2CAN061406, Arkansas, Unit 2 - Response to Request for Additional Information Adoption of National Fire Protection Association Standard NFPA-805

Arkansas, Unit 2 - Response to Request for Additional Information Adoption of National Fire Protection Association Standard NFPA-805

ML14181B318
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/30/2014
From:	Jeremy G. Browning Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN061406
Download: ML14181B318 (12)

Text

s Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Jeremy G. Browning Vice President - Operations Arkansas Nuclear One 2CAN061406 June 30, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Response to Request for Additional Information Adoption of National Fire Protection Association Standard NFPA-805 Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

Dear Sir or Madam:

By letter dated June 9, 2014 (Reference 8), the NRC requested additional information (RAI) associated with the Entergy Operations, Inc. (Entergy) request to amend the Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specifications (TS) and licensing bases to comply with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205, Risk-Informed Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants. The amendment request followed Nuclear Energy Institute (NEI) 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program under 10 CFR 50.48(c). The submittal (Reference 1) described the methodology used to demonstrate compliance with, and transition to, National Fire Protection Association (NFPA)-805, and included regulatory evaluations, probabilistic risk assessment (PRA), change evaluations, proposed modifications for non-compliances, and supporting attachments.

The June 9, 2014, RAI required a response to all questions except PRA 01.e.01 and PRA 21 by July 9, 2014. Attachment 1 includes the Entergy responses to the applicable Reference 8 questions.

Changes or additional information, as detailed in this letter, with respect to the original Entergy request (Reference 1) have been reviewed and Entergy has determined that the changes do not invalidate the no significant hazards consideration included in the Reference 1 letter.

In accordance with 10 CFR 50.91(b)(1), a copy of this application and the reasoned analysis about no significant hazards consideration is being provided to the designated Arkansas state official.

This letter contains no new commitments.

2CAN061406 Page 2 of 3 If you have any questions or require additional information, please contact Stephenie Pyle at 479-858-4704.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on June 30, 2014.

Sincerely, ORIGINAL SIGNED BY JEREMY G. BROWNING JGB/dbb

Attachment:

Responses to Request for Additional Information - ANO-2 Transition to NFPA-805

REFERENCES:

1. Entergy letter dated December 17, 2012, License Amendment Request to Adopt NFPA-805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) (2CAN121202)

(ML12353A041)

2. NRC letter dated September 11, 2013, Arkansas Nuclear One, Unit 2 -

Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA-805 (TAC No. MF0404)

(2CNA091301) (ML13235A005)

3. Entergy letter dated November 7, 2013, Response to Request for Additional Information - Adoption of National Fire Protection Association Standard NFPA-805 (2CAN111301) (ML13312A877)

4. Entergy letter dated December 4, 2013, Response to Request for Additional Information - Adoption of National Fire Protection Association Standard NFPA-805 (2CAN121302) (ML13338A432)

5. Entergy letter dated January 6, 2014, Response to Request for Additional Information - Adoption of National Fire Protection Association Standard NFPA-805 (2CAN011401) (ML14006A315)

6. NRC letter dated March 28, 2014, Arkansas Nuclear One, Unit 2 -

Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA-805 (2CNA031401)

(ML14085A225)

7. Entergy letter dated May 22, 2014, Response to Request for Additional Information - Adoption of National Fire Protection Association Standard NFPA-805 (2CAN051404) (ML14142A410)

8. NRC letter dated June 9, 2014, Arkansas Nuclear One, Unit 2 - Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA-805 (2CNA061402) (ML14155A133)

2CAN061406 Page 3 of 3 cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Peter Bamford MS O-8B3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Attachment to 2CAN061406 Responses to Request for Additional Information ANO-2 Transition to NFPA-805

Attachment to 2CAN061406 Page 1 of 8 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ANO-2 Transition to NFPA-805 By letter dated June 9, 2014 (Reference 8), the NRC requested additional information (RAI) associated with the Entergy Operations, Inc. (Entergy) request (Reference 1) to transition the Arkansas Nuclear One, Unit 2 (ANO-2), fire protection licensing basis to National Fire Protection Association (NFPA) Standard NFPA-805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition). The June 9, 2014, RAI required a response to all questions except PRA 01.e.01 and PRA 21 by July 9, 2014. The following include the responses to the subject questions that were designated as due by July 9, 2014.

The respective question is included for convenience.

Probabilistic Risk Assessment (PRA) RAI 01.f.01 In a letter dated November 7, 2013, the licensee responded to PRA RAI 01.f stating that the combustible controls procedure will specify "levels of control" but only identifies one level (i.e.,

Level 1 - Continuous fire watch for unattended transient combustibles). Requiring a continuous fire watch allows crediting rapid detection and suppression but does not justify a reduction in the expected heat release rate (HRR) of a transient fire.

It is also not clear from the response if the 69 kilowatt (kW) transient fires are assumed only for the listed locations which all have Level1 control, and the 317 kW transient fires are assumed for all other locations with other control levels. Please describe the controls that the control of combustibles procedure (EN-DC-161) will implement and how will these controls specifically address the types and locations of existing and potential transient combustibles in the areas where a reduced heat release rate is credited.

Response

Procedure EN-DC-161, Control of Combustibles, defines Combustible Control Zones as A defined plant area established to indicate levels of controls required based on potential fire related damage to plant components. There are four levels identified with the most stringent being Level 1, where transient combustibles are prohibited unless evaluated and approved in accordance with the guidance in EN-DC-161. Levels 2, 3, and 4 provide fewer restrictions than required for a Level 1 area.

The use of a 69 kW fire size was limited to only the listed locations (2109-U, 2111-T, 2096-M, 2098-L, 2112-BB, 2154-E, and 2183-J) which are all identified as Level 1 in EN-DC-161. The 317 kW transient fire size was assumed for all other locations.

EN-DC-161 notes Level 1 areas in NFPA-805 plants are also considered a Zero Transient Combustible zone. No combustible material may be left unattended in these areas without a transient combustible evaluation (TCE). The TCE may require storage of combustibles in a metal storage container, fire extinguishers, continuous fire watches, hot work limitations, staging combustibles away from ignition sources, or other limitations. This procedure references the ANO combustible loading calculations which provide limitations on what may be allowed to remain in these areas unattended. The combustible loading calculations and EN-DC-161 for the seven fire zones where a 69 kW fire size is credited will be revised to specifically state the unique concerns and special needs for these sensitive areas. Plant drawings are being developed to identify ignition sources in these zero transient combustible fire zones to support

Attachment to 2CAN061406 Page 2 of 8 consistent TCE performance. This will ensure the placement of transient combustibles will not be impacted by identified ignition sources. These drawing and document updates will occur as part of the NFPA-805 implementation related to the revision of technical documents commitment in LAR Table S2, Item S2-7.

PRA RAI 06.01 In a letter dated November 7, 2013, the licensee responded to PRA RAI 06 and stated that fire ignition frequencies were updated from the NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," frequencies (as part of the compensation for removing control power transformers (CPT) credit) to frequencies provided in Supplement 1 to NUREG/CR-6850 (frequently asked question (FAQ) 08-0048). Please indicate whether or not the guidance in FAQ 08-0048 is being followed. The NRC staff notes that if following FAQ 08-0048, a sensitivity study using the mean of the fire frequency for those fire frequency bins with an alpha value less than or equal to one should be performed and, if the acceptance guidelines are exceeded, possible defense-in-depth actions should be considered. Address the possibility that the acceptance guidelines could be exceeded in the sensitivity study and, if this is the case, identify any applicable defense-in-depth actions.

Response

The guidance provided in FAQ 08-0048 is being followed. As recommended, Entergy evaluated the sensitivity of the risk and delta-risk results using the mean fire bin ignition frequencies from EPRI 1011989, NUREG/CR-6850 (pre-Supplement 1 ignition frequencies) for those bins with an alpha term less than or equal to one. This evaluation was documented in Attachment 1 of PRA-A2-05-010, Rev. 0, ANO2 Revised Baseline and Sensitivity Analysis. The conclusion of the evaluation was that ANO-2 was within the Regulatory Guide (RG) 1.174 acceptance criteria for a Region II plant with total CDF below 1E-04 and total LERF below 1E-05 for overall plant risk, and CDF less than 1E-05 and LERF less than 1E-06, without credit for the revised ignition frequencies. The results met the acceptance criteria; therefore, no additional defense-in-depth actions were necessary.

Since Entergy is retaining the NUREG/CR-6850, Supplement 1, ignition frequencies in the model which resolves other RAI concerns, this sensitivity evaluation will be updated to determine if the risk metrics are still met using the pre-Supplement 1 ignition frequencies for those bins with an alpha term of less than or equal to one. Entergy does not expect the acceptance criteria to be exceeded in this sensitivity evaluation; but if exceeded, the analysis will address defense-in-depth as recommended in FAQ 08-0048. The updated evaluation will be provided as part of the response to PRA RAI 21.

PRA RAI 06.02 In a letter dated November 7, 2013, the licensee responded to PRA RAI 06.b and stated that non-suppression probabilities were updated and were applied to scenarios 2109-U-B/C/D/E/F and G, based on additional fire modeling. The staff notes that these appear to be changes related to additional modelling discussed in the response to PRA RAI 01.g.ii. The staff notes that the response does not refer to the discussion in LAR Section V.2.3 that states that non-suppression probabilities of "0" were used for manual suppression times greater than

Attachment to 2CAN061406 Page 3 of 8 60 minutes, contrary to the NUREG/CR-6850 guidelines of a minimum value of 1E-3. Please confirm that minimum non-suppression probabilities of 1E-3 are used in the PRA and will be used unless replaced by an alternative method acceptable to the NRC Staff.

Response

Minimum non-suppression probabilities of 1E-03 are used in the PRA.

ANO-2 LAR, Attachment V,Section V.2 explains that the model of record at the time of the LAR contained unapproved analysis methods (UAMs) and that a sensitivity analysis had been performed removing those UAMs. Sub-section V.2.3 indicates that minimum non-suppression probabilities of 0 (zero) were used in the LAR model of record, but were changed to 1.0E-03 in the Sensitivity Analysis supporting the revised LAR.

As indicated in the response to PRA RAI 06 (ADAMs Accession No. ML13312A877), the sensitivity analysis changes have been fully incorporated in the PRA model and included in the documentation for the new baseline model. The risk results (CDF, CDF, LERF, LERF) that will be provided in the response to PRA RAI 21.c will reflect these changes.

PRA RAI 08.01 In a letter dated December 4, 2013, the licensee responded to PRA RAI 08.c and stated that the calculation of the control room abandonment frequency assumes that half of the control room panel fires will be single cable bundle fires and the other half will be multiple cable bundle fires.

The licensee further states that this assumption is considered appropriate and provides a more realistic yet conservative result. The licensee further stated that it is highly implausible for a fire to start in multiple cable bundles, rather, it will first start as a single cable bundle fire with the potential to grow to involve multiple cable bundles and that since the control room is continuously manned, the rapid response to the fire will suppress most fires during the initial, single cable bundle stage and the fire will not become a multiple cable bundle fire. The NRC staff notes that without supporting fire modeling or event data, it is not clear what the opportunity for suppression is before multiple cables in a multiple bundle are involved in a fire, and whether the assumption is conservative compared to, for example, identifying single and multiple cable cabinets and modelling suppression. The NRC staff also notes that this assumption is a deviation from NUREG/CR-6850, which treats individual electrical panels as either a single cable bundle or a multiple cable bundle. Please provide updated risk results as part of the integrated analysis requested in PRA RAI 21, modeling all panels in the MCR as multiple bundle cable fires, or provide further justification that the assumption is conservative based on characterization of the actual cable bundle configurations in the MCR cabinets.

Response

The assumption that half of the control room panels are single cable bundle panels while the other half are multiple cable bundle panels will be eliminated from the control room abandonment frequency analysis.

A walk-down to identify the cabling configurations within each of the control room panels has been performed and the cable configurations will be reviewed against the guidance in NUREG/CR-6850, Appendix G, to determine which panels are single cable bundle panels and

Attachment to 2CAN061406 Page 4 of 8 which are multiple cable bundle panels. Any cabinets that were not visually inspected will be assumed to contain multi-cable bundles. Thus, except for those panels confirmed by walk-down to be single bundle panels, all panels in the MCR will be modeled as multiple bundle cable fires.

The risk results (CDF, CDF, LERF, LERF) that will be provided in the response to PRA RAI 21.c will reflect this change.

PRA RAI 08.02 In a letter dated December 4, 2013, the licensee responded to PRA RAI 08.a and stated that two main control board (MCB) scenarios were evaluated: one in which the fire impacted Panels 2C01, 2C02, 2C03, 2C04, and 2C100, and one in which the fire impacted Panel 2C09. The licensee's analysis indicates that NUREG/CR-6850 Appendix L was used to develop a non-suppression probability for the MCB fire scenarios and that the MCB (Bin 4) fire frequency was apportioned between these scenarios. According to NUREG/CR-6850 Appendix L, if Figure L-1 is used then each MCB cabinet scenario should have the entire MCB frequency assigned to it since the Appendix L method assumes that the single 60' by 10' MCB cabinet represents the entire frequency of all MCB cabinets [alternatively, if the MCB (Bin 4) frequency is apportioned to individual MCB cabinets, then Figure L-1 should be re-calculated for each cabinet to use the actual dimensions of each cabinet]. Please provide updated risk results as part of the integrated analysis requested in PRA RAI 21 appropriately applying NUREG/CR-6850 Appendix L, or fully define and justify an alternative method.

Response

The frequency for the MCB (Scenario ID: 2199-G-B) is being revised to combine the frequency and cable targets of all MCB panels (2C01, 2C02, 2C03, 2C04, 2C09, and 2C100) into one bounding scenario (2199-G-B). The scenario will implement the non-suppression probability associated with a zero separation distance from Figure L-1 of NUREG/CR-6850, Appendix L.

This scenario will incorporate the total Bin 4 ignition frequency for the single MCB scenario and will also include all cable (and associated component failures) impacts from all panels that make up the MCB.

The risk results (CDF, CDF, LERF, LERF) that will be provided in the response to PRA RAI 21.c will reflect this change.

PRA RAI 09.01 In a letter dated November 7, 2013, the licensee responded to PRA RAI 09 and explained how the licensee's "multiplier approach" was applied to internal events HEPs to calculate the increased probability of fire related Human Failure Events (HFEs). The licensee also presented an evaluation of how this treatment aligns with NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines," showing the licensee's "multiplier approach" cannot be considered consistent with or conservative in respect to approaches defined in NUREG-1921.

In the response to PRA RAI 09, under Section 4.0, "Implementation in NFPA-805 Model," the licensee indicates that HFE evaluations will be supplemented by significance determination, reanalysis of significant HFE, and a sensitivity study using these reanalyzed HEPs as part of transition - but that the current "multiplier approach" will be retained if the risk profile is

Attachment to 2CAN061406 Page 5 of 8 unaffected. The NRC staff notes that if the risk profile is unaffected for transition, it does not necessarily follow that the risk profile will be unaffected for self-approval which will evaluate as yet unknown changes and for which the acceptance guidelines are relatively small. Please provide additional information about possible impact's from retaining the method in the base line model for use in self-approval evaluations and whether the sensitivity study will be retained for self-approval.

Response

The HRA analysis has eliminated the use of the multiplier approach and is now consistent with the detailed approach specified in NUREG-1921 for all HEPs. This is documented in Entergy calculations PRA-A2-05-007 and PRA-A2-05-020. The NUREG-1921 HEPs will be retained in the model for use in self-approval.

The risk results (CDF, CDF, LERF, LERF) that will be provided in the response to PRA RAI 21.c will reflect this change. In addition, since the HRA method was changed from the multiplier method to the NUREG-1921 method, a focused scope peer review was performed and an update to Attachment V of the LAR will be provided with the PRA RAI 21.c response.

PRA RAI 16.01 In a letter dated November 7, 2013, the licensee responded to PRA RAI 15.c and stated that, "due to lack of cable routing information, some components are assumed to be failed in all fire scenarios, unless credited by exclusion." The response to PRA RAI 02 states that "credit by exclusion" was used to remove just 2,757 failures from a total population of 451,276 cable failure events.

The assumption that numerous systems and/or functions always fail with every fire because the routing of the cables is not known is not necessarily a conservative assumption with respect to the change in risk estimate. Risk reduction modifications that reduce the risk of scenarios with conservatory high risk estimates will overestimate the magnitude of the actual risk reduction that will be achieved.

The NRC staff notes that there are several substantive risk reduction modifications being implemented as part of the transition to NFPA-805 and recognizes that such modifications are a real safety benefit while tracing cables is not. In a letter dated January 6, 2014, several sensitivity results reported in the response to PRA RAI 16 indicate that the effects of the assumption of system and/or function failures based on lack of knowledge about cable routing may be investigated instead of tracing cables. This investigation would need to support a conclusion that any increase in risk is smaller than the acceptance guidelines.

Please provide quantitative results (i.e., CDF, LERF, CDF, and LERF values) of an assessment that demonstrate that the magnitude of the risk reduction from the improvements being implemented to reduce risk is greater than the magnitude of the risk of the retained VFDRs. This assessment should explicitly investigate and account for the overestimation in risk associated with the assumption that cables with unknown routing fail in all fire scenarios. The impact of the assumption on the additional risk of recovery actions should also be addressed.

Attachment to 2CAN061406 Page 6 of 8

Response

As indicated in the response to PRA RAI 16 (ADAMS Accession No. ML14006A315), Table 1, the CDF/LERF increase associated with the VFDRs and recovery actions is 1.36E-05 for CDF and 2.25E-07 for LERF. The total change in risk (combined risk increases and risk decreases) is -2.62E-04 for CDF and -9.04E-06 for LERF. The risk reduction of the new common feedwater (CFW) pump (formerly referred to as the new AFW pump) overwhelms the risk increase of the remaining VFDRs and recovery actions resulting in a negative delta risk.

Thus, the magnitude of the risk reduction from the improvements being implemented to reduce risk is greater than the magnitude of the risk of the retained VFDRs and recovery actions.

UNL (Unknown Location) Sensitivity Study II provided in Table 2 of PRA RAI 16 represents the worst-case impact as all UNL components are set to nominal non-fire impacted probabilities in the compliant case and failed in the post-transition variant case. This sensitivity shows that, without the unknown routing assumption, the total risk reduction for CDF/LERF is -1.42E-04 for CDF and -5.11E-06 for LERF. Thus, without the unknown routing assumption, the risk reduction of the new CFW pump still overwhelms the risk increase of the remaining VFDRs and recovery actions resulting in a negative delta risk. Also, the magnitude of the risk reduction from the improvements being implemented to reduce risk is still greater than the magnitude of the risk of the retained VFDRs and recovery actions. Thus, the safety benefit from the significant NFPA-805 modifications is the result of the proposed modification and not due to unknown cable routing assumptions.

As stated in the response to RAI PRA 16, these values do not reflect the post RAI model, and will be updated accordingly.

PRA RAI 17.a.01 In a letter dated January 6, 2014, the licensee responded to PRA RAI 17.a and stated that modifications may directly or indirectly correct a VFDR. The use of the term "indirectly" is confusing. A modification (or assumed modification) that protects or moves a cable directly resolves a VFDR and the lack of such a modification is modeled by removing fire affect failures from the compliant plant. If a modification provides an alternative redundant train, this may be called an indirect correction; but if categorized as related to a VFDR, the new train should be modeled in both the compliant plant and the post-transition plant. If categorized as a risk reducing modification, it may be modeled in only the post-transition plant. Please clarify what "indirect" means and how such modifications are reflected in the change in risk calculations.

Response

The new CFW pump (formerly referred to as the new AFW pump) (NFPA 805 LAR Submittal, Table S-1, Item S1-11) represents a future additional train of feedwater that indirectly corrects VFDR(s) related to Emergency Feedwater (EFW) and is also available in scenarios where only non-fire affected failures of the EFW system exist.

Attachment to 2CAN061406 Page 7 of 8 The post-transition plant model includes all of the plant modifications including the new CFW pump. The compliant plant model reflects the lack of the new CFW pump by removing the VFDRs related to EFW, in the same manner a modification to protect or move a cable is reflected in the compliant plant model. This allows the change in risk calculations to accurately represent the VFDR mitigation and risk reduction benefits of the new CFW pump.

The remaining risk-related modifications are included in the compliant configuration when calculating the delta risk except incipient detection, which is credited solely to reduce risk (not mitigating any VFDRs) in the post-transition configuration.

PRA RAI 20 The LAR, Section 4.5.1.1, states that a peer review against Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk- Informed Activities," Revision 1, was performed on the Internal Events PRA model used for the Fire PRA. Please clarify when this peer review was performed. Also, given that the peer review was performed against an earlier version of the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA Standard than recognized by RG 1.200 Revision 2, please explain how the change in Supporting Requirements (SRs) between the PRA standard used in the peer review and the 2009 version of the ASME/ANS PRA Standard and SR clarifications in RG 1.200 are reconciled.

Response

The ANO-2 internal events PRA model was peer reviewed using ASME RA-Sb-2005 and RG 1.200, Revision 1. This peer review occurred during the week of May 4, 2008.

A detailed comparison of Internal Events High Level Requirements and Supporting Requirements was performed between ASME RA-Sb-2005 and ASME/ANS RA-SA-2009.

Changes in section numbers or Supporting Requirement (SR) numbers were not identified as differences unless the cross-reference is logically different. Most of the changes incorporated clarifications and qualifications from RG 1.200, Revision 1. These clarifications and qualifications were considered in the Owners Group peer review. Only two substantive changes were identified in this comparison. The first change was to remove the requirement for LERF sensitivity studies (LE-F2). The other change is that an emphasis on Internal Flooding walkdown documentation was included in SRs IFPP-B2, IFSO-B2, IFSN-B2, IFEV-B2, and IFQU-B2. Two new references were added for data (NUREG/CR-6928 and NUREG-1715).

In addition, the new clarifications and qualifications in RG 1.200, Revision 2, were reviewed.

Clarifications are provided when the NRC has no objection to the requirement, but considers the requirement somewhat unclear or ambiguous. Qualifications indicate that the staff has a technical concern with the requirement. No new qualifications were added for Section 2 of the ASME/ANS standard. The new clarifications involved capitalizing an action verb and adding a reference and would not have impacted the peer review performed at ANO-2.

None of the above changes would invalidate the ANO-2 peer review or change any findings and observations.

Attachment to 2CAN061406 Page 8 of 8 Summary This letter provides the requested responses to NRC RAIs (Reference 8) that were designated as required by July 9, 2012, associated with the ANO-2 NFPA-805 LAR dated December 17, 2012 (Reference 1). As stated previously, responses to RAIs PRA 01.e.01 and PRA 21 will be provided by August 11, 2014, as required by Reference 8. In addition, the results of updates and/or changes associated with the RAI responses included in this letter will be included as part of the response to PRA 21.

REFERENCES

1. Entergy letter dated December 17, 2012, License Amendment Request to Adopt NFPA-805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) (2CAN121202) (ML12353A041)

2. NRC letter dated September 11, 2013, Arkansas Nuclear One, Unit 2 - Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA-805 (TAC No. MF0404) (2CNA091301) (ML13235A005)

3. Entergy letter dated November 7, 2013, Response to Request for Additional Information -

Adoption of National Fire Protection Association Standard NFPA-805 (2CAN111301)

(ML13312A877)

4. Entergy letter dated December 4, 2013, Response to Request for Additional Information -

Adoption of National Fire Protection Association Standard NFPA-805 (2CAN121302)

(ML13338A432)

5. Entergy letter dated January 6, 2014, Response to Request for Additional Information -

Adoption of National Fire Protection Association Standard NFPA-805 (2CAN011401)

(ML14006A315)

6. NRC letter dated March 28, 2014, Arkansas Nuclear One, Unit 2 - Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA-805 (2CNA031401) (ML14085A225)

7. Entergy letter dated May 22, 2014, Response to Request for Additional Information -

Adoption of National Fire Protection Association Standard NFPA-805 (2CAN051404)

(ML14142A410)

8. NRC letter dated June 9, 2014, Arkansas Nuclear One, Unit 2 - Request for Additional Information Regarding Adoption of National Fire Protection Association Standard NFPA-805 (2CNA061402) (ML14155A133)