W3F1-2012-0005, Supplemental Information in Support of the NRC Acceptance Review of Waterford 3 License Amendment Request to Adopt NFPA 805

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Supplemental Information in Support of the NRC Acceptance Review of Waterford 3 License Amendment Request to Adopt NFPA 805
ML12027A049
Person / Time
Site: Waterford Entergy icon.png
Issue date: 01/26/2012
From: Jacobs D
Entergy Operations
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
W3F1-2012-0005
Download: ML12027A049 (16)
v • d • e


Text

Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6660 Fax 504-739-6698 djacob2@entergy.com Donna Jacobs Vice President - Operations Waterford 3 W3F1-2012-0005 January 26, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Supplemental Information in Support of the NRC Acceptance Review of Waterford 3 License Amendment Request to Adopt NFPA 805 Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1. Entergy letter dated November 17, 2011 License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), Waterford Steam Electric Station, Unit 3 (W3F1-2011-0074)
2. NRC Transmittal to Entergy dated January 12, 2012, Supplemental Information Needed for Acceptance of Requested Licensing Action Re:

License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generation Plants (2001 Edition) (TAC No. ME7602)

Dear Sir or Madam:

In letter dated November 17, 2011 (Reference 1), Entergy Operations, Inc. (Entergy) submitted a License Amendment Request (LAR) for Waterford Steam Electric Station, Unit 3 (Waterford 3) to adopt a new risk informed-performance based fire protection licensing basis under NFPA 805.

In accordance with NRC LIC-109, Acceptance Review Procedures, the NRC staff performed their review of the Waterford 3 LAR to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review. On January 12, 2012, a teleconference was conducted between members of Entergy and the NRC staff to discuss supplemental information needed to complete the Acceptance Review. Based on this teleconference, the NRC transmitted a request for supplemental information dated January 12, 2012 (Reference 2). Entergy is providing this supplement in response to the NRCs request. provides the responses to the NRCs request for supplemental information. contains Table V-2 that was omitted in the November 17, 2011 License Amendment Request. Attachment 3 contains a revised license condition page reflecting the modification implementation schedule.

W3F1-2012-0005 Page 2 Subsequent to submittal of the Waterford 3 NFPA 805 LAR, Entergy documented two recent findings in the November 17, 2011 Waterford 3 LAR that have been entered into the Entergy corrective action program. The first deals with a statement in Section 4.7.3 regarding referencing and revising the Entergy Quality Assurance Program Manual (QAPM) to capture NFPA 805 quality assurance requirements. A more appropriate programmatic control is being considered in lieu of the QAPM. The second deals with the failure sequence for Scenario TGB-T004 in LAR Table W-1. The referenced emergency diesel generator (EDG 3A-S) was conservatively assumed to fail. In further review, a short or open circuit is not expected to cause a loss of critical components, thus the EDG is expected to remain available. The resolutions to these findings are being addressed under the Entergy corrective action program.

Entergy does not believe that either of these findings would alter the conclusions of the NRC Acceptance Review under LIC-109.

There are no new commitments contained in this submittal.

If you require additional information, please contact William J. Steelman at 504-739-6685.

I declare under penalty of perjury that the foregoing is true and correct. Executed on January 26, 2012.

Sincerely, DJ/sab Attachments:

1. Supplemental Information in Support of NRC Acceptance Review for Waterford 3 NFPA 805 License Amendment Application
2. Waterford 3 Transition Report, Table V-2, Fire PRA Category 1 Summary
3. Revised Operating License Page

W3F1-2012-0005 Page 3 cc: Mr. Elmo E. Collins, Jr. RidsRgn4MailCenter@nrc.gov Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 E. Lamar Blvd.

Arlington, TX 76011-4511 NRC Senior Resident Inspector Marlone.Davis@nrc.gov Waterford Steam Electric Station Unit 3 Dean.Overland@nrc.gov P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Kaly.Kalyanam@nrc.gov Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001 Louisiana Department of Environmental Ji.Wiley@LA.gov Quality Office of Environmental Compliance Surveillance Division P.O. Box 4312 Baton Rouge, LA 70821-4312

Attachment 1 to W3F1-2012-0005 Supplemental Information in Support of NRC Acceptance Review for Waterford 3 NFPA 805 License Amendment Application to W3F1-2012-0005 Page 1 of 7 Supplemental Information in Support of NRC Acceptance Review for Waterford 3 NFPA 805 License Amendment Application Supplemental information was requested by the NRC Staff on January 12, 2012 in support of the Acceptance Review for Waterford Steam Electric Station, Unit 3 (Waterford 3) License Amendment Request (LAR) dated November 17, 2011. The following provides the Entergy supplemental information requested by the NRC staff.

1. A fire Probabilistic Risk Assessment (PRA), including the methods used in the base PRA, must have been peer reviewed to support an LAR. The NRC staff requests the LAR be supplemented confirming that the peer review was sufficient in that any enhancements subsequent to the peer review or ongoing enhancements to reduce Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) involved the review of fire impacts on component failures for important fire scenarios and were performed within the methodology reviewed by the peer review team.

Entergy Supplemental Information: As discussed in Attachment V of the November 17, 2011, Waterford 3 LAR, a Westinghouse Peer Review was performed for the Waterford 3 Fire PRA under report LTR-RAM-II-11-003, Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard. This review was a full scope peer review which concluded that the Waterford 3 Fire PRA methodologies used were appropriate and sufficient to satisfy the ASME/ANS PRA Standard RA-Sa-2009.

Enhancements subsequent to the peer review or ongoing enhancements to reduce CDF and LERF involved review of fire impact on component failures for important fire scenarios and were performed within the methodology reviewed by the peer review team.

2. An acceptable LAR should include Table V-2: the not-met and CC-I SRs for fires. Please provide this Table.

Entergy Supplemental Information: During the final production of the Waterford 3 LAR, Table V-2, Fire PRA-Category I Summary of Attachment V of the November 17, 2011 LAR was inadvertently omitted. This table (one page) is provided in Attachment 2 of this letter. Additionally, Entergy is providing a reference page that provides references associated with Tables V-1 and V-2. This page is also contained in Attachment 2 of this letter.

3. The LAR identifies five deviations from currently reviewed methods such as those in NUREG/6850 and the Frequently Asked Question (FAQs). The NRC staff has agreed that unreviewed methods may be included in current submittals if accompanied by a sensitivity study. The LAR provides sensitivity studies on each of the five but the synergistic effects are not explored. Rather than try to identify relevant combinations, it would be simplest to provide an estimate of the changes in CDF and LERF where all five deviations are simultaneously replaced with currently reviewed methods.

to W3F1-2012-0005 Page 2 of 7 Entergy Supplemental Information:

The Waterford 3 NUREG/CR-6850 alternate method sensitivity analysis is documented in PRA-W3-05-009, Revision 0, Sensitivity Evaluation to Address Alternate Methodologies.

There are five alternate methods applied for Waterford 3 as discussed in Section V.2 of the November 17, 2011 LAR. The following results are performed against the NUREG/CR-6850 approved methods using PRA-W3-05-009 including credit for proposed physical or procedural changes discussed in Section V.2. This approach provides a more conservative estimate of change in CDF and LERF than quantifying the currently EPRI approved methods. In order to report the aggregate change in CDF and LERF for the methods that deviate from NUREG/CR-6850, the effects are accounted for and included individually according to the relevant fire areas/zones, and in some cases, per the relevant fire scenarios.

Alternate Method #1: Reduced Heat Release Rate for Transient Combustible Fires This alternate method only affects the Turbine Generator Building (TGB) scenarios. Based on the markings to be provided in the TGB for more stringent transient combustible controls, the change in CDF for this method is reduced to 4.0E-07/yr from that previously reported in the November 17, 2011 LAR. It is expected that the affect on LERF would be no worse than the 10% change to the CDF (i.e. 10% of 3.82E-08/yr).

Risk Impact for Alternate Method 1 - Reduced HRR Fire Area Delta CDF Delta LERF TGB 4.0E-07/yr 3.82E-09/yr Alternate Method #2: Adjustment Factor for the Transient Combustible Fire Ignition Frequency Based on proposed procedural changes as listed in Attachment S of the November 17, 2011 LAR, this alternate method will affect those scenarios with transient fire modeled excluding RAB 5 and RAB 6 which have been mitigated by procedural controls. The model has been re-quantified for CDF and LERF without the use of this factor. Realistically, the factor or some portion thereof, is applicable, so this evaluation is conservative. The results of this evaluation are as follows:

Risk Impact for Alternate Method 2 - Fire Ignition Frequency Fire Area/Zone Delta CDF Delta LERF CTA 2.93E-10 4.15E-12 CTB 7.80E-11 1.54E-12 RAB 1B 1.50E-09 2.79E-11 RAB 1C 5.36E-10 1.06E-11 RAB 1D 2.87E-10 5.62E-12 RAB 2 6.52E-09 1.67E-10 RAB 3 3.76E-09 4.02E-11 RAB 7A 1.88E-10 2.72E-12 to W3F1-2012-0005 Page 3 of 7 Fire Area/Zone Delta CDF Delta LERF RAB 7B 1.56E-07 3.67E-09 RAB 8A 1.09E-08 2.67E-10 RAB 8B 7.62E-10 1.56E-11 RAB 8C 1.90E-07 4.11E-09 RAB 9 1.16E-09 2.26E-11 RAB 12 1.55E-09 3.86E-11 RAB 17 1.86E-10 3.60E-12 RAB 18 1.25E-09 1.37E-11 RAB 19 2.13E-10 4.19E-12 RAB 20 1.50E-09 3.82E-11 RAB 21 1.45E-10 2.82E-12 RAB 23 7.49E-09 1.89E-10 RAB 24 9.80E-10 1.06E-11 RAB 25 3.36E-10 6.55E-12 RAB 27 2.69E-10 4.29E-12 RAB 30 4.14E-11 6.75E-13 RAB 31 2.27E-10 5.52E-12 RAB 32 1.07E-10 2.04E-12 RAB 33 4.56E-11 8.91E-13 RAB 34 1.31E-10 2.56E-12 RAB 35 7.06E-11 1.38E-12 RAB 38 1.68E-10 3.28E-12 RAB 39 3.82E-07 1.04E-08 TGB 2.38E-07 1.98E-09 Total 1.01E-06 2.11E-08 Alternate Method #3: Adjustment Factor for the Hot Work Ignition Frequency A change in CDF was previously reported in V.2.3 of the November 17, 2011 LAR.

However, based on proposed procedural controls provided in Attachment S for fire areas RAB 5 and RAB 6, the primary affect from these scenarios is eliminated. Thus, there is no subsequent reported impact to CDF and LERF for this methodology.

Alternate Method #4: Adjustment Factor for Electrical Cabinet Ignition Frequency As discussed in Section V.2.4 of the November 17, 2011 LAR, this method has only negligible impact on the CDF and LERF. Therefore, there is no change in CDF or LERF being reported for this method.

to W3F1-2012-0005 Page 4 of 7 Alternate Method #5: Severity Factor for Severe Pump Oil Fires This method affects pump fire scenarios in fire areas RAB 2, RAB 35, RAB 36, RAB 39, and TGB. The most significant affect applies to scenarios E017M and E017S in RAB 2.

Based on the proposed modification in Attachment S for RAB 2 and the post-modification Fire Risk Evaluation for RAB 2, scenarios E017M and E017S will be eliminated, thus eliminating the majority of the CDF increase due to this method. The remaining effects of this method on the Waterford 3 Fire CDF, taken from PRA-W3-05-009, are as follows:

Alternate Method 5 - Severe Pump Oil Fire Affects Area Scenario Delta CDF Area Scenario Delta CDF RAB 2 E017 -1.19E-09 TGB E058 -1.20E-11 RAB 2 E018 -1.33E-09 TGB E058M 2.39E-11 RAB 2 E019 -5.66E-11 TGB E058S 9.75E-12 RAB 2 E019M 1.17E-09 TGB E062 -1.20E-11 RAB 2 Total -1.41E-09 TGB E062M 1.20E-11 RAB 35 E005 -1.20E-11 TGB E066 -1.20E-11 RAB 35 E005M 1.20E-11 TGB E066M 2.39E-11 RAB 35 E005S 3.25E-12 TGB E066S 9.75E-12 RAB 35 Total 3.25E-12 TGB E067 -1.20E-11 RAB 36 E004 -1.20E-11 TGB E067M 1.20E-11 RAB 36 E004M 1.19E-11 TGB E068 -1.20E-11 RAB 36 E004S 3.31E-12 TGB E068M 2.39E-11 RAB 36 Total 3.21E-12 TGB E072 -8.98E-12 RAB 39 E003 -1.20E-11 TGB E072M 2.39E-11 RAB 39 E003M 1.20E-11 TGB Total 7.01E-11 RAB 39 E003S 4.29E-12 RAB 39 Total 4.29E-12 The fire area delta CDF totals, as shown in the table, are three or more orders of magnitude below the total CDF. Therefore, the reported CDF and LERF values in Attachment W, Table W-2 are not significantly affected by this alternate method and no change in CDF or LERF is being subsequently reported.

Aggregate Impact on Fire Risk Results:

The aggregate change in CDF and LERF for the alternate methods applied at Waterford 3 is given in the table below. Due to planned modifications and changes, the impact from the NUREG/CR-6850 alternate methods applied at Waterford 3 has been reduced and satisfies RG 1.174 acceptance criteria.

to W3F1-2012-0005 Page 5 of 7 Aggregate Impact for Alternate Methods Fire Table W-2 Adjusted Table W-2 Adjusted Area/Zone Delta CDF Delta LERF CDF CDF LERF LERF CTA 2.93E-10 4.15E-12 1.37E-07 1.37E-07 2.43E-09 2.43E-09 CTB 7.80E-11 1.54E-12 1.30E-07 1.30E-07 2.35E-09 2.35E-09 RAB1B 1.50E-09 2.79E-11 4.85E-08 5.00E-08 9.33E-10 9.61E-10 RAB1C 5.36E-10 1.06E-11 1.24E-10 6.60E-10 2.42E-12 1.30E-11 RAB1D 2.87E-10 5.62E-12 9.65E-11 3.84E-10 1.89E-12 7.51E-12 RAB3 3.76E-09 4.02E-11 7.99E-09 1.17E-08 1.13E-10 1.53E-10 RAB7A 1.88E-10 2.72E-12 2.97E-06 2.97E-06 3.30E-08 3.30E-08 RAB7B 1.56E-07 3.67E-09 6.27E-07 7.83E-07 1.96E-08 2.33E-08 RAB8A 1.09E-08 2.67E-10 2.50E-06 2.51E-06 7.23E-08 7.26E-08 RAB8B 7.62E-10 1.56E-11 2.06E-06 2.06E-06 5.54E-08 5.54E-08 RAB8C 1.90E-07 4.11E-09 1.17E-05 1.19E-05 3.10E-07 3.14E-07 RAB9 1.16E-09 2.26E-11 1.52E-09 2.68E-09 2.98E-11 5.24E-11 RAB12 1.55E-09 3.86E-11 7.53E-09 9.08E-09 1.91E-10 2.30E-10 RAB17 1.86E-10 3.60E-12 1.26E-10 3.12E-10 2.46E-12 6.06E-12 RAB18 1.25E-09 1.37E-11 8.53E-10 2.10E-09 9.34E-12 2.30E-11 RAB19 2.13E-10 4.19E-12 7.27E-09 7.48E-09 2.65E-09 2.65E-09 RAB2 6.52E-09 1.67E-10 3.00E-06 3.01E-06 7.95E-08 7.97E-08 RAB20 1.50E-09 3.82E-11 3.83E-10 1.88E-09 8.49E-11 1.23E-10 RAB21 1.45E-10 2.82E-12 1.62E-07 1.62E-07 1.62E-09 1.62E-09 RAB23 7.49E-09 1.89E-10 5.97E-09 1.35E-08 1.34E-10 3.23E-10 RAB24 9.80E-10 1.06E-11 1.00E-08 1.10E-08 1.83E-10 1.94E-10 RAB25 3.36E-10 6.55E-12 5.44E-09 5.78E-09 1.12E-10 1.19E-10 RAB27 2.69E-10 4.29E-12 1.67E-08 1.70E-08 3.57E-10 3.61E-10 RAB30 4.14E-11 6.75E-13 1.22E-10 1.63E-10 9.80E-11 9.87E-11 RAB31 2.27E-10 5.52E-12 3.55E-08 3.57E-08 7.35E-10 7.41E-10 RAB32 1.07E-10 2.04E-12 1.68E-08 1.69E-08 3.42E-10 3.44E-10 RAB33 4.56E-11 8.91E-13 1.18E-09 1.23E-09 2.30E-11 2.39E-11 RAB34 1.31E-10 2.56E-12 8.94E-11 2.20E-10 1.75E-12 4.31E-12 RAB35 7.06E-11 1.38E-12 1.04E-08 1.05E-08 2.15E-10 2.16E-10 RAB38 1.68E-10 3.28E-12 2.36E-09 2.53E-09 4.07E-11 4.40E-11 RAB39 3.82E-07 1.04E-08 1.51E-07 5.33E-07 3.92E-09 1.44E-08 TGB 6.38E-07 5.8E-09 3.20E-06 3.84E-06 3.82E-08 4.40E-08 Total 1.41E-06 2.49E-08 Plant Total - All Fire Areas Table W-2 CDF Adjusted CDF Table W-2 LERF Adjusted LERF 3.4E-05 3.5E-05 7.3E-07 7.6E-07 to W3F1-2012-0005 Page 6 of 7

4. Waterford 3s increase in CDF is estimated to be greater than 10-6/yr. According to Regulatory Guide (RG) 1.174, the NRC staff needs an estimate of the total risk from all hazards, e.g., seismic and external flooding to review the acceptability of this result. The licensees estimate of the total risk should be provided.

Entergy Supplemental Information:

Section W.1 of the November 17, 2011 LAR provides a qualitative discussion of the affects from external events on site risk for the Waterford 3 site. The following provides further explanation and an estimate of the overall internal and external event site risk.

Site Risk from Internal Events The combined site risk from the fire PRA as reported in Table W-2 of the November 17, 2011 LAR is 3.4E-5/yr for CDF and 7.3 E-7/yr for LERF. The internal events PRA as provided in the Waterford 3 Level 1 Plant Safety Assessment (PSA) model is 4.0E-6/yr for CDF and 5.3E-7/yr for LERF for a combined site internal events CDF of 3.8E-5/yr and LERF of 1.3E-06/yr. Note that the evaluation models performed for the Fire PRA provide more conservative results than that performed for Level 1 PSA internal events models.

Site Risk from External Events Seismic - The Waterford site is within a region of infrequent and minor seismic activity.

Waterford 3 is designed for a maximum horizontal ground surface acceleration of 0.1 g.

Based on the updated seismic hazard curve provided by EPRI in November 2009, the likelihood of a seismic event exceeding 0.1 g peak ground acceleration is 1.5E-5/yr.

Because Waterford was designated a Reduced Scope Seismic plant in NUREG-1407, no detailed seismic fragilities for systems, structures, and components were required. Seismic walkdowns performed did not identify any seismic vulnerabilities or seismic design outliers.

Given the low seismic frequency with no seismic design outliers, the seismic CDF is estimated to be less than 1E-6/yr and LERF is estimated to be 1E-7/yr.

Flooding and other External Events - Other external events have been assessed for the Waterford 3 site for risk susceptibility. As discussed in Section W.1 of our November 17, 2011 LAR, high winds, floods, or off-site industry facility accidents do not contribute significantly to Waterford 3 site risk. For the external events that are within the acceptance criteria for the SRP (1975), the CDF is also estimated to be less than 1E-6/yr. This is consistent with the discussions of the events in Sections 2.3 through 2.11 of NUREG-1407.

In addition, Waterford has calculated the likelihood of core damage due to a toxic chemical release from nearby facilities and transportation accidents as 1.9E-9/yr.

Therefore, a bounding estimate of the overall CDF risk due to external events (including seismic, external flooding, and off-site industry facility accidents) is estimated to be less than 3E-6/yr. A total bounding estimate for LERF external events is assumed to be 0.1 of the total CDF which would be less than 3E-7/yr.

to W3F1-2012-0005 Page 7 of 7

5. Appendix M in the submittal states:

The licensee shall implement the following modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) by _____________. [INSERT DATE. To be based on Attachment S].

There are no following modifications listed. The licensee should clarify what modifications are planned.

Entergy Supplemental Information:

Attachment S, and in particular, Table S-1 provides the proposed modifications that are necessary to comply with the transition to NFPA 805 for Waterford 3. The language chosen was based on RG 1.205. The intent of the approach was to provide a final list of modifications based on the remaining modifications not implemented at time of license amendment issuance. However, Entergy proposes to modify Item (2) under the heading Transition License Conditions as reflected in Attachment M with the following statement.

(2) The licensee shall implement the modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) as provided in Table S-1, Plant Modifications of the Waterford 3 License Amendment Request dated November 17, 2011 prior to startup from the first refueling outage greater than 12 months following SER issuance.

A revised Page 8 of the License Condition previously provided in Enclosure 3 of the November 17, 2011 LAR is provided in Attachment 3 of this letter.

Attachment 2 to W3F1-2012-0005 Waterford 3 Transition Report, Table V-2, Fire PRA Category 1 Summary (2 pages) to 1 W3F1-2012-0005 Page 1 of 2 Entergy Attachment V - Fire PRA Quality Table V-2 Fire PRA- Category I Summary SR Topic Status FSS-B2 Main Control Room Abandonment The Waterford 3 Fire Probabilistic Risk Assessment Fire Scenarios Report (PRA-W3-05-006, Revision

0) Sections 13.2.1 and 13.2.2 discuss the use of a 0.1 CCDP for MCR abandonment. A detailed human reliability analysis for shutdown outside of the MCR does not exist. Shutdown outside the MCR was judged to result in a relatively high CCDP. This judgment was due to the combination of the HFEs, random failures associated with the remaining plant capability, and the damage to other systems by fire. The approach for applying a bounding 0.1 CCDP for MCR abandonment is judged to be appropriate for these cases.

A Capability Category 1 is considered acceptable for the Fire PRA application.

FSS-C1 Use of Multi-point Heat Release Rate Section 14 of the Fire Scenarios Report discusses the use of generic fire as opposed to more detailed Treatment fire modeling. Waterford 3 applied the Hughes Associates Generic Fire Modeling Treatment. This treatment offers a means for incorporation of fire modeling into the fire PRA in a manner that eliminates the need for separate scenario specific analyses. While the results are slightly more conservative, they are consistent with the results of the more detailed fire modeling.

A Capability Category I is acceptable for this application.

FSS-C2 Peak Heat Release Rates Versus Time The use of fire growth curves are not part of the Generic Fire Modeling Treatments used at Waterford Dependent Fire Growth. 3. Section 14 of the Fire Scenarios Report discusses the use of generic fire modeling versus detailed fire modeling and justifies the approach for the Waterford 3 Fire PRA application.

A Capability Category I is considered acceptable for this application.

FSS-C3 Fire Development - Burnout/Growth/Decay The use of fire growth curves are not part of the Generic Fire Modeling Treatments used at Waterford

3. Section 14 of the Fire Scenarios Report discusses the use of generic fire modeling versus detailed fire modeling and justifies the approach for the Waterford 3 Fire PRA application.

A Capability Category I is considered acceptable for this application.

FSS-E3 Quantitative Uncertainty Complete discussion of uncertainty and sensitivity is included in the Summary Report (PRA-W3 007, Revision 0). However, consistent with industry discussions with ACRS, quantitative treatment of uncertainty intervals for fire modeling parameters is not required at this time.

FSS-H2 Plant Specific Damage Criteria Plant specific damage thresholds were not developed for the Waterford 3 FPRA. This is in line with the application of the Hughes generic fire modeling approach (no detailed fire modeling was done).

A Capability Category I for FSS-H2 is considered acceptable for the FPRA application.

FSS-H5 Fire Scenario Uncertainty The Waterford 3 FPRA uses generic fire modeling for individual fire scenarios. With this approach, the Waterford 3 FPRA only meets Category I of SR FSS-H5. Section 14 of the Fire Scenarios Report (PRA-W3-05-006, Revision 0) provides the basis for this approach and the justification for its use in the FPRA methodology.

A Capability Category I for FSS-H5 is acceptable for the FPRA application.

to 2 W3F1-2012-0005 Page 2 of 2 Entergy Attachment V - Fire PRA Quality References

1. LTR-RAM-II-11-003, Fire PRA Peer Review of Waterford Steam Electric Station Unit 3 Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard
2. PRA-W3-05-003, Revision 0, WSES3 Fire Probabilistic Risk Assessment Quantification Model Preparation and Database Development
3. PRA-W3-05-006, Revision 0, WSES3 Fire Probabilistic Risk Assessment Fire Scenarios Report
4. PRA-W3-05-002, Revision 0, WSES3 Fire Probabilistic Risk Assessment Component and Cable Selection Report
5. PRA-W3-05-001, Revision 0, WSES3 Fire Probabilistic Risk Assessment Plant Partitioning and Fire Ignition Frequency Development
6. PRA-W3-05-007, Revision 0, WSES3 Fire Probabilistic Risk Assessment Summary Report
7. ASME/ANS RASa-2009, ASME and ANS Combined PRA Standard Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment

Attachment 3 to W3F1-2011-0005 Revised Operating License Page (1 page)

Transition License Conditions (1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.

(2) The licensee shall implement the modifications to its facility to complete the transition to full compliance with 10 CFR 50.48(c) as provided in Table S-1, "Plant Modifications" of the Waterford 3 License Amendment Request dated November 17, 2011 prior to startup from the first refueling outage greater than 12 months following SER issuance.

(3) The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.

10. Post-Fuel-Loading Initial Test Program (Section 14. SSER 10)

Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

11. Emergency Response Capabilities (Section 22. SSER 8)

EOI shall comply with the requirements of Supplement 1 to NUREG-0737 for the conduct of a Detailed Control Room Design Review (DCRDR).

Prior to May 1, 1985, the licensee shall submit for staff review and approval the DCRDR Summary Report, including a description of the process used in carrying out the function and task analysis performed as a part of both the DCRDR and the Procedures Generation Package efforts.

12. Reactor Coolant System (RCS) Depressurization Capabilitv (Section 5.4.3. SSER 8)

By June 18, 1985, the licensee shall submit the results of confirmatory tests regarding the depressurization capability of the auxiliary pressurizer spray (APS) system. This information must demonstrate that the APS system can perform the necessary depressurization to meet the steam generator single-tube rupture accident acceptance criteria (SRP 15.6.3) with loop charging isolation valve failed open. Should the test results fail to demonstrate that the acceptance criteria are met, the licensee must provide for staff review and approval, justification for interim operation, and a schedule for corrective actions.

AMENDMENT NO. 134,203,218, Re 0 i3eel bj letter elated Oetober 28, 2004 Revised bj letter elated dtJl, 26,2007

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