sequences is 1 L a (Type A acceptable leakage) because a new Class 3 has been added to account for increased leakage due to Type A inspections [2]. 2. The maximum containment leakage for Class 3a sequences is 10 L a based on the EPRI guidance. 3. The maximum containment leakage for Class 3b sequences is 100 L a based on the NEI guidance contained within the EPRI report. 4. Class 3b is conservatively categorized as LERF based on the NEI guidance and previously approved EPRI methodology.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 14 Printed: 4/23/2015
: 5. Containment leakage due to EPRI Classes 4 and 5 are considered negligible based on the NEI guidance and the previously approved EPRI methodology. 6. The containment releases are constant and continuous and are not impacted with time. The duration of the release is defined by the LERF definition provided in the PRA. 7. The containment releases for EPRI Classes 2, 6, 7, and 8 are not impacted by the ILRT Type A Test frequency. These classes already include containment failure with release consequences equal or greater than those impacted by Type A. 8. Because EPRI Class 8 sequences are containment bypass sequences, potential releases are directly to the environment. Therefore, the containment structure will not impact the release magnitude. 9. The WF3 IF PRA model [23] is developed separately and was not assessed using the internal events Level 1 model [16]. Based on similar CDF scenarios and the relative independence of core damage and containment phenomenology [24], the PDS distribution is based on the internal transient events; additionally, the transient initiator is the only initiator assumed for internal flooding. This assignment provides for fifty (50) percent of the IF CDF contribution (1.24E-6/yr) to be binned as INTACT, while the
 
remaining fifty (50) percent (1.24E-6/yr) is binned as LATE.
4.0  CALCULATIONS This calculation applies the WF3 PRA release category information in terms of frequency and person-rem estimates to determine the changes in risk due to increasing the ILRT test interval.
The changes in risk are assessed consistent with the guidance provided in the EPRI guidance
 
document [2]. 4.1  CALCULATIONAL STEPS The analysis employs the steps provided in EPRI's ILRT guidance document and uses associated risk metrics to evaluate the impact of a proposed change on plant risk. These measures are the change in release frequency, the change in risk as defined by the change in person-rem, the change in LERF, and the change in the conditional containment failure probability (CCFP). Additionally EPRI also lists the change in CDF as a measure to be considered [2]. Since the testing addresses the ability of the containment to maintain its function, the proposed change has no measurable impact on core damage frequency. Therefore, this attribute remains constant and has no risk significance. The overall analysis process is documented as outlined below:  Define and quantify the baseline plant damage classes and person-rem estimates. Calculate baseline leakage rates and estimate probability to define the analysis baseline. Develop baseline population dose (person-rem) and population dose rate (person-rem/yr).
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 15 Printed: 4/23/2015 Modify Type A leakage estimate to address extension of the Type A test frequency and calculate new population dose rates, LERF and conditional containment failure probability. Compare analysis metrics to estimate the impact and significance of the increase related to those metrics. The first step in the analysis is to define the baseline plant damage classes and person-rem dose measures. Plant damage state information is developed using the WF3 PRA Level 2 PRA results [5]. The containment endstate information and the results of the containment analysis are used to define the representative sequences. The population person-rem dose estimates for the key plant damage classes are based on the application of the method described in the EPRI ILRT guidance document [2]. The product of the person-rem for the plant damage classes and the frequency of the plant damage state is used to estimate the annual person-rem for the particular plant damage state. Summing these estimates produces the annual person-rem dose based on the sequences defined in the WF3 PRA. The PRA plant damage state definitions considered isolation failures due to Type B and Type C faults and examined containment challenges occurring after core damage and/or reactor vessel failure. These sequences are grouped into key plant damage classes. Using the plant damage state information, bypass, isolation failures and phenomena-related containment failures are identified. Once identified, the sequence was then classified by the EPRI release category definitions. With this information developed, the PRA baseline inputs are completed. The second step expands the baseline model to address Type A leakage. The PRA did address Type A (liner-related) faults, represented by INTACT accident sequences, and this contribution has been binned into EPRI Class 1. A new estimation using the EPRI methodology must be incorporated to provide a complete baseline. In order to define leakage that can be linked directly to the Type A testing, it is important that only failures that would be identified by Type A testing exclusively be included. The EPRI ILRT guidance document [2] provides the estimate for the probability of a leakage contribution that could only be identified by Type A testing based on industry experience. This probability is then used to adjust the intact containment category of the WF3 PRA to develop a baseline model including Type A faults. The release, in terms of person-rem, is developed based on information contained in EPRI's report and is estimated as a leakage increase relative to allowable dose (L a) defined as part of the ILRT. The predicted probability of Type A leakage is then modified to address the expanded time between testing. This is accomplished by a ratio of the existing testing interval and the proposed test interval. This assumes a constant failure rate and that the failures are randomly dispersed during the interval between the test. The change due to the expanded interval is calculated and reported in terms of the change in release due to the expanded testing interval, the change in the population person-rem and the change in large early release frequency. The change in the conditional containment failure Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 16 Printed: 4/23/2015 probability is also developed. From these comparisons, a conclusion is drawn as to the risk significance of the proposed change. Using this process, the following were performed: 1. Map the WF3 release categories into the 8 release classes defined by the EPRI Report. 2. Calculate the Type A leakage estimate to define the analysis baseline.
: 3. Calculate the Type A leakage estimate to address the current testing frequency.
: 4. Modify the Type A leakage estimates to addr ess extension of the Type A test interval. 5. Calculate increase in risk due to extending Type A testing intervals.
: 6. Estimate the change in LERF due to the Type A testing. 7. Estimate the change in CCFP due to the Type A testing. 4.2  SUPPORTING CALCULATIONS Step 1: Map the release categories into the 8 release classes defined by the EPRI Report [2]  EPRI defines eight (8) release classes as presented in Table 10.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 17 Printed: 4/23/2015 Table 10 EPRI Containment Failure Classifications EPRI Failure Classification Description Interpretation for Assigning WF3 Release Category 1 Containment remains intact with containment initially isolated Intact containment bins or late basemat attack sequences. 2 Dependent failure modes or common cause failures Isolation faults that are related to a loss of power or other isolation failure mode that is not a direct failure of an isolation component 3 Independent containment isolation failures due to Type A related failures Isolation failures identified by Type A testing 4 Independent containment isolation failures due to Type B related failures Isolation failures identified by Type B testing 5 Independent containment isolation failures due to Type C related failures Isolation failures identified by Type C testing 6 Other penetration failures Isolation failure with scrubbing or small isolation fails 7 Induced by severe accident phenomena Early containment failure sequences as a result of hydrogen burn or other early phenomena 8 Bypass Bypass sequence or SGTR Table 11 presents the WF3 release category mapping for these eight accident classes. Person-rem per year is the product of the frequency (per year) and the person-rem.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 18 Printed: 4/23/2015 Table 11 WF3 PRA Release Category Grouping to EPRI Classes Class EPRI Description Frequency Person-Rem Person-Rem/yr 1 Intact containment 1.57E-6 6.33E+3 9.92E-3 2 Large containment isolation failures 1.64E-8 2.69E+6 4.42E-2 3a Small isolation failures (liner breach) To be Determined  0.00E+0 3b Large isolation failures (liner breach) To be Determined  0.00E+0 4 Small isolation failures - failure to seal (type B) -    5 Small isolation failures - failure to seal (type C) -    6 Containment isolation failures (dependent failure, personnel
 
errors) 1.76E-9 2.69E+5 4.74E-4 7 Severe accident phenomena-induced failure (early) 4.57E-6 6.95E+7 3.18E+2 8 Containment bypass 4.98E-7 5.66E+7 2.82E+1  Total 6.66E-6  3.46E+2 Step 2: Calculate the Type A leakage estimate to define the analysis baseline (3 year test interval) As displayed in Table 11, the WF3 PRA did not identify any release categories specifically associated with EPRI Classes 4 or 5 and the estimate for Class 3 was redistributed back into INTACT. Therefore each of these classes mu st be evaluated for applicability to this study.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 19 Printed: 4/23/2015 Class 3: Containment failures in this class are due to leaks such as liner breaches that could only be detected by performing a Type A ILRT. In order to determine the impact of the extended testing interval, the probability of Type A leakage must be calculated. In order to better assess the range of possible leakage rates, the Class 3 calculation is divided into two classes. Class 3a is defined as a small liner breach and Class 3b is defined as a large liner breach. This division is consistent with the EPRI methodology [2]. The calculation of Class 3a and Class 3b probabilities is presented below. Calculation of Class 3a Probability Data presented in the EPRI report [2] contains 2 Type A leakage events out of 217 tests. Using the data a mean estimate for the probability is determined for Class 3a as shown in Equation 3.  (eq. 3) This probability, however, is based on three tests over a ten (10) year period and not the one per ten-year frequency currently employed at WF3 [3]. The probability (0.0092) must be adjusted to reflect this difference and is adjusted in step 3 of this calculation. Multiplying the CDF times the probability of a Class 3a leak develops the Class 3a frequency contribution in accordance with guidance provided by EPRI. The total CDF includes contributions already binned to LERF. To include these contributions would result in a potentially conservative result. Therefore, the LERF contribution (Class 2 and Class 8) from CDF is removed (1.64E-8/yr and 4.98E-7/yr). The CDF for WF3 is 4.18E-6/yr as presented in Table 11 and is adjusted to remove the LERF contribution. Therefore the frequency of a Class 3a failure is calculated as:
FREQ class3a = PROBclass3a x (CDF- Class 2 - Class 8)  = 0.0092 x (6.66E-6/yr - 1.64E-8/yr - 4.98E-7/yr) = 5.66E-8/yr (eq. 4)
Calculation of Class 3b Probability To estimate the failure probability given that no failures have occurred, the guidance provided in the EPRI report [2] suggests the use of a non-informative prior. This approach essentially updates a uniform distribution (no bias) with the available evidence (data) to provide a better estimation of an event. A beta distribution is typically used for the uniform prior with the parameters =0.5 and =1. This is then combined with the existing data (no Class 3b events, 217 tests) using Equation 5.
0023.0 218 5.0 1 217 5.0 0 3 N n p b Class (eq. 5)
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 20 Printed: 4/23/2015 where: N is the number of tests, n is the number of events (faults) of interest,  are the parameters of the non-informative prior distribution. From this solution, the frequency for Class 3b is generated using Equation 6 and is adjusted appropriately to address LERF sequences.
FREQ class3b = PROBclass3b x (CDF - Class 8 - Class 2)  = 0.0023 x (6.66E-6/yr - 4.98E-7/yr - 1.64E-8/yr) = 1.41E-8/yr (eq. 6)
Class 1: Although the frequency of this class is not directly impacted by Type A testing and the frequency for Class 1 should be reduced by the estimated frequencies in the new Class 3a and Class 3b in order to preserve the total CDF. The revised Class 1 frequency is therefore:
FREQ class1 = FREQ class1 - (FREQ class3a + FREQclass3b)  (eq. 7)
FREQ class1 = 1.57E-6/yr - (5.66E-8/yr + 1.41E-8/yr) = 1.50E-6/yr Class 2: Class 2 represents large containment isolation failures. Class 2 contains contribution to LERF related to isolation failures without scrubbing credited. The frequency of Class 2 is the sum of those release categories identified in Table 3 as Class 2.
FREQ class2 = 1.64E-8/yr (eq. 8) Class 4: This group consists of all core damage accidents for which a failure-to-seal containment isolation failure of Type B test components occurs. By definition, these failures are dependent on Type B testing, and Type A testing will not impact the probability. Therefore this group is not
 
evaluated further, consistent with the approved methodology. Class 5: This group consists of all core damage accidents for which a failure-to-seal containment isolation failure of Type C test components occurs. By definition, these failures are dependent on Type C testing, and Type A testing will not impact the probability. Therefore this group is not evaluated further, consistent with the approved methodology. Class 6: The Class 6 group is comprised of isolation faults that occur as a result of the accident sequence progression. For WF3, this class is defined by the WF3 SERF category.
FREQclass6 = 1.76E-9/yr (eq. 9)
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 21 Printed: 4/23/2015 Class 7: Class 7 represents early and late containment failure sequences involving phenomena related containment breach. Class 7 contains contributions to LERF related to early release phenomena. The frequency of Class 7 is the sum of those release categories identified in Table 3 as Class 7 and the frequency associated with LATE failures.
FREQ class7 = 4.57E-6/yr (eq. 10) Class 8: The frequency of Class 8 is the sum of those release categories identified in Table 3 as Class 8.
 
FREQ class8 = 4.98E-7/yr (eq. 11) Table 12 summarizes the above information by the EPRI defined classes. This table also presents dose exposures previously calculated. Class 3a and 3b person-rem values are developed based on the design basis assessment of the intact containment as defined in the EPRI guidance report [2]. The Class 3a and 3b doses are represented as 10L a and 100L a respectively. Table 12 also presents the person-rem frequency data determined by multiplying the failure class frequency by the corresponding exposure.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 22 Printed: 4/23/2015 Table 12 Baseline Risk Profile Class Description Frequency (/yr) Person-rem Person-rem (/yr) 1 No Containment Failure 1.50E-6 6.33E+3 9.47E-3 2 Large Containment Isolation Failures 1.64E-8 2.69E+6 4.42E-2 3a Small Isolation Failures (Liner breach) 5.66E-8 6.33E+4 2 3.58E-3 3b Large Isolation Failures (Liner breach) 1.41E-8 6.33E+5 3 8.91E-3 4 Small isolation failures - failure to seal (type B)  1    5 Small isolation failures - failure to seal (type C)  1    6 Containment Isolation Failures (dependent failure, personnel errors) 1.76E-9 2.69E+5 4.74E-4 7 Severe Accident Phenomena- induced Failure (Early and
 
Late) 4.57E-6 6.95E+7 3.18E+2 8 Containment Bypass 4.98E-7 5.66E+7 2.82E+1  Total 6.66E-6  3.46E+2
: 1. represents a probabilistica lly insignificant value. 2. 10 times L
: a. 3. 100 times L
: a. The percent risk contribution due to Type A testing is defined as follows:
 
%RiskBASE =[( Class3aBASE + Class3bBASE) / TotalBASE] x 100 (eq. 12)
Where: Class3aBASE = Class 3a person-rem/yr for baseline interval = 3.58E-3 person-rem/yr Class3bBASE = Class 3b person-rem/yr for baseline interval = 8.91E-3 person-rem/yr TotalBASE = total person-rem/yr for baseline interval = 3.46E+2 person-rem/yr
%RiskBASE = [(3.58E-3 + 8.91E-3) / 3.46E+2] x 100 =
0.004 percent (eq. 13)
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 23 Printed: 4/23/2015 Step 3: Calculate the Type A leakage estimate to address the current inspection interval The current surveillance testing requirement for Type A testing and allowed by 10 CFR 50, Appendix J is at least once (1) per  ten (10) years based on an acceptable performance history (defined as two consecutive periodic Type A tests at least twenty four (24) months apart in which the calculated performance leakage was less than 1.0L a). According to the ERRI report [2], extending the Type A ILRT interval from three (3) in ten (10) years to one (1) in ten (10) years will increase the average time that a leak detectable only by an ILRT goes undetected from eighteen (18) to sixty (60) months. Multiplying the testing interval by 0.5 and multiplying by twelve (12) to convert from "years" to "months" calculates the average time for an undetected condition to exist. The increase for a ten (10) year ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (from  eighteen (18) months to sixty (60) months) multiplied by the existing Class 3a probability as shown in Equation 14. 0.0307 (eq. 14) A similar calculation is performed for the Class 3b probability as presented in Equation 15. 0.0077 (eq. 15)
Risk Impact due to ten (10) year Test Interval Based on the approved EPRI methodology [2] and the NEI guidance [4], the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences. The risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage. Additionally the Class 1 frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in
 
Table 13 below.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 24 Printed: 4/23/2015 Table 13 Risk Profile for Once in Ten Year Testing Class Description Frequency (/yr) Person-rem 2 Person-rem (/yr) 1 No Containment Failure 1 1.33E-6 6.33E+3 8.43E-3 2 Large Containment Isolation Failures 1.64E-8 2.69E+6 4.42E-2 3a Small Isolation Failures (Liner breach) 1.89E-7 6.33E+4 1.19E-2 3b Large Isolation Failures (Liner breach) 4.70E-8 6.33E+5 2.97E-2 4 Small isolation failures - failure to seal (type B)  3    5 Small isolation failures - failure to seal (type C)  3    6 Containment Isolation Failures (dependent failure, personnel errors) 1.76E-9 2.69E+5 4.74E-4 7 Severe Accident Phenomena- induced Failure (Early and
 
Late) 4.57E-6 6.95E+7 3.18E+2 8 Containment Bypass 4.98E-7 5.66E+7 2.82E+1  Total 6.66E-6  3.46E+2 1. The PRA frequency of Class 1 has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF. 2. From Table 12.
: 3. represents a probabilistically insignificant value. Using the same methods as for the baseline, and the data in Table 13 the percent risk contribution due to Type A testing is as follows:
%Risk 10 = [(Class3a 10 + Class3b
: 10) / Total 10] x 100 (eq. 16)
Where:
Class3a 10 = Class 3a person-rem/yr for current 10-year interval = 1.19E-2 person-rem/yr Class3b 10 = Class 3b person-rem/yr for current 10-year interval = 2.97E-2 person-rem/yr Total 10 = total person-rem/yr for current 10-year interval = 3.46E+2 person-rem/yr 
%Risk 10 = [(1.19E-2 + 2.97E-2) / 3.46E+2] x 100 =
0.01 percent (eq. 17)
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 25 Printed: 4/23/2015 The percent risk increase (%Risk 10) due to a ten (10) year ILRT over the baseline case is as follows: %Risk 10 = [((Class1 10 + Class3a 10-+ Class3b
: 10) - (Class1BASE + Class3aBASE + Class3bBASE))/ TotalBASE] x 100.0 (eq. 18)
Where:
Class1 10 = Class 1 person-rem/yr for current 10-year interval = 8.43E-3 person-rem/yr Class3a 10 = Class 3a person-rem/yr for current 10-year interval = 1.19E-2 person-rem/yr Class3b 10 = Class 3b person-rem/yr for current 10-year interval = 2.97E-2 person-rem/yr Class1BASE = Class 1 person-rem/yr for baseline interval = 9.47E-3 person-rem/yr (Table 12)
Class3aBASE = Class 3a person-rem/yr for baseline interval = 3.58E-3 person-rem/yr (Table 12)
Class3bBASE = Class 3b person-rem/yr for baseline interval = 8.91E-3 person-rem/yr (Table 12)
TotalBASE = total person-rem/yr for baseline interval = 3.46E+2 person-rem/yr (Table 12) %Risk 10 =  [(8.43E-3 + 1.19E-2 + 2.97E-2) - (9.47E-3 + 3.58E-3 + 8.91E-3)] / 3.46E+2 x 100.0
= 0.008 percent (eq. 19) Step 4: Calculate the Type A leakage estimate to address extended inspection intervals If the test interval is extended to one (1) per fifteen (15) years, the average time that a leak detectable only by an ILRT test goes undetected increases to ninety (90) months (0.5 x 15 x 12). For a fifteen (15) year test interval, the result is the ratio (90/18) of the exposure times as was the case for the 10 year case. Increasing the ILRT test interval from once (1) every three (3) years to once (1) per fifteen (15) years results in a proportional increase in the overall probability of leakage. The approach for developing the risk contribution for a fifteen (15) year interval is the same as that for the ten (10) year interval. The increase for a fifteen (15) year ILRT interval is the ratio of the average time for a failure to detect for the increased ILRT test interval (from eighteen (18) months to ninety (90) months) multiplied by the existing Class 3a probability as shown in Equation 20.  (eq. 20) A similar calculation is performed for the Class 3b probability as presented in Equation 21.  (eq. 21) Risk Impact due to 15-year Test Interval As stated for the ten (10) year case, the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 26 Printed: 4/23/2015 The increased risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage. Additionally the Class 1 frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in Table 14 below. Table 14 Risk Profile for Once in Fifteen Year Testing Class Description Frequency (/yr) Person-rem 2 Person-rem (/yr) 1 No Containment Failure 1 1.21E-6 6.33E+3 7.68E-3 2 Large Containment Isolation Failures 1.64E-8 2.69E+6 4.42E-2 3a Small Isolation Failures (Liner breach) 2.83E-7 6.33E+4 1.79E-2 3b Large Isolation Failures (Liner breach) 7.04E-8 6.33E+5 4.46E-2 4 Small isolation failures - failure to seal (type B)  3    5 Small isolation failures - failure to seal (type C)  3    6 Containment Isolation Failures (dependent failure, personnel errors) 1.76E-9 2.69E+5 4.74E-4 7 Severe Accident Phenomena- induced Failure (Early and Late) 4.57E-6 6.95E+7 3.18E+2 8 Containment Bypass 4.98E-7 5.66E+7 2.82E+1  Total 6.66E-6  3.46E+2 1. The PRA frequency of Class 1 has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF. 2. From Table 12.
: 3. represents a probabilistically insignificant value. Using the same methods as for the baseline, and the data in Table 14 the percent risk contribution due to Type A testing is as follows:
%Risk 15 =[( Class3a 15 + Class3b
: 15) / Total 15] x 100 (eq. 22)
Where: Class3a 15 = Class 3a person-rem/yr for 15-year interval = 1.79E-2 person-rem/yr Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 27 Printed: 4/23/2015 Class3b 15 = Class 3b person-rem/yr for 15-year interval = 4.46E-2 person-rem/yr Total 15 = total person-rem year for 15-year interval = 3.46E+2 person-rem/yr
%Risk 15 = [(1.79E-2 + 4.46E-2) / 3.46E+2] x 100 =
0.018 percent (eq. 23) The percent risk increase (%Risk 15) due to a fifteen-year ILRT over the baseline case is as follows: %Risk 15 = [((Class1 15 + Class3a 15 + Class3b 15) - (Class1BASE + Class3aBASE + Class3bBASE))/ TotalBASE] x 100.0 (eq. 24)
Where:
Class1 15 = Class 1 person-rem/yr for current 15-year interval = 7.68E-3 person-rem/yr Class3a 15 = Class 3a person-rem/yr for current 15-year interval = 1.79E-2 person-rem/yr Class3b 15 = Class 3b person-rem/yr for current 15-year interval = 4.46E-2 person-rem/yr Class1BASE = Class 1 person-rem/yr for baseline interval = 9.47E-3 person-rem/yr (Table 12)
Class3aBASE = Class 3a person-rem/yr for baseline interval = 3.58E-3 person-rem/yr (Table 12)
Class3bBASE = Class 3b person-rem/yr for baseline interval = 8.91E-3 person-rem/yr (Table 12)
TotalBASE = total person-rem/yr for baseline interval = 3.46E+2 person-rem/yr (Table 12) %Risk 15 = [(7.68E-3 + 1.79E-2 + 4.46E-2) - (9.47E-3 + 3.58E-3 + 8.91E-3)] / 3.46E+2 x 100.0
= 0.014 percent (eq. 25) Step 5: Calculate increase in risk due to extending Type A inspection intervals Based on the guidance in the EPRI guidance document [2], the percent increase in the total integrated plant risk from a fifteen-year ILRT over a current ten-year ILRT is computed as follows: %Total 10-15 = [((Class1 15 + Class3a 15 + Class3b 15) - (Class1 10 + Class3a 10 + Class3b 10))/ Total 10] x 100.0 (eq. 26) Where:
Class1 15 = Class 1 person-rem/yr for current 15-year interval = 7.68E-3 person-rem/yr Class3a 15 = Class 3a person-rem/yr for current 15-year interval = 1.79E-2 person-rem/yr Class3b 15 = Class 3b person-rem/yr for current 15-year interval = 4.46E-2 person-rem/yr Class1 10 = Class 1 person-rem/yr for current 10-year interval = 8.43E-3 person-rem/yr (Table 13)
Class3a 10 = Class 3a person-rem/yr for current 10-year interval = 1.19E-2 person-rem/yr  (Table 13)
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 28 Printed: 4/23/2015 Class3b 10 = Class 3b person-rem/yr for current 10-year interval = 2.97E-2 person-rem/yr  (Table 13)
Total 10 = total person-rem/yr for 10-year interval  = 3.46E+2 person-rem/yr (Table 13)
% Total 10-15 = [(7.68E-3 + 1.79E-2 + 4.46E-2) - (8.43E-3 + 1.19E-2 + 2.97E-2)] / 3.46E+2 x 100
= 0.006 percent (eq. 27) Step 6: Calculate the change in Risk in terms of Large Early Release Frequency (LERF) The risk impact associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from containment could in fact result in a larger release due to failure to detect a pre-existing leak during the relaxation period. From the EPRI Report, the Class 3a dose is assumed to be ten (10) times the intact containment leakage, L a (or 6.33E+4 person-rem) and the Class 3b dose is assumed to be 100 times L a (or 6.33E+5 person-rem). The method for defining the dose equivalent for allowable leakage (L a) is developed in the EPRI report. This compares to a historical observed average of twice L a. Therefore, the estimate is somewhat conservative. Based on EPRI guidance, only Class 3 sequences have the potential to result in large releases if a pre-existing leak were present. Class 1 sequences are not considered as potential large release pathways because for these sequences the containment remains intact. Therefore, the containment leak rate is expected to be small (less than 2L a). A larger leak rate would imply an impaired containment, such as Classes 2, 3, 6 and 7. Late releases are excluded regardless of the size of the leak because late releases are, by definition, not a LERF event. Therefore, the change in the frequency of Class 3b sequences is used as the increase in LERF for WF3, and the change in LERF can be determined by the differences. The EPRI guidance document [2] identifies that Class 3b is considered to be the main contributor to LERF. Table 15 summarizes the results of the LERF evaluation that Class 3b is indicative of a LERF sequence. Table 15 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Years 15 Years Class 3b (Type A LERF) 1.41E-8/yr 4.70E-8/yr 7.04E-8/yr LERF (3 year baseline)  3.29E-8/yr 5.64E-8/yr LERF (10 year baseline) 2.35E-8/yr Regulatory Guide 1.174 [6] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. The EPRI report [2] cites Regulatory Guide 1.174 and defines very small changes in risk as resulting in increases of CDF below 1E-6/yr and increases in LERF below 1E-7/yr. Since the ILRT does not impact CDF, the relevant metric is LERF.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 29 Printed: 4/23/2015 Calculating the increase in LERF requires determining the impact of the ILRT interval on the leakage probability. By increasing the ILRT interval from the currently acceptable ten (10) years to a period of fifteen (15) years results in an increase in the contribution to LERF of 2.35E-8/yr. This value meets the guidance in Regulatory Guide 1.174 defining very small changes in LERF. The LERF increase measured from the original three (3) in ten (10) year interval to the fifteen (15) year interval is 5.64E-8/yr, which is also less than the criterion presented in Regulatory Guide 1.174. Step 7: Calculate the change in Conditional Containment Failure Probability (CCFP) The conditional containment failure probability (CCFP) is defined as the probability of containment failure given the occurrence of an accident. This probability can be expressed using the following equation: CDF ncf f CCFP)(1 (eq. 28)
Where, f(ncf) is the frequency of those sequences which result in no containment failure. This frequency is determined by summing the Class 1 and Class 3a results, and CDF is the total frequency of all core damage sequences. Therefore the change in CCFP for this analysis is the CCFP using the results for fifteen (15) years (CCFP
: 15) minus the CCFP using the results for ten (10) years (CCFP 10). This can be expressed by the following:
10 15 15 10 CCFP CCFP CCFP (eq. 29) Using the data previously developed the change in CCFP from the current testing interval is calculated and presented in Table 16. Table 16 Impact on Conditional Containment Failure Probability due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Years 15 Years f(ncf) (/yr) 1.55E-6 1.52E-6 1.50E-6 f(ncf)/CDF 2.33E-1 2.29E-1 2.25E-1 CCFP 7.67E-1 7.71E-1 7.75E-1 CCFP (3 year baseline)  4.94E-3 8.47E-3 CCFP (10 year baseline) 3.53E-3
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 30 Printed: 4/23/2015 The EPRI guidance document [2] provides insight for determining acceptable levels of increase in CCFP. The guidance states that an increase in CCFP less than 1.5 percent is considered small based on past ILRT submittals accepted by the NRC. By increasing the ILRT interval from the currently acceptable ten (10) years to a period of fifteen (15) years results in a CCFP increase of 3.53E-3 or 0.46 percent. This value meets the guidance contained in the EPRI report for small changes in CCFP. The CCFP increase measured from the original three (3) in ten (10) year interval to the fifteen (15) year interval is 8.47E-3 or 1.10 percent, which is also less than the criterion presented in the guidance
 
document. 5.0  SENSITIVITY STUDIES This section presents sensitivity studies suggested in the EPRI report [2] for the WF3 ILRT extension assessment. This includes an evaluation of assumptions made in relation to liner corrosion, the use of the expert elicitation, and the impact of external events. 5.1  LINER CORROSION The analysis approach utilizes the Calvert Cliffs Nuclear Plant (CCNP) methodology [19] as modified by EPRI. This methodology is an acceptable approach to incorporate the liner corrosion issue into the integrated leak rate test (ILRT) extension risk evaluation, but more instances of corrosion have occurred since the EPRI report was published. Therefore the methodology used by CCNP and EPRI will remain unchanged, but the inputs will are updated using a data collection period that begins in September of 1996 and ends on December 31 st 2013. Thus the data collection period is extended from 5.5 years to 17.25 years. Over the 17.25 years, more containment liner corrosion events occurred. In 2011, the NRC published a technical letter report that analyzed containment liner corrosion events occurring at operating nuclear power plants in the USA [12]. The results of this analysis were five (5) containment liner corrosion events in almost fifteen (15) years at sixty six (66) possible sites in the Unites States. Two (2) of the five (5) events are the same existences of corrosion used by CCNP in their liner corrosion analysis (North Anna Power Station Unit 2 and Brunswick Steam Electric Plant Unit 2). The next event took place at D.C. Cook Unit 2 in March of 2001. A small hole was discovered in the liner plate that the plant suspected was man made. In 2009, a through-wall penetration caused by a piece of wood embedded in the concrete was identified at Beaver Valley. It should be noted that in 2006 during the Beaver Valley Unit 1 steam generator replacement surface corrosion was identified. This corrosion had yet to cause penetration in the liner, but since the discovery of this corrosion occurred during a steam generator replacement and not a normal inspection, the event will be included with the conservative assumption that the corrosion would have been discovered after it penetrated the steel liner. The last event occurred in the fall of 2013 at Beaver valley Unit 1 [13]. Thus over the 17.25 year data collection period six (6) liner corrosion events occurred at a possible sixty six (66) plant locations. Table 17 summarizes the results obtained from the CCNP methodology utilizing a more recent data collection period.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 31 Printed: 4/23/2015 Table 17 WF3 Liner Corrosion Risk Assessment Results Using CCNP Methodology Step Description Containment Cylinder and Dome (85%) Containment Basemat (15%) 1 Historical liner flaw likelihood Failure data:  containment location specific Success data:  based on 70 steel-lined containments and 5.5 years since the 10CFR 50.55a requirements of periodic visual inspections of containment surfaces Events  6 6 / (66 x 17.25) = 5.27E-3/yr Events:  0 Assume a half failure 0.5 / (66 x 17.25) = 4.39E-4/yr 2 Aged adjusted liner flaw likelihood During the 15-year interval, assume failure rate doubles every five years (14.9% increase per year). The average for the 5 th to 10 th year set to the historical failure rate.
Year 1 average 5-10 15 Failure rate 2.14E-3/yr 5.27E-3/yr
 
1.49E-2/yr Year 1 average 5-10 15 Failure rate 1.78E-4/yr 4.39E-4/yr
 
1.24E-3/yr 15 year average = 6.42E-3/yr 15 year average = 5.58E-4/yr 3 Increase in flaw likelihood between 3 and 15 years Uses aged adjusted liner flaw likelihood (Step 2),
assuming failure rate doubles every five years. 0.74% (1 to 3 years) 4.24% (1 to 10 years) 9.63% (1 to 15 years) 0.06% (1 to 3 years) 0.36% (1 to 10 years) 0.84% (1 to 15 years) 4 Likelihood of breach in containment given liner flaw 1% 0.1%
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 32 Printed: 4/23/2015 Table 17 (continued) WF3 Liner Corrosion Risk Assessment Results Using CCNP Methodology Step Description Containment Cylinder and Dome (85%) Containment Basemat (15%) 5 Visual inspection detection failure likelihood 10% 5% failure to identify visual flaws plus 5% likelihood that the flaw is not visible (not through-cylinder but could be detected by ILRT) All events have been detected through visual inspection. 5% visible failure detection is a conservative assumption. 100% Cannot be visually inspected 6 Likelihood of non-detected containment leakage (Steps 3 x 4 x 5) 0.00074% (3 years) 0.74% x 1% x 10%
0.00424% (10 years) 4.24% x 1% x 10% 0.00963% (15 years)
 
9.63% x 1% x 10% 0.00006% (3 years) 0.06% x 0.1% x 100%
0.00036% (10 years) 0.36% x 0.1% x 100% 0.00084% (15 years)
 
0.84% x 0.1% x 100%
The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for containment cylinder and dom e and the containment basemat. Total likelihood of non-detected containment leakage (3 yr) = 0.00074% + 0.00006% = 0.0008%
Total likelihood of non-detected containment leakage (10 yr) = 0.00424% + 0.00036% =
0.0046% Total likelihood of non-detected containment leakage (15 yr) = 0.00963% + 0.00084% =
0.01047% This likelihood is then multiplied by the non-LERF containment failures for WF3. This value is calculated by the following equation for each period of interest. LERF is comprised of Class 2, Class 8, and Class 3b cases as shown below in Equation 30. Non-LERF = CDF - Class 2 - Class 8 - Class 3b (eq. 30)
A final adjustment could be made to address cases with successful containment spray operation. It is conservatively not addressed as it would not be expected to substantially alter the overall results. Table 18 presents the data and the resultant increase in LERF due to liner corrosion for each case.
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 33 Printed: 4/23/2015 Table 18 Liner Corrosion LERF Adjustment Using CCNP Methodology Case CDF (/yr)
Class 2 (/yr) Class 8 (/yr) Class 3b (/yr) Likelihood of Non-detected Corrosion Leakage Increase in LERF (/yr) 3-years 6.66E-6 1.64E-8 4.98E-7 1.41E-8 8.00E-6 4.90E-11 10-years 6.66E-6 1.64E-8 4.98E-7 4.70E-8 4.60E-5 2.80E-10 15-years 6.66E-6 1.64E-8 4.98E-7 7.04E-8 1.05E-4 6.36E-10 This contribution is added to the Class 3b LERF cases and the sensitivity analysis performed. Table 19 provides a summary of the base case as well as the corrosion sensitivity case. The "Delta Person-Rem" column provides the change in person-rem between the case without corrosion and the case that considers corrosion. Values within parentheses "( )" indicate the LERF change or delta between the without corrosion and corrosion cases.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 34 Printed: 4/23/2015 Table 19 WF3 Summary of Base Case and Corrosion Sensitivity Cases EPRI Class Base Case (3 per 10 years) 1 per 10 years 1 per 15 years Without Corrosion With Corrosion Without Corrosion With Corrosion Without Corrosion With Corrosion Frequency Person-rem per year Frequency Person-rem per year Delta Person-Rem per year Frequency Person-rem per year Frequency Person-rem per year Delta Person-Rem per year Frequency Person-rem per year Frequency Person-rem per year Delta Person-Rem per year 1 1.50E-6 9.47E-3 1.50E-6 9.47E-3 -3.10E-7 1.33E-6 8.43E-3 1.33E-6 8.43E-3 -1.77E-6 1.21E-6 7.68E-3 1.21E-6 7.69E-3 -4.02E-6 2 1.64E-8 4.42E-2 1.64E-8 4.42E-2 N/A 1.64E-8 4.42E-2 1.64E-8 4.42E-2 N/A 1.64E-8 4.42E-2 1.64E-8 4.42E-2 N/A 3a 5.66E-8 3.58E-3 5.66E-8 3.58E-3 N/A 1.89E-7 1.19E-2 1.89E-7 1.19E-2 N/A 2.83E-7 1.79E-2 2.83E-7 1.79E-2 N/A 3b 1.41E-8 8.91E-3 1.41E-8 8.95E-3 3.10E-5 4.70E-8 2.97E-2 4.72E-8 2.99E-2 1.77E-4 7.04E-8 4.46E-2 7.11E-8 4.50E-2 4.02E-4 6 1.76E-9 4.74E-4 1.76E-9 4.74E-4 N/A 1.76E-9 4.74E-4 1.76E-9 4.74E-4 N/A 1.76E-9 4.74E-4 1.76E-9 4.74E-4 N/A 7 4.57E-6 3.18E+2 4.57E-6 3.18E+2 N/A 4.57E-6 3.18E+2 4.57E-6 3.18E+2 N/A 4.57E-6 3.18E+2 4.57E-6 3.18E+2 N/A 8 4.98E-7 2.82E+1 4.98E-7 2.82E+1 N/A 4.98E-7 2.82E+1 4.98E-7 2.82E+1 N/A 4.98E-7 2.82E+1 4.98E-7 2.82E+1 N/A CDF 6.66E-6 3.46E+2 6.66E-6 3.46E+2 3.07E-5 6.66E-6 3.46E+2 6.66E-6 3.46E+2 1.76E-4 6.66E-6 3.46E+2 6.66E-6 3.46E+2 3.98E-4 Class 3b LERF 1.41E-8 1.41E-8 (4.90E-11) 4.70E-8 4.72E-8 (2.80E-10) 7.04E-8 7.11E-8 (6.36E-10) Delta LERF (from base case of 3 per 10 years) 3.29E-8 3.31E-8 (2.31E-10) 5.64E-8 5.69E-8 (5.87E-10) Delta LERF from 1 per 10 years N/A 2.35E-8 2.38E-8 (3.55E-10)
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 35 Printed: 4/23/2015 The inclusion of corrosion does not result in an increase in LERF sufficient to invalidate the baseline analysis and the overall impact of corrosion inclusion is negligible. 5.2  DEFECT SENSITIVITY AND EXPERT ELICIATION SENSITIVITY A second sensitivity case on the impacts of assumptions regarding pre-existing containment defect or flaw probabilities of occurrence and magnitude, or size of the flaw, is performed as described in the EPRI guidance document [2]. The expert elicitation contained in the EPRI report developed probabilities for pre-existing containment defects that would be detected by the ILRT only based on the historical testing data. Using the expert knowledge, this information was extrapolated into a probability versus magnitude relationship for pre-existing containment defects. The failure mechanism analysis also used the historical ILRT data augmented with expert judgment to develop the results.
Details of the expert elicitation process and results are contained in the EPRI report. The expert elicitation process has the advantage of considering the available data for small leakage events, which have occurred in the data, and extrapolates those events and probabilities of occurrence to the potential for large magnitude leakage events. The expert elicitation results are used to develop sensitivity cases for the risk impact assessment. Employing the results requires the application of the ILRT interval methodology using the expert elicitation to change in the probability of pre-existing leakage in the
 
containment. The baseline assessment uses the Jefferys non-informative prior and the expert elicitation sensitivity study uses the results of the expert elicitation. In addition, given the relationship between leakage magnitude and probability, larger leakage that is more representative of large early release frequency, can be reflected. For the purposes of this sensitivity, the same leakage magnitudes that are used in the basic methodology (i.e., 10 La for small and 100 La for large) are used here. Table 20 presents the magnitudes and probabilities associated with the Jefferys non-informative prior and the expert elicitation use in the base methodology and this sensitivity case. Table 20 Comparison of Jefferys Non-Informative Prior and Expert Elicitation Values Leakage Size (L a) Jefferys Non-Informative Prior Expert Elicitation Mean Probability of Occurrence Percent Reduction 10 9.22E-3 3.88E-3 58% 100 2.29E-3 2.47E-4 89%
Taking the baseline analysis and using the values provided in Table 20 for the expert elicitation, the results in Table 21 are developed.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 36 Printed: 4/23/2015 Table 21 WF3 Summary of ILRT Extension Using Expert Elicitation Values Accident Class ILRT Interval 3 per 10 Years 1 per 10 years 1 per 15 Years Base Frequency Adjusted Base Frequency Dose (person-rem)
Dose Rate (person-rem/yr) Frequency Dose Rate (person-rem/yr) Frequency Dose Rate (person-rem/yr) 1 1.57E-6 1.54E-6 6.33E+3 9.76E-3 1.48E-6 9.39E-3 1.44E-6 9.12E-3 2 1.64E-8 1.64E-8 2.69E+6 4.42E-2 1.64E-8 4.42E-2 1.64E-8 4.42E-2 3a N/A 2.39E-8 6.33E+4 1.51E-3 7.97E-8 5.04E-3 1.19E-7 7.56E-3 3b N/A 1.52E-9 6.33E+5 9.63E-4 5.07E-9 3.21E-3 7.61E-9 4.81E-3 6 1.76E-9 1.76E-9 2.69E+5 4.74E-4 1.76E-9 4.74E-4 1.76E-9 4.74E-4 7 4.57E-6 4.57E-6 6.95E+7 3.18E+2 4.57E-6 3.18E+2 4.57E-6 3.18E+2 8 4.98E-7 4.98E-7 5.66E+7 2.82E+1 4.98E-7 2.82E+1 4.98E-7 2.82E+1 Totals 6.66E-6 6.66E-6 1.30E+8 3.46E+2 6.66E-6 3.46E+2 6.66E-6 3.46E+2  LERF (3 per 10 yrs base)  3.55E-9 6.08E-9  LERF (1 per 10  yrs base) 2.54E-9 CCFP 7.65E-1 7.65E-1 7.66E-1 The results illustrate how the expert elicitation reduces the overall change in LERF and the overall results are more favorable with regard to the change in risk. 5.3  POTENTIAL IMPACTS FROM EXTERNAL EVENTS An assessment of the impact of external events is performed. The primary basis for this investigation is the determination of the total LERF following an increase in the ILRT testing interval from three (3) in ten (10) years to one (1) in fifteen (15) years. External events were evaluated in the WF3 Individual Plant Examination of External Events (IPEEE) [14]. The IPEEE program was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and an understanding of severe accident risk. The primary areas of external event analysis for the WF3 IPEEE were seismic and internal fires, and other external events. Seismic and fire were considered to be the most limiting due to their frequency of occurrence and their potential impact on plant operability. Therefore it is assumed that they bound the risk contribution from other external events. Both seismic and internal fire were examined but the analysis contained conservative Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 37 Printed: 4/23/2015 assumptions related to consequential failures due to external events such that the absolute CDF is considered an understatement of plant performance and an over estimation of CDF. The WF3 site is a very low seismicity site and the potential for a seismic event of significance is very low relative to more active locations. Seismic events were addressed through a Seismic Margin Analysis (SMA) as part of the IPEEE for WF3. The Seismic PRA method screened all the components that met a high confidence low probability of failure (HCLPF) for the review level seismic event occurring with a magnitude of 0.3g. The remaining components were grouped together as a proxy component. It was assumed that if this proxy component failed it would result in core damage. This method is considered conservative. The SMA information is used in conjunction with the improvements that have been incorporated into the internal event model since the IPEEE was performed. Prior seismic analyses have indicated that for a well-designed plant, seismic contributions are a combination of low acceleration events with random failures and higher acceleration events with dependent component or structural failures due to forces associated with the seismic event. As cited in NUREG-1742 [15], the controlling failure typically involves prolonged loss of ac power leading to a station blackout. Low acceleration events lead to a disruption of offsite power sources and result in a prolonged need for onsite sources. This contribution has been estimated utilizing the current internal events analysis and based on the loss of offsite power (LOSP) initiating events analysis to define a conditional core damage probability (CCDP). This value is then combined with a typical estimation for the median capacity of the offsite power supply (0.3g, median capacity) [22]. The frequency is multiplied by 0.5 for the likelihood of failure of offsite sources given a seismic event. The CCDP is calculated by modifying the WF3 CAFTA model [16] to only calculate the CCDP associated with loss of offsite power scenarios. The model contains seven (7) unique initiating events (IEs) that are associated with LOSP. Since the impact of any of the seven (7) initiating events is the same, only one event (%T5) is set to a value of 1.0 to represent a condition reflecting a loss of offsite power and the quantification yields the CCDP due to LOSP. The quantification assumes that offsite power cannot be restored within twenty four (24) hours.
Since the standard recovery techniques utilize non-seismic data, it is not applicable. The calculated CCDP for SBO without recovery is 1.35E-2. From the seismic hazard curve [17], a 0.3g seismic event has a median frequency of 1.20E-5/yr. At this seismicity level, the best estimate fragility for loss of power yields a probability of 0.5. Combining the frequency, the CCDP and the probability of losing offsite power yields an estimate for the frequency contribution for low acceleration seismic events. The seismic CDF estimate assuming a 0.3g event is 8.07E-8/yr. In addition to the prolonged loss of offsite power case, at higher accelerations the seismic forces result in component and/or structural concerns. For most safety-related components, the structures are not limiting and the impact can be based on component-level fragility. Reference 22 utilized existing seismic fragility information to arrive at a generic estimate for component capacities. A review of this report indicates that major equipment exhibits at least 1.0g median capacity given standard assumptions related to anchorage and location. To develop an estimate for multiple component and/or structural seismic failures for WF3 a median capacity of 1.0g is utilized. The corresponding recurrence frequency of a seismic event of this acceleration or greater is 1.21E-6/yr. This is again multiplied by the probability of failures Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 38 Printed: 4/23/2015 at that acceleration (0.5) to arrive at a value of 6.07E-7/yr. This represents the frequency of core damage due to seismically-induced component and/or structural failures. This estimate is considered a bounding contribution for seismically induced failures, because the probability of a seismically induced component failure associated with a seismic event of this magnitude would dominate postulated random failure probability. A typical assumption of one-fails-all-fail typically assumed for seismic faults would also tend to defeat redundant components and again lead to the conclusion that for this seismicity range the seismic failures would provide a reasonable estimate for the contribution to core damage and LERF. Summing the estimates for lower acceleration seismic events which would be dominated by prolonged station blackout with the contribution from higher acceleration seismic events involving seismically induced component failures yields an estimated CDF contribution of 6.87E-7/yr (8.07E-8/yr + 6.07E-7/yr) and is controlled by higher acceleration seismic initiating event. The findings contained in NUREG-1742 [15] indicate that the fire CDF is primarily determined by plant transient type of events such as those from assessed plant transients. The judgment is made based on this observation that it is reasonable to assume that the ratio of intact to impaired containments will be similar for fire as for the internal events such that the total CDF and the breakdown by EPRI Class will be equivalent to that presented for the internal events. Since both fire and seismic are considered in this sensitivity study, the CDF contribution for fire is taken from the WF3 Fire PRA [18]. The value used in this study is the non-compliant fire risk evaluation CDF of 1.62E-5/yr. Per the guidance contained in the EPRI report [2] the figure-of-merit for the risk impact assessment of extended ILRT intervals is given as: delta LERF = The change in frequency of Accident Class 3b Using the percentage of total CDF contributing to LERF for the fire, seismic, and other external events as an approximation for the early CDF applicable to EPRI Accident Class 3b yields the following:
CDFFIRE = 1.62E-5/yr (eq. 31)
CDF SEISMIC = 8.07E-8/yr + 6.07E-7/yr = 6.87E-7/yr (eq. 32) Class 3b Frequency = [(CDFFIRE) + (CDF SEISMIC)]
* Class 3b Leakage Probability (eq. 33) Class 3b Frequency = [(1.62E-5/yr) + (6.87E-7/yr)]
* 2.3E-03 = 3.88E-8/yr (eq. 34) No adjustment is made to the CDF values since LERF sequences are typically associated with SGTR or interfacing system LOCA sequences which are not represented by the external event assessments. This is potentially conservative, but is reasonable based on the simplified assessment, the conservative nature of the external events studies and the fact that many of the external event scenarios are long term station blackout and long term level of analysis detail.
The change in LERF is estimated for the one (1) in ten (10) year and one (1) in fifteen (15) year cases and the change defined for the external events in Table 22.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 39 Printed: 4/23/2015 Table 22 WF3 Upper Bound External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 1 per 10 years) 3 per 10 year 1 per 10 year 1 per 15 year External Events 3.88E-8 1.29E-7 1.94E-7 6.47E-8 Internal Events 1.41E-8 4.70E-8 7.04E-8 2.35E-8 Combined 5.29E-8 1.76E-7 2.65E-7 8.82E-8 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined the total change in LERF does not exceed the guidance for very small change in risk and does not exceed the 1.0E-7/yr change in LERF. The LERF increase supports the conclusion that the increased duration between tests does not result in a significant change in risk and the increase is acceptable per the criterion defined in the EPRI guidance document [2].
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 40 Printed: 4/23/2015
 
==6.0  REFERENCES==
: 1. Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooling Power Reactors, U.S. Nuclear Regulatory Commission (USNRC), 10 CFR Part 50, Appendix J, January 2006. 2. Gisclon, J. M., et al, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325, Electric Power Research Institute, 1018234, October 2008. 3. Containment Integrated Leak Rate Test, Rev. 4 Change 9, Entergy Operations Incorporated, PE-005-001, August 2006. 4. Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, Nuclear Energy Institute, NEI 94-01, July 2012. 5. WF3 Large Early Release Frequency (LERF) Model, Rev. 1, Entergy Operations Incorporated, PRA-W3-01-001S12, June 2009. 6. An Approach for Using Probabilistic Risk Assessment in Risk-Informed decisions on Plant-Specific Changes to the Licensing Basis, USNRC, Regulatory Guide 1.174, July 1998. 7. Reactor Risk Reference Document, Appendices J-O, Draft for Comment, USNRC, NUREG-1150, January 1987. 8. Park, C. K., et al, Evaluation of Severe Accident Risks: Zion, Unit 1, Rev. 1, USNRC, NUREG/CR-4551, Vol. 7, March 1993. 9. Waterford 3 Emergency Plan: Revision 37, Entergy Operators Incorporated, February 2013. 10. Sicard, P., Loss of Coolant Accident (LOCA) Alternative Source Term (AST) Radiological Dose Consequences for 3716 MWt Extended Power Uprate (EPU), Entergy Operations Incorporated, ECS04-001, August 2004. 11. Summitt, R., Comanche Peak Steam Electric Station Probabilistic Safety Assessment Evaluation of Risk Significance of ILRT Extension, Reliability and Safety Consulting Engineers Inc., RSC 01-47/R&R-PN-110, November 2001. 12. Dunn, D. S., et al, Containment Liner Corrosion Operating Experience Summary Technical Letter Report - Revision 1, USNRC, August 2011. 13. Sepelak, B., Containment Liner Through Wall Defect Discovered During Planned Visual Inspection, FirstEnergy Nuclear Operating Company (FENOC), LER 2013-002-01, February 2014. 14. Underwood, D., IPEEE Reduced Scope Seismic Margins Assessment (SMA) Waterford 3, Entergy Operations Incorporated, WF3-CS-12-00001, February 2012. 15. Perspectives Gained from the IPEEE Program, USNRC, NUREG-1742, April 2002.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 41 Printed: 4/23/2015
: 16. WF3 at Power Fault Tree Model  Size  Date WF3Rev5.caf    407 KB 2/21/2013 11:07 am WF3Rev5.rr    6296 KB 7/15/2014 6:32 pm F-MASTR5.caf    3 KB  1/31/2013 10:07 am recovery_rules4.txt    27 KB  1/31/2013 10:08 am Mutex5.txt      28 KB  1/31/2013 6:31 pm COREDAMAGE.CUT    7294 KB 7/17/2014 1:48 pm 17. Entergy Seismic Hazard Curve  Size  Date Entergy USGShazard.xlsx  29 KB  4/7/2011 1:18 pm 18. Stephens, P., Comparison of Waterford 3 MOR and FRE CDF and LERF Results, Reliability and Safety Consulting Engineers, Inc., RSC-CALKNX-2013-0810, October 2013. 19. Letter to NRC from Calvert Clifffs Nuclear Power Plant Unit No.1. Docket No. 50-317, Response to Request for Additional Inform ation Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, dated March 27, 2002. 20. RSC ILRT Excel Calculations for Waterford 3  Size  Date ILRT Calculation Sheet WF3_R1.xlsx  104.5 KB 4/20/2015 21. Calculation of Reactor Accident Consequences: Appendix VI to Reactor Safety Study, USNRC, WASH-1400 (NUREG 75/014), October 1975. 22. Harrison, D., Generic Component Fragilities for the GE Advanced BWR Seismic Analysis, International Technology Corporation, International Technology Corporation, September 1988. 23. Allen, D., W3 Internal Flooding Analysis, Rev. 3, Entergy Operations Incorporated, PRA-W3-01-002. 24. Young, V., Large Early Release Frequency (LERF) & Level 2 Analysis, Rev. 1, Reliability and Safety Consulting (RSC) Engineers, Inc., RSC 13-12/PSA-WF3-01-LE, August 2014.
 
Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 42 Printed 04/23/15 Review Comments and Resolution Reviewer Directions: Provide detailed technical or global editorial comments here. Individual editorial or illustrative comments may be electronically provided (tracking) or attached to this review sheet. Resolution Process:
Originator must provide resolutions for all comments. Reviewer is to approve all proposed resolutions prior to completing the review process. No review is complete until this step is accomplished. Reviewer Comment Originator Resolution of Comment Reviewer Concurrence Editorial comments provided in markup. Updated report with all editorial changes. RS Page 3, the last bulleted item discusses "small" changes to CCFP but does not give any reference or actual baseline for comparison of what "small" is. Does such a metric exist? In addition, suggest adding a discussion after Table 16 related to the CCFP results similar to what exists for the delta LERF metrics. Added more information to the last bullet and after Table 16 detailing what the EPRI guidance document classifies as a small change in CCFP.
RS It would be beneficial to a casual reader if some items were defined early in the report. Suggest defining what Type A, B, C testing are, as well as what EPRI Class 1, 2, 3, etc are. Added a paragraph at the beginning of Section 1.1 that outlines the different type of containment leakage testing. The EPRI classifications are defined in Table 10 of the
 
report.
RS Table 6 header needs a reference filled in place of "XXX". Table 6 header title is now "Predicted Dose Rates from Reference 10" RS The short paragraph after Table 8 needs further explanation of how the calculation was performed as it is not possible to recreate it currently. Added in an equation and calculation to clear up how the INTACT dose was developed.
RS Should other noted assumptions throughout the report such as population evacuation levels be included in Section 3.0? These are the assumption that the EPRI guidance document sets for the user.
RS Consider moving some noted text from Section 4.1 into Section 1.1 to give more of the methodology up front. The current formatting is approved by the NRC and will remain unchanged.
RS Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 43 Printed: 4/23/2015 Equation #3 is not reproduced correctly in the supporting spreadsheet, updates are required which will slightly change the report's results. Update was made to the excel spreadsheet and the report and all subsequent calculation and numbers are updated.
RS Footnotes #2 and 3 in Table 12 are missing in the table text. Added the superscripts in the correct locations of the table.
RS Equations #13 and 14 require updates for the Class 3 probabilities which should match those presented in Equations #2 and 4 respectively. Similarly, Equations #19 and 20 require updates. All mentioned equations have been updated RS The Jefferys Prior column in Table 20 requires updates for the Class 3 probabilities which should match those presented in Equations #2 and 4. Any changes to the results presented in Tables 20 and 21 from this update should also be made. Tables 20 and 21 have been updated to reflect the correct Class 3 probabilities.
RS What is the source for the 0.5 probability of loss of offsite power given a seismic event as discussed in Section 5.3? The median capacity of LOSP is assumed to be 0.3g. Since this is a median capacity failure only occurs 50 percent of the time.
Therefore a 0.5 multiplier is applied to the probability.
RS Reference #18 is a duplicate of #1 and should be removed. Removed reference RS The dose constant equation (Eq. 1) appears to be the inverse of that in the documentation in Reference 11. Double checked the equation by hand against the reference the scaling factor is correct.
RS The values for X and Y should be dose not dose rate since they are for an accumulation of so many hours. See Reference 11 Appendix C. Change the values to be dose instead of dose rate. RS For Equation 19 and the like, you need to figure out some way to show that his is not zero because it is my by looking!  Need a footnote or additional precision, something. Change formatting from scientific to general number formatting with four significant digits.
RS Added suggested text to highlight that WF3 is a low seismicity area. Agreed and accepted the suggested text. RS Revision 1 1. Editorial comments. Updated report with all editorial changes. RS Evaluation of Risk Significance of an ILRT Extension RSC 14-12/ECS14-010 (Rev.1) 44 Printed: 4/23/2015
: 2. Section 2.0: I think it is good to be a bit more descriptive of how the 50/50 split was derived. Once that is done the Table 2 notes can be removed I believe. Additional wording for 50/50 split description added. RS 3. Please check the change made to Reference 4. No change required; the EPRI report is Reference 2 (Revision 2-A of 1009325). Reference 4 is the NEI ILRT extension task force document (Revision 3-A).
RS    : Sensitivity Cases File Name:  RSC-CALKNX-2015-0403.docx CALCULATION COVER SHEETPage  1 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calculation No: RSC-CALKNX-2015-0403 Revision: 0 Title:  ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results Facility:  Waterford Steam Electric Station Unit 3 Client: Entergy Project: RSC  15  -  03 Document Control Information This Calculation: Calculation Abstract and Search Keywords  Contains RSC Proprietary Information  Is a New Document  Supports,  Amends,  Supersedes,  Or Supplements RSC Document/Calculations(s):RSC 14-12/ECS14-010, Rev. 1 To provide an estimation of the change in risk associated with extending the Type A integrated leak rate test interval beyond the current 10 years specified by 10 CFR 50, Appendix J, Option B, using the internal flooding results and updated Level 2 results.
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Originator Reviewer Approver Electronic Signature Name Vincent Young Ricky Summitt Ricky Summitt Date 04  /  17  /  15 04  /  20  /  15 04  /  20  /  15 File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  2 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited.
PURPOSE OF ANALYSIS This sensitivity calculation provides a summary of the change in risk associated with extending the Type A integrated leak rate test interval beyond the current 10 years specified by 10 CFR 50, Appendix J, Option B 1 for Waterford Steam Electric Station Unit 3 (WF3), using the Level 2 probabilistic risk assessment (PRA) results from RSC 13-02/PSA-WF3-01-LE 2 and adding the internal flooding PRA results from PRA-W3-01-002
: 7. A second sensitivity is also performed, using the same Level 2 PRA results while doubling the internal flooding core damage frequency (CDF) contribution from PRA-W3-01-002. The assessment is consistent with the processes described in the methodology identified in EPRI's guidance document, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals
: 3. The complete ILRT extension risk analysis is provided in RSC 14-12/ECS14-010, Rev. 1
: 8. METHODOLOGY/APPROACH/PROCESS (define analysis steps) The reactor containment leakage test program consists of three tests: Type A, Type B, and Type C (Reference 1). These tests periodically verify the leak-tight integrity of the primary reactor containment, and the systems (and their components) penetrating the containment. Type A testing is intended to measure the overall integrated leak rate which is the summation of leakage through all potential leakage paths including containment welds, valves, fittings, and components which penetrate containment. The Type B test measures leakage across each pressure-containing or leakage-limiting boundary for a magnitude of containment penetration seals (i.e. resilient seals, gaskets, sealant compounds, flexible metal seal assemblies, air lock door seals, etc.). The final type of testing, Type C, measures containment isolation valve leakage rates. This type of testing is applicable for any valves that provide a direct connection between the inside and outside atmospheres of the primary reactor containment under normal operation, are required to close automatically upon receipt of a containment isolation signal, are required to operate intermittently under post-accident conditions, and are in main steam, feedwater, and other system piping which penetrate containment of direct-cycle boiling water power reactors. 10 CFR 50, Appendix J allows individual plants to extend Type A surveillance testing requirements and to provide for performance-based leak testing. This calculation documents a risk-based evaluation of the proposed change of the ILRT interval for the WF3, specifically using the WF3 internal flooding PRA results provided in PRA-W3-01-002 (Reference 7) and the WF3 Level 2 PRA results in RSC 13-02 (Reference 2). The proposed change would only impact testing associated with the current su rveillance tests for Type A l eakage, proce dure PE-005-001
: 4. This summary utilizes the guidelines set forth in NEI 94-01 5, the methodology used in the EPRI Report, and considers the submittals generated by other utilities. The complete ILRT extension risk analysis is provided in RSC 14-12/ECS14-010, Rev. 1 (Reference 8). ANALYSIS WORK AREA Sensitivity Case #1 (Updated Level 2 Results + Internal Flooding Results) - Summary The sensitivity results that combine the WF3 internal flooding model results (Reference 7) with the updated WF3 Level 2 PRA results (Reference 2) are provided in Table 1 below. Type A testing risk is comprised of EPRI Class 3a and Class 3b. Class 3b is defined as the large early release (LERF) contribution to Type A testing. Note that for this sensitivity case, the entire CDF contribution from the internal flooding PRA model (Reference 7) is split between the INTACT plant damage state (PDS) category and the LATE PDS category; 50 percent of the contribution (1.24E-6/yr) is binned as INTACT, while the remaining 50 percent (1.24E-6/yr) is binned as LATE.
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  3 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 1. Sensitivity #1: Summary of Risk Impact on Extending Type A ILRT Test Frequency Risk Impact for 3-years (baseline) Risk Impact for 10-years (current requirement) Risk Impact for 15-years Total integrated risk (person-rem/yr) 3.76E+2 3.76E+2 3.76E+2 Type A testing risk (person-rem/yr) 1.70E-2 5.68E-2 8.51E-2 % total risk  (Type A / total) 0.005% 0.015% 0.023% Type A LERF (Class 3b) (per year) 1.92E-8 6.40E-8 9.60E-8 Changes due to extension from 10 years (current)  Risk from current (Person-rem/yr) 2.74E-2 % Increase from current
( Risk / Total Risk) 0.007%  LERF from current (per year) 3.20E-8  CCFP from current 3.79E-3 Changes due to extension from 3 years (baseline)  Risk from baseline (Person-rem/yr) 6.57E-2 % Increase from baseline
( Risk / Total Risk) 0.017%  LERF from baseline (per year) 7.68E-8  CCFP from baseline 9.08E-3  The person-rem/year increase in risk contribution from extending the ILRT test frequency from the current ten (10) year interval to a fifteen (15) year interval is 2.74E-2 person-rem/year.
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  4 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. The risk increase in LERF from extending the ILRT test frequency from the current ten (10) year interval to a fifteen (15) year interval is 3.20E-8/yr. The change in conditional containment failure probability (CCFP) from the current ten (10) year interval to a fifteen (15) year interval is 3.79E-3/yr. The change in Type A test frequency from once (1) per ten (10) years to once (1) per fifteen (15) years increases the risk impact on the total integrated plant risk by only 0.007 percent. Also, the change in Type A test frequency from the original three (3) per ten (10) years to once (1) per fifteen (15) years increases the risk only 0.017 percent. Therefore, the risk impact when compared to other severe accident risks is negligible. Regulatory Guide 1.174 6 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10
-6/yr and increases in LERF below 10
-7/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from a once (1) per ten (10) years to once (1) per fifteen (15) years is 3.20E-8/yr. Guidance in Regulatory Guide 1.174 defines very small changes in LERF as below 10
-7/yr; therefore, increasing the ILRT interval from ten (10) to fifteen (15) years is considered non-risk significant, and the results support this determination. In addition, the change in LERF resulting from a change in the Type A ILRT test interval from a three (3) per ten (10) years to once (1) per fifteen (15) years is 7.68E-8/yr. The delta LERF is also below the guidance classification of a very small change. Regulatory Guide 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy. Consistency with defense-in-depth philosophy is maintained by demonstrating that the balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in conditional containment failure probability was estimated to be 3.79E-3 (0.57 percent increase) for the proposed change and 9.08E-3 (1.39 percent increase) for the cumulative change of going from a test interval of three (3) in ten (10) years to one (1) in fifteen (15) years. Both CCFP changes meet the criterion of less than 1.5 percent increase obtained from the EPRI guidance document (Reference 3). Therefore, the changes in CCFP are considered small and demonstrate that the defense-in-depth philosophy is maintained. In reviewing the results for sensitivity case #1, the WF3 analysis demonstrates that the change in plant risk is small as a result of this proposed extension of ILRT testing sensitivity. The change in LERF defined in the analysis for both the baseline and the current cases is within the acceptance criterion. Sensitivity Case #1 - Detailed Analysis The WF3 release states are summarized in Table 2. WF3 Level 2 results are grouped into four accident sequence states that represent the summation of individual accident categories. The number of sequences comprising each sequence state is also presented in Table 2.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  5 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 2. Release Category Frequencies Release Category Contributing WF3 Accident Categories Frequency (/yr) EPRI Classification INTACT (S) 1 10 2.94E-6 Class 1 LERF 2 18 8.25E-8 Class 8 SERF 9 8.87E-8 Class 6 LATE 3 14 5.34E-6 Class 7  Total N/A 8.45E-6 N/A 1. 50 percent of the CDF contribution from the internal flooding PRA model [(0.5) * (2.48E-6/yr)] was binned in the INTACT release category. 2. The LERF contribution for WF3 contains early containment failures due to containment phenomenon; per the EPRI guidance, these should be collected in Class 7. To accurately classify the contributions, the LERF contribution is separated to be consistent with the EPRI guidance document (Reference 3). 3. 50 percent of the CDF contribution from the internal flooding PRA model [(0.5) * (2.48E-6/yr)] was binned in the LATE release category. Table 3 contains the release category dose information. Class 1 dose information is derived from a scaling factor based on plant specific data. Class 2, Class 7, and Class 8 are developed by multiplying the Zion dose for these classes (Table 5 of Reference 8) by the population dose factor. Class 6 applies a decontamination factor of 0.1 to the dose associated with Class 2 based on an assumption that 10 percent of the release would be scrubbed. 
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  6 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 3. WF3 Dose for EPRI Accident Classes Release Category Frequency (/yr)
EPRI Class WF3 Dose (person-rem) INTACT 2.94E-6 Class 1 6.33E+3 LERF 1 5.81E-9 Class 2 2.69E+6 SERF 2 8.87E-8 Class 6 2.69E+5 3 LERF + LATE 4 5.34E-6 Class 7 6.95E+7 LERF 5 7.67E-8 Class 8 5.66E+7 1. The EPRI Class 2 category consists of the WF3 assigned LERF contribution associated with isolation failures (Table 24 of Reference 2). 2. The EPRI Class 6 category consists of WF3 assigned scrubbed isolation failures in SERF. 3. The EPRI Class 6 dose rate is derived from the Class 2 dose rate. A decontamination factor of 0.1 is applied with the assumption that 10 percent of the release would be scrubbed.
: 4. The EPRI Class 7 category consists of the WF3 assigned LERF contribution associated with phenomenological failures (Table 24 of Reference 2). Per the EPRI guidance document, LATE failures are also classified as Class 7. 5. The EPRI Class 8 category consists of the WF3 assigned LERF contribution associated with bypass or SGTR failures (Table 24 of Reference 2). Table 4 summarizes the information in Section 4.2 of Reference 8, by the EPRI-defined classes. This table also presents dose exposures calculated. Class 3a and 3b person-rem values are developed based on the design basis assessment of the intact containment as defined in the EPRI guidance report (Reference 3). The Class 3a and 3b doses are represented as 10L a and 100L a, respectively. Table 4 also presents the person-rem frequency data determined by multiplying the failure class frequency by the corresponding exposure.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  7 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 4. Baseline Risk Profile Class Description Frequency (/yr) Person-rem Person-rem (/yr) 1 No Containment Failure 2.84E-6 6.33E+3 1.80E-2 2 Large Containment Isolation Failures 5.81E-9 2.69E+6 1.56E-2 3a Small Isolation Failures (Liner breach) 7.71E-8 6.33E+4 2 4.88E-3 3b Large Isolation Failures (Liner breach) 1.92E-8 6.33E+5 3 1.21E-2 4 Small isolation failures - failure to seal (type B)  1    5 Small isolation failures - failure to seal (type C)  1    6 Containment Isolation Failures (dependent failure, personnel errors) 8.87E-8 2.69E+5 2.38E-2 7 Severe Accident Phenomena- induced Failure (Early and Late) 5.34E-6 6.95E+7 3.71E+2 8 Containment Bypass 7.67E-8 5.66E+7 4.35E+0  Total 8.45E-6  3.76E+2 1. Represents a probabilistically insignificant value. 2. 10 times L
: a. 3. 100 times L
: a. Based on the approved EPRI methodology (Reference 3) and the NEI guidance (Reference 5), the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences. The risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage. Additionally, the Class 1 frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in Table 5 below.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  8 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 5. Risk Profile for Once in Ten Year Testing Class Description Frequency (/yr) Person-rem 2 Person-rem (/yr) 1 No Containment Failure 1 2.62E-6 6.33E+3 1.66E-2 2 Large Containment Isolation Failures 5.81E-9 2.69E+6 1.56E-2 3a Small Isolation Failures (Liner breach) 2.57E-7 6.33E+4 1.63E-2 3b Large Isolation Failures (Liner breach) 6.40E-8 6.33E+5 4.05E-2 4 Small isolation failures - failure to seal (type B)  3    5 Small isolation failures - failure to seal (type C)  3    6 Containment Isolation Failures (dependent failure, personnel errors) 8.87E-8 2.69E+5 2.38E-2 7 Severe Accident Phenomena- induced Failure (Early and Late) 5.34E-6 6.95E+7 3.71E+2 8 Containment Bypass 7.67E-8 5.66E+7 4.35E+0  Total 8.45E-6  3.76E+2 1. The PRA frequency of Class 1 has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF. 2. From Table 4. 3. Represents a probabilistically insignificant value.
As stated for the ten (10) year case, the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences. The increased risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage. Additionally, the Class 1 frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in Table 6 below.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  9 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 6. Risk Profile for Once in Fifteen Year Testing Class Description Frequency (/yr) Person-rem 2 Person-rem (/yr) 1 No Containment Failure 1 2.46E-6 6.33E+3 1.55E-2 2 Large Containment Isolation Failures 5.81E-9 2.69E+6 1.56E-2 3a Small Isolation Failures (Liner breach) 3.86E-7 6.33E+4 2.44E-2 3b Large Isolation Failures (Liner breach) 9.60E-8 6.33E+5 6.07E-2 4 Small isolation failures - failure to seal (type B)  3    5 Small isolation failures - failure to seal (type C)  3    6 Containment Isolation Failures (dependent failure, personnel errors) 8.87E-8 2.69E+5 2.38E-2 7 Severe Accident Phenomena- induced Failure (Early and Late) 5.34E-6 6.95E+7 3.71E+2 8 Containment Bypass 7.67E-8 5.66E+7 4.35E+0  Total 8.45E-6  3.76E+2 1. The PRA frequency of Class 1 has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF. 2. From Table 4. 3. Represents a probabilistically insignificant value.
Sensitivity Case #2 (Updated Level 2 Results + Doubled Internal Flooding Results) - Summary The sensitivity results that combine the WF3 internal flooding model results (Reference 7) with the updated WF3 Level 2 PRA results (Reference 2) are provided in Table 7 below. Type A testing risk is comprised of EPRI Class 3a and Class 3b. Class 3b is defined as the LERF contribution to Type A testing. Note that for this sensitivity case, the entire CDF contribution from the internal flooding PRA model (Reference 7) is doubled, then split between the INTACT plant damage state (PDS) category and the LATE PDS category; 50 percent (2.48E-6/yr) is binned as INTACT, while the remaining 50 percent (2.48E-6/yr) is binned as LATE.
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  10 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 7. Summary of Risk Impact on Extending Type A ILRT Test Frequency Risk Impact for 3-years (baseline) Risk Impact for 10-years (current requirement) Risk Impact for 15-years Total integrated risk (person-rem/yr) 4.62E+2 4.62E+2 4.62E+2 Type A testing risk (person-rem/yr) 2.21E-2 7.36E-2 1.10E-1 % total risk  (Type A / total) 0.005% 0.016% 0.024% Type A LERF (Class 3b) (per year) 2.49E-8 8.29E-8 1.24E-7 Changes due to extension from 10 years (current)  Risk from current (Person-rem/yr) 3.55E-2 % Increase from current
( Risk / Total Risk) 0.008%  LERF from current (per year) 4.15E-8  CCFP from current 3.79E-3 Changes due to extension from 3 years (baseline)  Risk from baseline (Person-rem/yr) 8.51E-2 % Increase from baseline
( Risk / Total Risk) 0.018%  LERF from baseline (per year) 9.95E-8  CCFP from baseline 9.11E-3  The person-rem/year increase in risk contribution from extending the ILRT test frequency from the current ten (10) year interval to a fifteen (15) year interval is 3.55E-2 person-rem/year.
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  11 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. The risk increase in LERF from extending the ILRT test frequency from the current ten (10) year interval to a fifteen (15) year interval is 4.15E-8/yr. The change in conditional containment failure probability (CCFP) from the current ten (10) year interval to a fifteen (15) year interval is 3.79E-3/yr. The change in Type A test frequency from once (1) per ten (10) years to once (1) per fifteen (15) years increases the risk impact on the total integrated plant risk by only 0.008 percent. Also, the change in Type A test frequency from the original three (3) per ten (10) years to once (1) per fifteen (15) years increases the risk only 0.018 percent. Therefore, the risk impact when compared to other severe accident risks is negligible. Regulatory Guide 1.174 6 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10
-6/yr and increases in LERF below 10
-7/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from a once (1) per ten (10) years to once (1) per fifteen (15) years is 4.15E-8/yr. Guidance in Regulatory Guide 1.174 defines very small changes in LERF as below 10
-7/yr; therefore, increasing the ILRT interval from ten (10) to fifteen (15) years is considered non-risk significant, and the results support this determination. In addition, the change in LERF resulting from a change in the Type A ILRT test interval from a three (3) per ten (10) years to once (1) per fifteen (15) years is 9.95E-8/yr. The delta LERF is also below the guidance classification of a very small change. Regulatory Guide 1.174 also encourages the use of risk analysis techniques to help ensure and show that the proposed change is consistent with the defense-in-depth philosophy. Consistency with defense-in-depth philosophy is maintained by demonstrating that the balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation. The change in conditional containment failure probability was estimated to be 3.79E-3 (0.61 percent increase) for the proposed change and 9.11E-3 (1.47 percent increase) for the cumulative change of going from a test interval of three (3) in ten (10) years to one (1) in fifteen (15) years. Both CCFP changes meet the criterion of less than 1.5 percent increase obtained from the EPRI guidance document (Reference 3). Therefore, the changes in CCFP are considered small and demonstrate that the defense-in-depth philosophy is maintained. In reviewing the results for sensitivity case #2, the WF3 analysis demonstrates that the change in plant risk is small as a result of this proposed extension of ILRT testing sensitivity. The change in LERF defined in the analysis for both the baseline and the current cases is within the acceptance criterion. Sensitivity Case #2 - Detailed Analysis The WF3 release states are summarized in Table 8. WF3 Level 2 results are grouped into four accident sequence states that represent the summation of individual accident categories. The number of sequences comprising each sequence state is also presented in Table 8.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  12 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 8. Release Category Frequencies Release Category Contributing WF3 Accident Categories Frequency (/yr) EPRI Classification INTACT (S) 1 10 4.18E-6 Class 1 LERF 2 18 8.25E-8 Class 8 SERF 9 8.87E-8 Class 6 LATE 3 14 6.58E-6 Class 7  Total N/A 1.09E-5 N/A 1. 50 percent of the CDF contribution from the internal flooding PRA model was doubled [(0.5) * (2.48E-6/yr) * (2)] and binned in the INTACT release category.
: 2. The LERF contribution for WF3 contains early containment failures due to containment phenomenon; per the EPRI guidance, these should be collected in Class 7. To accurately classify the contributions, the LERF contribution is separated to be consistent with the EPRI guidance document (Reference 3).
: 3. 50 percent of the CDF contribution from the internal flooding PRA model was doubled [(0.5) * (2.48E-6/yr) * (2)] and binned in the LATE release category. Table 9 contains the release category dose information. Class 1 dose information is derived from a scaling factor based on plant specific data. Class 2, Class 7, and Class 8 are developed by multiplying the Zion dose for these classes (Table 5 of Reference 8) by the population dose factor. Class 6 applies a decontamination factor of 0.1 to the dose associated with Class 2 based on an assumption that 10 percent of the release would be scrubbed. 
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  13 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 9. WF3 Dose for EPRI Accident Classes Release Category Frequency (/yr)
EPRI Class WF3 Dose (person-rem) INTACT 4.18E-6 Class 1 6.33E+3 LERF 1 5.81E-9 Class 2 2.69E+6 SERF 2 8.87E-8 Class 6 2.69E+5 3 LERF + LATE 4 6.58E-6 Class 7 6.95E+7 LERF 5 7.67E-8 Class 8 5.66E+7 1. The EPRI Class 2 category consists of the WF3 assigned LERF contribution associated with isolation failures (Table 24 of Reference 2). 2. The EPRI Class 6 category consists of WF3 assigned scrubbed isolation failures in SERF. 3. The EPRI Class 6 dose rate is derived from the Class 2 dose rate. A decontamination factor of 0.1 is applied with the assumption that 10 percent of the release would be scrubbed.
: 4. The EPRI Class 7 category consists of the WF3 assigned LERF contribution associated with phenomenological failures (Table 24 of Reference 2). Per the EPRI guidance document, LATE failures are also classified as Class 7. 5. The EPRI Class 8 category consists of the WF3 assigned LERF contribution associated with bypass or SGTR failures (Table 24 of Reference 2). Table 10 summarizes the information in Section 4.2 of Reference 8, by the EPRI-defined classes. This table also presents dose exposures calculated. Class 3a and 3b person-rem values are developed based on the design basis assessment of the intact containment as defined in the EPRI guidance report (Reference 3). The Class 3a and 3b doses are represented as 10L a and 100L a, respectively. Table 10 also presents the person-rem frequency data determined by multiplying the failure class frequency by the corresponding exposure.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  14 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 10. Baseline Risk Profile Class Description Frequency (/yr) Person-rem Person-rem (/yr) 1 No Containment Failure 4.06E-6 6.33E+3 2.57E-2 2 Large Containment Isolation Failures 5.81E-9 2.69E+6 1.56E-2 3a Small Isolation Failures (Liner breach) 1.00E-7 6.33E+4 2 6.33E-3 3b Large Isolation Failures (Liner breach) 2.49E-8 6.33E+5 3 1.57E-2 4 Small isolation failures - failure to seal (type B)  1    5 Small isolation failures - failure to seal (type C)  1    6 Containment Isolation Failures (dependent failure, personnel errors) 8.87E-8 2.69E+5 2.38E-2 7 Severe Accident Phenomena- induced Failure (Early and Late) 6.58E-6 6.95E+7 4.58E+2 8 Containment Bypass 7.67E-8 5.66E+7 4.35E+0  Total 1.09E-5  4.62E+2 1. Represents a probabilistically insignificant value. 2. 10 times L
: a. 3. 100 times L
: a. Based on the approved EPRI methodology (Reference 3) and the NEI guidance (Reference 5), the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences. The risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage. Additionally, the Class 1 frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in Table 11 below.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  15 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 11. Risk Profile for Once in Ten Year Testing Class Description Frequency (/yr) Person-rem 2 Person-rem (/yr) 1 No Containment Failure 1 3.76E-6 6.33E+3 2.38E-2 2 Large Containment Isolation Failures 5.81E-9 2.69E+6 1.56E-2 3a Small Isolation Failures (Liner breach) 3.33E-7 6.33E+4 2.11E-2 3b Large Isolation Failures (Liner breach) 8.29E-8 6.33E+5 5.25E-2 4 Small isolation failures - failure to seal (type B)  3    5 Small isolation failures - failure to seal (type C)  3    6 Containment Isolation Failures (dependent failure, personnel errors) 8.87E-8 2.69E+5 2.38E-2 7 Severe Accident Phenomena- induced Failure (Early and Late) 6.58E-6 6.95E+7 4.58E+2 8 Containment Bypass 7.67E-8 5.66E+7 4.35E+0  Total 1.09E-5  4.62E+2 1. The PRA frequency of Class 1 has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF. 2. From Table 10. 3. Represents a probabilistically insignificant value.
As stated for the ten (10) year case, the increased probability of not detecting excessive leakage due to Type A tests directly impacts the frequency of the Class 3 sequences. The increased risk contribution is determined by multiplying the Class 3 accident frequency by the increase in the probability of leakage. Additionally, the Class 1 frequency is adjusted to maintain the overall core damage frequency constant. The results of this calculation are presented in Table 12 below.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  16 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. Table 12. Risk Profile for Once in Fifteen Year Testing Class Description Frequency (/yr) Person-rem 2 Person-rem (/yr) 1 No Containment Failure 1 3.56E-6 6.33E+3 2.25E-2 2 Large Containment Isolation Failures 5.81E-9 2.69E+6 1.56E-2 3a Small Isolation Failures (Liner breach) 5.00E-7 6.33E+4 3.16E-2 3b Large Isolation Failures (Liner breach) 1.24E-7 6.33E+5 7.87E-2 4 Small isolation failures - failure to seal (type B)  3    5 Small isolation failures - failure to seal (type C)  3    6 Containment Isolation Failures (dependent failure, personnel errors) 8.87E-8 2.69E+5 2.38E-2 7 Severe Accident Phenomena- induced Failure (Early and Late) 6.58E-6 6.95E+7 4.58E+2 8 Containment Bypass 7.67E-8 5.66E+7 4.35E+0  Total 1.09E-5  4.62E+2 1. The PRA frequency of Class 1 has been reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF. 2. From Table 10. 3. Represents a probabilistically insignificant value.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  17 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. REFERENCES 1. Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooling Power Reactors, U.S. Nuclear Regulatory Commission (USNRC), 10 CFR Part 50, Appendix J, January 2006. 2. Young, V., Large Early Release Frequency (LERF) & Level 2 Analysis, Rev. 1, Reliability and Safety Consulting (RSC) Engineers, Inc., RSC 13-02/PSA-WF3-01-LE, August 2014. 3. Gisclon, J. M., et al, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325, Electric Power Research Institute (EPRI), 1018243, October 2008. 4. Containment Integrated Leak Rate Test, Rev. 4 Change 9, Entergy, PE-005-001, August 2006. 5. Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, Nuclear Energy Institute (NEI), NEI 94-01, July 2012. 6. An Approach for Using Probabilistic Risk Assessment in Risk-Informed decisions on Plant-Specific Changes to the Licensing Basis, USNRC, Regulatory Guide 1.174, July 1998. 7. Allen, D., W3 Internal Flooding Analysis, Rev. 3, Entergy, PRA-W3-01-002. 8. Young, V., Evaluation of Risk Significance of an ILRT Extension, Rev. 1, RSC Engineers, Inc., RSC 14-12/ ECS14-010, April 2015.
 
File Name:  RSC-CALKNX-2015-0403.docx CALCULATION SHEETPage  18 of 18 PRINTED APRIL 23, 2015 File Saved: 4/23/2015 3:56:00 PM FORM  NO.: RSC-CALC99-02 Rev 25 Calc. No.: RSC-CALKNX-2015-0403 Revision: 0 Calculation Title: ILRT Sensitivity - Summary of Risk Impact on Extending A ILRT Test Frequency Using Internal Flooding Results and Updated Level 2 Results NOTICE: If so identified on the cover sheet, this document contains RSC Engineers, Inc. proprietary information. Any retention, duplication, reproduction, or distribution of this information either electronically or mechanically to parties either  internally or externally without the prior written consent of RSC Engineers, Inc. is prohibited. SCANS OR ATTACHMENTS (as necessary) toW3F1-2015-0021Revised Section 4.5.3 of License Amendment Request 4.5.3 Summary of Plant-Specific Risk Assessment ResultsThe findings of the WF3 risk assessment confirm the general findings of previous studies thatthe risk impact associated with extending the ILRT interval from three in ten years to one in15 years is small. The WF3 plant-specific results for extending ILRT interval from the current 10years to 15 years are summarized below. The person-rem/year increase in risk contribution from extending the ILRT testfrequency from the current ten (10) year interval to a fifteen (15) year interval is 2.01E-2person-rem/year. The risk increase in LERF from extending the ILRT test frequency from the current ten(10) year interval to a fifteen (15) year interval is 2.35E-8/yr. The change in conditional containment failure probability (CCFP) from the current ten(10) year interval to a fifteen (15) year interval is 3.53E-3/yr. The change in Type A test frequency from once (1) per ten (10) years to once (1) perfifteen (15) years increases the risk impact on the total integrated plant risk by only 0.006percent. Also, the change in Type A test frequency from the original three (3) per ten(10) years to once (1) per fifteen (15) years increases the risk only 0.014 percent.Therefore, the risk impact when compared to other severe accident risks is negligible. Regulatory Guide 1.174 [6] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very smallchanges in risk as resulting in increases of core damage frequency (CDF) below 10
-6/yrand increases in LERF below 10
-7/yr. Since the ILRT does not impact CDF, the relevantcriterion is LERF. The increase in LERF resulting from a change in the Type A ILRT testinterval from a once (1) per ten (10) years to once (1) per fifteen (15) years is 2.35E-8/yr.Guidance in Regulatory Guide 1.174 defines very small changes in LERF as below 10
-7/yr, increasing the ILRT interval from ten (10) to fifteen (15) years is thereforeconsidered non-risk significant and the results support this determination. In addition,the change in LERF resulting from a change in the Type A ILRT test interval from a three(3) per ten (10) years to once (1) per fifteen (15) years is 5.64E-8/yr. The delta LERF isalso below the guidance classification of a very small change. Regulatory Guide 1.174 also encourages the use of risk analysis techniques to helpensure and show that the proposed change is consistent with the defense-in-depthphilosophy. Consistency with defense-in-depth philosophy is maintained bydemonstrating that the balance is preserved among prevention of core damage,prevention of containment failure, and consequence mitigation. The change inconditional containment failure probability was estimated to be 3.53E-3 (0.46 percentincrease) for the proposed change and 8.47E-3 (1.10 percent increase) for thecumulative change of going from a test interval of three (3) in ten (10) years to one (1) infifteen (15) years. Both CCFP changes meet the criterion of less than 1.5 percentincrease obtained from the EPRI guidance document [2]. Therefore the changes inCCFP are considered small and demonstrate that the defense-in-depth philosophy ismaintained.In reviewing these results, the WF3 analysis demonstrates that the change in plant risk is smallas a result of this proposed extension of ILRT testing. The change in LERF defined in the analysis for both the baseline and the current cases is within the acceptance criterion.Details of the WF3 risk assessment are contained in Attachment 6 to this enclosure.}}

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