2CAN081401, Response to Request for Additional Information Adoption of National Fire Protection Association Standard NFPA-805

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

ML14219A635
Person / Time
Site:	Arkansas Nuclear
Issue date:	08/07/2014
From:	Jeremy G. Browning Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN081401
Download: ML14219A635 (135)

Text

s 2CAN081401 August 7, 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 01e.01 and PRA 21 by July 9, 2014. Reference 9 contained the RAI responses due on or before July 9, 2014.

of this letter includes the Entergy responses to the remaining two questions.

Responses to the various RAIs and incorporation of additional insights have resulted in changes to the PRA model, supporting documents, and overall risk results. Therefore, updates to affected attachments included in previous correspondence are included in attachments to this letter.

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.

Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Jeremy G. Browning Vice President - Operations Arkansas Nuclear One

2CAN081401 Page 2 of 3 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.

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 August 7, 2014.

Sincerely, ORIGINAL SIGNED BY DALE E. JAMES FOR JEREMY G. BROWNING JGB/dbb Attachments:

1.

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

2.

Updated Table 4 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features

3.

Updated Attachment C Pages - Fire Area Transition

4.

Updated Attachment G, Table G Recovery Actions and Activities

5.

Updated Attachment S, Table S1 - Plant Modifications

6.

Updated Attachment V - Fire PRA Quality

7.

Updated Attachment W - Fire PRA Insights

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)

2CAN081401 Page 3 of 3

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)

9.

Entergy letter dated June 30, 2014, Response to Request for Additional Information - Adoption of National Fire Protection Association Standard NFPA-805 (2CAN061406) 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

to 2CAN081401 Responses to Request for Additional Information ANO-2 Transition to NFPA-805 to 2CAN081401 Page 1 of 10 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 01e.01 and PRA 21 by July 9, 2014. Reference 9 contained the RAI responses due on or before July 9, 2014. The following include the responses to the remaining two questions that were designated as due by August 11, 2014.

The respective question is included for convenience.

Probabilistic Risk Assessment (PRA) RAI 01.e.01 By letter dated November 7, 2013, the licensee responded to PRA RAI 01.e, and noted the use of a single conditional core damage probability (CCDP) estimate of 6.97E-2 for failing to successfully shut down the reactor following a main control room (MCR) abandonment scenario.

The response to PRA RAI 15a in the same letter further states:

The control room abandonment compliant case represents the existing as-built, as operated plant if all of the fire area variances from deterministic requirements (VFDRs) in the control room were eliminated; in other words, if the control room was deterministically compliant...

For the control room abandonment scenario, all of the specific VFDR related components, which if protected would eliminate the VFDR, were set to their random failure probability instead of to "failed by the fire." Setting these components to their random failure probability provides an estimate of the fire risk if individual modifications were made to protect or reroute the components, thereby eliminating the VFDRs. The other components in the Fire Probabilistic Risk Assessment (FPRA) model that are impacted by the fire scenario were set to "failed by the fire."

The second paragraph above indicates that VFDRs in the MCR room are known and that their removal could be modeled by setting the related components to their random failure probability instead of failed by fire. Using the random failure probability instead of failed by fire is one acceptable method of estimating the compliant plant risk for use in estimating the change in risk associated with retaining VFDRs. However, in a letter dated January 6, 2014, the licensee responded to PRA RAI 16b and stated:

Full room burn-up scenarios in the Cable Spreading Room (CSR) and the MCR have the highest (and the same) compliant case CCDP in Fire Area G. Though a few scenarios in Fire Area G have a higher Core Damage Frequency (CDF) than the MCR full room burn-up scenario because of varying ignition frequencies and non-suppression probabilities, those scenarios are bounded by the CCDP of the MCR and CSR full room burn-up fires. The MCR full room burn-up scenario is also known as the MCR abandonment scenario.

to 2CAN081401 Page 2 of 10 The paragraph above indicates that the full room burn-up scenario is the only scenario included in the compliant plant MCR abandonment scenario. In contrast to the full room-burn up scenario, the response to PRA RAI 1.d.i in the letter dated December 4, 2013, indicates that full room burn-up is not the only MCR fire scenario that can cause loss of habitability. The licensee stated:

Abandonment times were calculated based on the limiting habitability condition of either smoke (visibility) or heat. The various configurations, the output from the CFAST runs, and the resultant abandonment times are documented in CALC-AN02-FP-09-00013, "ANO-2 Control Room Abandonment Times." This calculation includes an evaluation of impact of a fire initiating in the ANO-2 MCR, as well as a fire initiating in the ANO-1 MCR, and the transfer of heat and smoke to the ANO-2 MCR.

a)

Please explain how the CCDPs and conditional large early release probabilities (CLERPs) are estimated for fires that lead to MCR abandonment, in both the post-transition and the compliant case. Provide the range of CCDPs and CLERPS that have been developed and explain how the process and the range of estimates eventually developed address the types of MCR abandonment scenarios that the NRC staff has identified as appropriately characterizing such fires. These types of scenarios are:

i.

Scenarios where the fire fails few functions aside from forcing MCR abandonment and successful shutdown is straightforward.

ii.

Scenarios where the fire could cause some recoverable functional failures or spurious operations that complicate the shutdown but successful shutdown is likely.

iii.

Scenarios where the fire induced failures cause great difficulty for shutdown by failing multiple functions and/or complex spurious operations that make successful shutdown unlikely.

b)

Using the frequency for full room burn-up with the relatively low Human Error Probabilities (HEP) of 0.0697 seems inconsistent. It would appear that the scenario following full room burn-out would be the most complex and confusing and that the highest HEP should be assigned because the greatest fire damage possible is achieved. Please describe the types of spurious operations and instrumentation failures and access issues that are associated with full MCR burn-up and how these have been evaluated and included in the HEP.

c)

Please describe the total frequency of fires that initiate in the MCR and the total frequency of fires that lead to MCR abandonment.

Response

a)

The compliant case CCDP and CLERP are based on a review of VFDRs and elimination of the associated fire induced hardware failures to allow only random failures of the associated components to define the associated risk. For operator actions credited, such as the tripping of a Reactor Coolant Pump (RCP), the HEP for the control room action was applied as a conservative surrogate when performing the action at a primary control station.

to 2CAN081401 Page 3 of 10 The variant case CCDP and CLERP are based on credit for control of the new AFW pump at its control panel, outside the control room, using its associated HEP (developed using the NUREG-1921 methodology) as well as credit for HEPs associated with tripping of the RCPs at the switchgear and de-energizing the letdown valves at the respective power supply panel (also developed using the NUREG-1921 methodology).

With respect to the total abandonment frequency associated with fires in the ANO-1 and ANO-2 control rooms, the use of a single MCR abandonment scenario to define the delta risk provides a bounding risk and delta risk estimate that is based on the assumption that all cables in the control room are damaged. In this worst case damage scenario the delta risk is maximized since the credit given to the VFDRs maximizes the offset of the risk of the variant case, which includes all VFDRs in the respective variant condition. Evaluation of single scenarios with less damage would result in a lower variant risk but the same compliant risk, since credit is taken for an ideal primary control station for which the risk is not dependent on the control room damage incurred. This approach is similar to the quantification of a scenario where the fire induced failures result in multiple functions failing and/or complex spurious operations that ultimately make successful shutdown unlikely (Item III of the RAI). The evaluation of I.

scenarios where the fire fails few functions aside from forcing MCR abandonment and successful shutdown is straightforward, and II.

scenarios where the fire could cause some recoverable functional failures or spurious operations that complicate the shutdown but successful shutdown is likely will reflect more limited damage, which will result in a smaller delta risk since in the limit the delta is zero for a scenario which causes no damage. The primary advantage in applying a scenario-by-scenario MCR abandonment analysis for evaluation of delta risk is that a more accurate, but less conservative, delta risk results from each scenario evaluated. Therefore, the current methodology provides a bounding delta risk value. In addition, the large negative delta risk value for the ANO-2 FPRA due to the incorporation of significant modifications which are not required for a compliant plant will offset the most conservative estimate of the delta risk which can be obtained, one in which the compliant case risk is assumed to be zero (i.e., the MCR abandonment variant case CDF risk is 1.3E-05 and the total delta risk is a negative delta, the absolute value of which exceeds the MCR abandonment variant case risk).

b)

The 0.0697 CCDP applicable to the MCR abandonment variant case has been updated based on revision of associated HEPs using the methodology specified in NUREG-1921.

The updated CCDP is 0.151. The operator actions credited post-abandonment in the variant case are:

RHF2LTDWNP - Operators fail to isolate letdown flow outside the MCR RHF2RCPSLP - Operators fail to trip RCPs at the switchgear QHF2P75BFP - Operator fails to start and align new AFW pump locally following MCR abandonment to 2CAN081401 Page 4 of 10 The letdown isolation action and the RCP trip actions are required to mitigate multiple spurious operations (MSOs) which could otherwise impact the shutdown from outside the control room. The operator control of the new AFW pump (QHF2P75BFP) includes the required instrumentation to monitor secondary heat removal during shutdown. Other instrumentation required for post fire shutdown will be available at the new AFW pump control panel. No access issues arise from the control room fire since paths required for the above actions are not impacted by the control room fire.

As discussed in item a) above, the delta risk calculated based on a zero compliant case risk would still result in an acceptable delta risk given the large negative delta associated with the significant, beyond compliance modifications implemented as part of the NFPA 805 transition for ANO-2.

c)

The total control room ignition frequency for ANO-1 (Fire Zone 197-F) is 4.76E-03/reactor year. The total control room ignition frequency for ANO-2 (Fire Zone 2199-G) is 4.79E-03/reactor year.

The total abandonment frequency is 8.20E-05 per reactor year for the ANO-2 control room.

PRA RAI 21 Section 2.4.3.3 of the NFPA 805 standard incorporated by reference into 50.48(c) states that the probabilistic safety assessment (PSA) (PSA is also referred to as PRA) approach, methods, and data shall be acceptable to the authority having jurisdiction (AHJ), which is the NRC.

a)

For each method (i.e., each bullet) below, please indicate how the issue will be addressed in: (i) the final composite analysis results provided in support of the LAR and (ii) the PRA that will be used at the beginning of the self-approval of post-transition changes. Please note that the licensee may replace any of these methods and weaknesses with a method or model previously accepted by the NRC by modifying the FPRA.

PRA RAI 01.g regarding removal of the electric panel factors and additional credit for suppression and detection.

PRA RAI 01.k regarding state of knowledge correlations (SOKC) associated with parametric data propagation for fire ignition frequency bins, spurious operations, non-suppression probabilities, HFEs, and internal events.

PRA RAI 03 regarding removal of credit for incipient detection in the control element drive mechanism (CEDM) room cabinets to limit cable damage within a cabinet.

PRA RAI 06 Regarding removal credit of CPT credit in assessment of circuit failure probabilities.

PRA RAI 06.02 regarding minimum non-suppression probabilities of 1E-3 (RAI 6.02 requests clarification of the values used, not justification for any proposed deviation).

to 2CAN081401 Page 5 of 10 FM RAI 01.b use of 10 minutes propagation time between MCR cabinets in lieu of 15 minutes.

FM RAI 01.c MCR transient fire growth rates changed to be consistent with NUREG/CR-6850.

Additionally, please explain the statement "Because the revised transient growth rates are consistent with the current NUREG guidance, it is not necessary to compute the impact on fire risk results."

FM RAI 01.d updated MCR abandonment calculation that no longer uses the hot gas layer smoke concentration and temperature modifications.

FM RAI 01.f new transient zone of influence (ZOI) tables.

FM RAI 01.g and FM RAI 04 new cable ignition ZOIs and hot gas layers (HGLs).

PRA RAI 01.e.01 regarding MCR abandonment CCDP/CLERP estimates.

PRA RAI 01.f.01 regarding using 69 kW instead of 317 kW heat release rates (HRRs) for transient fires.

PRA RAI 08.01 regarding treatment of electrical cabinet fires in the MCR abandonment analysis.

PRA RAI 08.02 regarding treatment of MCB fires.

PRA RAI 09.01 regarding evaluation of non-significant HFEs using guidance in NUREG-1921 opposed the licensee's multiplier approach.

PRA RAI 15.01 regarding modelling of the failure of cables with unknown locations.

b)

In a letter dated November 7, 2013, the licensee responded to RAI 01.h and indicated that the PRA will be revised under implementation item S2-9 to address all modifications and implementation items, including the procedure changes in implementation item S2-6. The response does not indicate when this model will be available and how the model will be consistent with the response to a) above. Please provide a method to assure that all changes will be made and that a focused-scope peer review will be performed on changes that are PRA upgrades as defined in the PRA standard, and that any findings will be resolved before self-approval of post-transition changes (For example, a license condition).

c)

Please provide final risk estimates (CDF, LERF, CDF, LERF, and additional risk of recovery actions) based on a PRA that uses only acceptable methods for any method that impacts the transition change in risk estimates. Discuss the likelihood that the risk increase in any individual fire area would exceed the acceptance guidelines and, if so, why exceeding the guidelines is acceptable.

to 2CAN081401 Page 6 of 10

Response

a)

The disposition of each of the items listed in a) above is provided in the table below, including how each item is addressed in the final composite analysis and the post-transition analysis.

Item Final Composite Analysis Treatment Post Transition Analysis Treatment PRA 01.g regarding removal of the electric panel factors and additional credit for suppression and detection Electrical panel factors eliminated Electrical panel factors eliminated PRA RAI 01.k regarding SOKC associated with parametric data propagation for fire ignition frequency bins, spurious operations, non-suppression probabilities, human failure events (HFEs), and internal events Incorporated in uncertainty analysis Incorporated in uncertainty analysis PRA RAI 03 regarding removal of credit for incipient detection in the CEDM room cabinets to limit cable damage within a cabinet Eliminated credit for incipient detection for non-severe fire (credited only for severe fire, impacting targets outside of cabinet)

Eliminated credit for incipient detection for non-severe fire (credited only for severe fire, impacting targets outside of cabinet)

PRA RAI 06 Regarding removal credit of control power transformer (CPT) credit in assessment of circuit failure probabilities Eliminated CPT credit Eliminated CPT credit PRA RAI 06.02 regarding minimum non-suppression probabilities of 1E-3 (RAI 6.02 requests clarification of the values used, not justification for any proposed deviation)

Incorporated floor value non-suppression probability of 1E-3 Incorporated floor value non-suppression probability of 1E-3 FM RAI 01.b use of 10 minutes propagation time between MCR cabinets in lieu of 15 minutes 10 minute propagation time used 10 minute propagation time used to 2CAN081401 Page 7 of 10 Item Final Composite Analysis Treatment Post Transition Analysis Treatment FM RAI 01.c MCR transient fire growth rates changed to be consistent with NUREG/CR-6850 Additionally, explain the statement Because the revised transient growth rates are consistent with the current NUREG guidance, it is not necessary to compute the impact on fire risk results Revised transient fire growth rates consistent with NUREG/CR-6850 The statement is intended to indicate that the updated risk numbers will use the revised transient growth rates and therefore a specific calculation of the impact of this change will not be addressed individually, but in the overall risk impact of all RAIs Revised transient fire growth rates consistent with NUREG/CR-6850 FM RAI 01.d updated MCR abandonment calculation that no longer uses the hot gas layer smoke concentration and temperature modifications The MCR abandonment calculation no longer uses the hot gas layer smoke concentration and temperature modifications The MCR abandonment calculation no longer uses the hot gas layer smoke concentration and temperature modifications FM RAI 01.f new transient ZOI tables Incorporated new transient zone of influence tables for scenarios with secondary combustible cable Incorporated new transient zone of influence tables for scenarios with secondary combustible cable FM RAI 01.g and FM RAI 04 new cable ignition ZOIs and HGLs Incorporated new cable ignition ZOI and time to HGL analyses for scenarios including secondary combustible cable Incorporated new cable ignition ZOI and time to HGL analyses for scenarios including secondary combustible cable PRA RAI 01.e.01 regarding MCR abandonment CCDP/CLERP estimates See response to RAI 01.e.01; no specific change to analysis resulting from this RAI See response to RAI 01.e.01; no specific change to analysis resulting from this RAI to 2CAN081401 Page 8 of 10 Item Final Composite Analysis Treatment Post Transition Analysis Treatment PRA RAI 01.f.01 regarding using 69kW instead of 317kW HRRs for transient fires 69 kW HRRs used in specified zones with additional controls, as discussed in RAI 01.f.01 response; all other transient fires use the 317 kW HRR 69 kW HRRs used in specified zones with additional controls, as discussed in RAI 01.f.01 response; all other transient fires use the 317 kW HRR PRA RAI 08.01 regarding treatment of electrical cabinet fires in the MCR abandonment analysis Analysis now incorporates multiple cable bundles for all control room panels; please refer to the updated Scenario Report Analysis now incorporates multiple cable bundles for all control room panels; please refer to the updated Scenario Report PRA RAI 08.02 regarding treatment of MCB fires.

All MCB panels have been combined into one scenario with the MCB frequency applied and an Appendix L zero distance factor also applied All MCB panels have been combined into one scenario with the MCB frequency applied and an Appendix L zero distance factor also applied PRA RAI 09.01 regarding evaluation of non-significant HFEs using guidance in NUREG-1921 opposed the licensees multiplier approach All HFEs reflect application of NUREG-1921 methodology All HFEs reflect application of NUREG-1921 methodology PRA RAI 16.01 regarding modelling of the failure of cables with unknown locations (the NRC RAI referenced PRA RAI 15.01; the correct reference is PRA RAI 16.01)

Please see PRA RAI 16.01 response for evaluation of delta risk without assumption of unknown cable failures in compliant case1 Please see PRA RAI 16.01 response for evaluation of delta risk without assumption of unknown cable failures in compliant case1 1 The values provided in the response to PRA RAI 16.01 have been recalculated using the post-RAI model. The CDF/LERF increase associated with the VFDRs and recovery actions is 3.02E-05 for CDF and 5.75E-07 for LERF. The total change in risk (combined risk increases and risk decreases) is -1.59E-04 for CDF and

-5.30E-06 for LERF. Absent the unknown routing assumption, the total risk reduction for CDF/LERF is

-7.85E-05 for CDF and -3.18E-06 for LERF.

b)

The referenced commitments (S2-6 and S2-9) will be closed only after review of modification implementation items and update of analysis as required. The resulting risk will be confirmed to be within the reported post transition risk. Any changes in method which represent an upgrade will be subjected to a peer review. No new methods are anticipated. ANO Procedure EN-DC-151 ensures the model is maintained and requires that a peer review be initiated for a methods change in accordance with ASME/ANS PRA Standard RA-Sa-2009.

to 2CAN081401 Page 9 of 10 c)

The final Fire PRA quantification results based on the methods as outlined in Item a) above are included in a revised NFPA 805 LAR Attachment W provided in Attachment 7 of this letter. The results meet the acceptance guidelines of Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.

Summary This letter provides the requested responses to NRC RAIs (Reference 8) associated with the ANO-2 NFPA-805 LAR dated December 17, 2012 (Reference 1) not included in Entergy response letter dated June 30, 2014 (Reference 9).

Responses to the various RAIs and incorporation of additional insights have resulted in changes to the PRA model, supporting documents, and overall risk results. Therefore, updates to affected attachments included in previous correspondence are enclosed to this letter as listed below. The enclosed pages include changes identified in all RAI response correspondence since submittal of the original LAR. The attached pages supersede all like-pages previously submitted.

Att. 2 - Table 4-3 Revised to be consistent with LAR Att. C Required Fire Protection Systems and Features. The Fire Risk Evaluations (FREs) that support LAR Att. C have been revised to reflect the results of the updated engineering report PRA-A2-05-003, Rev. 1, ANO-2 Fire Probabilistic Risk Assessment Fire Scenarios Report NUREG/CR-6850 Tasks 8 and 11 (ERIN Report Number 0247-06-0006.05, Rev. 4),

Table 9-3, Zones That Credit Suppression and Table 9-4, Zones That Credit Detection.

Att. 3 - Attachment C Updated to reflect changes to the FREs, which were revised to support responses to RAIs.

Att. 4 - Attachment G Revision to Table G-1 was required to eliminate those recoveries determined to not be risk significant as a result of updating the fire PRA model.

Att. 5 - Attachment S CALC-09-E-0008-10, Rev. 1, ANO NFPA 805 Fire Risk Evaluation -

Fire Area G has eliminated modification of two motor-operated valves (MOVs) for Fire Area G (Table S1, Item S1-7). In addition, the use of non-metallic ceramic braiding is being considered for use to isolate target conductors (Table S1, Item S1-8). Finally, the ANO engineering design process determined that upgrading the existing hydrogen bottle storage location to meet applicable code requirements is preferable to relocation (Table S1, Item S1-14).

Att. 6 - Attachment V Updated to eliminate discussion of unapproved methods and incorporate new focused scope peer review for the Human Reliability Analysis (HRA). The entire revised Att. V is included, but only pages V-1, V-2, V-25, and V-27 contain new information.

Att. 7 - Attachment W Updated to reflect responses to RAIs and changes to the FREs.

to 2CAN081401 Page 10 of 10 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)

9.

Entergy letter dated June 30, 2014, Response to Request for Additional Information -

Adoption of National Fire Protection Association Standard NFPA-805 (2CAN061406)

to 2CAN081401 Updated Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features

Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements to 2CAN121202 Page 74 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D)

Required Detection System (S, L, E, R, D)

Required Fire Protection Feature (S, L, E, R, D)

Required Fire Protection Feature and System Details 2MH01E Between Aux Bldg and Intake Structure 2MH01E Concrete Manhole East 4.2.4.2 None None N/A 2MH01W Between Aux Bldg and Intake Structure 2MH01W Concrete Manhole West 4.2.3.2 None None N/A 2MH02E Between Aux Bldg and Intake Structure 2MH02E Concrete Manhole East 4.2.4.2 None None N/A 2MH02W Between Aux Bldg and Intake Structure 2MH02W Concrete Manhole West 4.2.3.2 None None N/A 2MH03E Between Aux Bldg and Intake Structure 2MH03E Concrete Manhole East 4.2.4.2 None None N/A 2MH03W Between Aux Bldg and Intake Structure 2MH03W Concrete Manhole West 4.2.3.2 None None N/A AA B HPSI, LPSI and Containment Spray Pump Room and Gallery 2007-LL "B" HPSI, LPSI and Containment Spray Pump Room and Gallery 4.2.4.2 E

E, R, D N/A Detection and Partial Suppression AAC Alternate AC Diesel 2MH12 Manhole near SBO diesel 4.2.3.2 None None N/A SBOD Alternate AC Diesel 4.2.3.2 N/R N/R N/A ADMIN Administration Building ADMIN Administration Building 4.2.3.2 None N/R N/A B-2 Unit 2 General Plant Multiple Elevations 2045-XX Turbine Lube Oil Storage Tank Room 4.2.4.2 N/R None N/A 2078-QQ Heat Exchanger Equipment Room 4.2.4.2 None None N/A 2092-PP Chiller Water System Equipment Room 4.2.4.2 None None N/A 2147-A Chemical Storage Room 4.2.4.2 E, R None N/A Suppression 2148-A Corridor 4.2.4.2 None None N/A 2151-A Fuel Handling Room (El. 404) 4.2.4.2 None E, R, D N/A Detection

Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements to 2CAN121202 Page 75 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D)

Required Detection System (S, L, E, R, D)

Required Fire Protection Feature (S, L, E, R, D)

Required Fire Protection Feature and System Details 2152-D Computer Room 4.2.4.2 None E, R, D N/A Detection 2153-A Ventilation Equipment Room 4.2.4.2 None None N/A 2155-A Steam Pipe Room 4.2.4.2 None None N/A 2156-A Containment Purge Air Equipment Room 4.2.4.2 None R, D N/A Detection 2172-ZZ Storage and Shop Room 4.2.4.2 N/R None N/A 2177-YY Neutralizer Tank Room 4.2.4.2 None None N/A 2178-AAA Lube Oil Reservoir 4.2.4.2 N/R None N/A 2200-MM Turbine Building 4.2.4.2 E

E N/A Partial Suppression and Detection 2201-B ANO-2 Operations Support Facility 4.2.4.2 None N/R N/A 2223-KK Pipeway Equipment Access Room (Aux.

Bldg. Extension) 4.2.4.2 None E, R, D N/A Detection 2225-WW Regenerative Waste Pump & Tank Room 4.2.4.2 None None N/A 2229-SS Storage Room 2232 4.2.4.2 E

None N/A Partial Suppression 2230-RR Drum Filling Room 4.2.4.2 None None N/A 2231-TT Plant Heating Boiler Room 4.2.4.2 N/R None N/A 2242-OO H&V Mechanical Equipment Room, AO Shack, Lab & Storage Room 4.2.4.2 R

N/R N/A Partial Suppression 2243-NN Chemistry Lab, Kitchen & Offices 4.2.4.2 N/R None N/A 2261-UU Plant Heating Boiler Day Tank 4.2.4.2 N/R None N/A B-3 North Penetration Areas 2091-BB North Electrical Equipment Room 4.2.4.2 None E, R, D N/A Detection 2112-BB Lower North Electrical Penetration Room 4.2.4.2 E, R E, R, D N/A Suppression and Detection 2183-J Upper North Electrical Penetration Room 4.2.4.2 E, R E, R, D N/A Suppression and Detection B-4 CEDM Room 2154-E CEDM Equipment Room 4.2.4.2 None E, R, D N/A Detection

Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements to 2CAN121202 Page 76 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D)

Required Detection System (S, L, E, R, D)

Required Fire Protection Feature (S, L, E, R, D)

Required Fire Protection Feature and System Details B-5 North and South Aux Bldg Stair 2149-B Stairwell No. 2001 4.2.3.2 None None N/A 2158-F Stairwell No. 2055 4.2.3.2 None None N/A B-6 Aux Bldg General Access Area, A & C Pump Rooms 2006-LL General Access Room 4.2.4.2 E

E, R, D N/A Detection and Partial Suppression 2010-LL "C" HPSI Pump Room 4.2.4.2 None E, R, D N/A Detection 2011-LL Tendon Gallery Access 4.2.4.2 None E, R, D N/A Detection 2014-LL "A" HPSI, LPSI, & Containment Spray Pump Room 4.2.4.2 None E, R, D N/A Detection CC Emergency Feedwater Pump Room (Turbine Driven) 2024-JJ Emergency Feedwater Pump Room (Turbine Driven) 4.2.3.2 None E, R N/A Detection DD Unit 2 General Area 335' Elevation 2019-JJ Boric Acid Condensate Tank Room 4.2.4.2 None None N/A 2032-JJ Spent Resin Storage Tank Room 4.2.4.2 None None N/A 2040-JJ Corridor 4.2.4.2 None E, D N/A Partial Detection 2068-DD Hot Machine Shop 4.2.4.2 None E, R, D N/A Detection EE-L South Piping Penetration Rooms 2055-JJ Lower South Piping Penetration Room 4.2.4.2 None E, R, D N/A Detection 2084-DD Upper South Piping Penetration Room and Waste Gas Equipment Room 4.2.4.2 None E, D N/A Partial Detection EE-U Lower South Electrical Penetration 2111-T Lower South Electrical Penetration Room 4.2.4.2 E, R E, R, D N/A Suppression and Detection FF Emergency Feedwater Pump Room (Motor Driven) 2025-JJ Emergency Feedwater Pump Room (Motor Driven) 4.2.3.2 None E, R N/A Detection

Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements to 2CAN121202 Page 77 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D)

Required Detection System (S, L, E, R, D)

Required Fire Protection Feature (S, L, E, R, D)

Required Fire Protection Feature and System Details G

Unit 2 Alternate Shutdown Areas 97-R Cable Spreading Room 4.2.4.2 N/R N/R N/A 129-F Control Room 4.2.4.2 N/R N/R N/A 2098-C CPC Room 4.2.4.2 E, R, D E, R, D N/A Detection and Suppression 2098-L Cable Spreading Room 4.2.4.2 E, R, D E, R, D N/A Detection and Suppression 2119-H CR Printer Room 4.2.4.2 None E, R, D N/A Detection 2136-I Health Physics Corridor 4.2.4.2 E

E, R, D N/A Detection and Partial Suppression 2137-I USEP Room, Decon, Hot Instrument Shop 4.2.4.2 E, R, D E, R, D N/A Detection and Suppression 2150-C Old CPC Room 4.2.4.2 None E, R, D N/A Detection 2199-G Unit 2 Control Room 4.2.4.2 None E, R, D N/A Detection GG Unit 2 North Electrical and Piping Penetration Area 2076-HH Electrical Equipment Room 4.2.4.2 None E, R, D N/A Detection 2081-HH Upper North and Lower North Piping Penetration Room 4.2.4.2 None E, R, D N/A Detection HH Unit 2 General Area 354' Elevation 2063-DD Sample Room 4.2.4.2 None None N/A 2072-R Upper Volume Control Tank Room, Lower Tank and Pump Room 4.2.4.2 None E, R, D N/A Detection 2073-DD Access Room, Pump Room, Tank Room (2B62 & Resin Addition Room) 4.2.4.2 E

E, R, D N/A Detection and Partial Suppression 2096-M Motor Control Center (2B63) 4.2.4.2 None E, R, D N/A Detection 2106-R Degasifier Vacuum Pump Room 4.2.4.2 None E, R, D N/A Detection 2107-N Corridor (North of Stairway 2001) 4.2.4.2 None E, R, D N/A Detection II North Switchgear Room 2101-AA North Switchgear Room 4.2.4.2 None R, D N/A Detection

Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements to 2CAN121202 Page 78 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D)

Required Detection System (S, L, E, R, D)

Required Fire Protection Feature (S, L, E, R, D)

Required Fire Protection Feature and System Details JJ Corridor 2109-U Corridor 4.2.4.2 E, R, D E, R, D N/A Detection and Partial Suppression K

Tank Rooms 16-Y Clean Waste Receiver Tank Room 4.2.3.2 None None N/A 2020-JJ Boron Holdup Tank Vault 4.2.3.2 None None N/A KK Unit 2 South Emergency Diesel Generator and Boric Acid Makeup Tank Rooms 2093-P South Emergency Diesel Generator Room 4.2.4.2 E

E, R, D N/A Detection and Partial Suppression 2114-I EDG Air Intake Room 4.2.4.2 None E, R, D N/A Detection 2115-I Boric Acid Makeup Tank Room 4.2.4.2 None E, R, D N/A Detection L

Diesel Fuel Storage Vault Area TKVLT Diesel Fuel Storage Vault 4.2.3.2 N/R N/R N/A MM West Battery and DC Equipment Rooms 2099-W West D.C. Equipment Room 4.2.4.2 None E, R, D N/A Detection 2103-V West Battery Room 4.2.4.2 None R, D N/A Detection NN Unit 2 Containment Building 2032-K Containment Building South Side 4.2.4.2 N/R E, R, D N/A Partial Detection 2033-K Containment Building North Side 4.2.4.2 N/R E, R, D N/A Partial Detection OO Unit 2 Intake Structure INTAKE Intake Structure (Unit 2) 4.2.4.2 N/R D

N/A Detection QQ North Emergency Diesel 2094-Q North Emergency Diesel Generator Room 4.2.3.2 E, R E, R N/A Suppression and Detection 2114-I EDG Air Intake Room 4.2.3.2 None E, R N/A Detection

Arkansas Nuclear One - Unit 2 4.0 Compliance with NFPA 805 Requirements to 2CAN121202 Page 79 Table 4-3 Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features Fire Area Fire Zone Description NFPA 805 Regulatory Basis Required Suppression System (S, L, E, R, D)

Required Detection System (S, L, E, R, D)

Required Fire Protection Feature (S, L, E, R, D)

Required Fire Protection Feature and System Details SS South Switchgear and East DC Equipment and Battery Rooms 2097-X East D.C. Equipment Room 4.2.4.2 None E, R, D N/A Detection 2100-Z South Switchgear Room 4.2.4.2 None R, D N/A Detection 2102-Y East Battery Room 4.2.4.2 None E, R, D N/A Detection TT Electrical Equipment (2B9/2B10) Room 2108-S Electrical Equipment (2B9/2B10) Room 4.2.4.2 None E, R, D N/A Detection YD Miscellaneous Yard Locations YARD Miscellaneous Yard Locations 4.2.3.2 N/R N/R N/A Legend:

Fire Protection Features are features required to meet NFPA 805 Chapter 3 requirements.

S - Credited Separation Criteria is derived from PRA in Attachment C - Table B-3 VFDRs.

E - EEEE Criteria: Credited Systems/Features are derived from Attachment A - Table B-1 and/or Attachment C - Table B-3.

L - NRC approved Licensing Action is derived from Attachment K and/or Attachment A - Table B-1 VFDRs.

R - Risk Criteria is derived from PRA in Attachment C - Table B-3.

D - Defense-In-Depth Criteria is derived from PRA in Attachment C - Table B-3.

N/R - System is operational in fire area, however it is Not Required.

None - Fire protection feature is not present in the fire zone.

to 2CAN081401 Updated Attachment C Pages Fire Area Transition

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-38 Fire Area ID:

B Turbine Building & General Areas Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2045-XX Turbine Lube Oil Storage Tank Room P

No No No No No No No No No No No 2078-QQ Heat Exchanger Equipment Room No No No No No No No No No No No No 2092-PP Chiller Water System Equipment Room No No No No No No No No No No No No 2147-A Chemical Storage Room Yes No No No No No Yes No Yes No No No 2148-A Corridor No No No No No No No No No No No No 2151-A Fuel Handling Room (El. 404)

No Yes No No No No No Yes No Yes No Yes 2152-D Computer Room No Yes No No No No No Yes No Yes No Yes 2153-A Ventilation Equipment Room No No No No No No No No No No No No 2155-A Steam Pipe Room No No No No No No No No No No No No 2156-A Containment Purge Air Equipment Room No Yes No No No No No No No Yes No Yes 2172-ZZ Storage and Shop Room P

No No No No No No No No No No No 2177-YY Neutralizer Tank Room No No No No No No No No No No No No 2178-AAA Lube Oil Reservoir P

No No No No No No No No No No No 2200-MM Turbine Building P

P No No No No Yes Yes No No No No 2201-B Operations Support Facility No Yes No No No No No No No No No No 2223-KK Pipe-way Equipment Access Room (Aux.

Bldg. Extension)

No Yes No No No No No Yes No Yes No Yes 2225-WW Regenerative Waste Pump & Tank Room No No No No No No No No No No No No 2229-SS Storage Room 2232 P

No No No No No Yes No No No No No 2230-RR Drum Filling Room No No No No No No No No No No No No 2231-TT Plant Heating Boiler Room P

No No No No No No No No No No No 2242-OO H&V Mechanical Equipment Room, AO Shack, Lab & Storage Room Yes P

No No No No No No Yes No No No 2243-NN Chemistry Lab, Kitchen & Offices P

No No No No No No No No No No No 2261-UU Plant Heating Boiler Day Tank P

No No No No No No No No No No No P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-39 Fire Area ID:

B Turbine Building & General Areas Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-21

Title:

ANO-2 Fire Area B-2 Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions The following equipment is recovered in the post transition baseline case:

2A-309 switchgear breaker 2P-32A reactor coolant pump 2P-32B reactor coolant pump 2P-32C reactor coolant pump 2P-32D reactor coolant pump Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection system located in Fire Zones 2151-A, 2152-D, 2156-A, and 2223-KK was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

The suppression systems in Fire Zones 2147-A and 2242-OO were credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-48 Fire Area ID:

B North Penetration Areas Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-22

Title:

ANO-2 Fire Area B-3 Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions The following equipment is recovered in the post transition baseline case:

2CV-0789-1 EFW pump condensate suction valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

The following modification is area specific and credited to reduce risk in this fire area:

2CV-4698-1 circuit modified to prevent spurious opening and resolve IN 92-18 issue.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. The modification to 2CV-4698-1 will prevent spurious operation.

Additional Fire Area Considerations The detection system located in Fire Area B-3 was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

The suppression systems in Fire Zones 2112-BB and 2183-J were credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-56 Fire Area ID:

B CEDM Room Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-05

Title:

ANO-2 Fire Area B-4 Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs.

Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

The following modification is area specific and credited to reduce risk in this fire area:

2C70, 2C71, 2C72, 2C73, 2C75, 2C80, and 2C409 panels will have incipient detection installed.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection system located in Fire Area B-4 was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-58 Fire Area ID:

B CEDM Room Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs VFDR ID:

B4-01 VFDR:

Fire damage to control cables for RCPs 2P-32A (breaker 2H-11), 2P-32B (breaker 2H-21), 2P-32C (breaker 2H-22) and 2P-32D (breaker 2H-12) can result in a spurious re-start of 2P-32A, 2P-32B, 2P-32C and 2P-32D, respectively. Securing the pumps assures normal pressurizer spray is secured and prevents potential RCP seal damage.

Loss of these functions could challenge the Pressure and Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.

Disposition:

This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with operating procedure changes (reference Table S-2, Item S2-7).

VFDR ID:

B4-02 VFDR:

Fire damage to control cables associated with 2CV-4824-2 (IN 92-18) can result in a spurious failure of the valve in the open position.

This results in a diversion path from the primary inventory control path to auxiliary pressurizer spray using the charging system.

Makeup using the charging pump is through the high pressure safety injection (HPSI) header by manually opening 2CVC-115 and closing 2CV-4840-2 from Control Room to isolate auxiliary spray.

Loss of these functions could challenge the Inventory and Pressure Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.

Disposition:

No further action is required as 2CV-4824-2 is a component in the chemical and volume control system (CVCS) system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios.

End of Fire Area B-4

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-70 Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk*

DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2024-JJ EFW Pump Room (Turbine Driven)

No Yes No No No No No Yes No Yes No No

• Detection credited in HGL/MCA P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.

VFDRs This fire area is in deterministic compliance and has no VFDRs.

End of Fire Area CC

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-79 Fire Area ID:

DD - Unit 2 General Area 335 Elevation Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID:

CALC-ANOC-FP-09-00003 Units 1 & 2 HELB Doors Fire Protection Engineering Evaluation Summary:

Purpose:

This evaluation is to document in an engineering report form the fire protection engineering evaluation for the ANO-1 and ANO-2 fire doors that are also classified as high energy line break (HELB) doors with modified or missing door latches.

Basis for Acceptability: The room is protected by a smoke detection system that alarms in the Control Room and the prompt response by the fire brigade with access to manual firefighting equipment would prevent any fire (in the unlikely event one does occur) from building sufficient pressure to open the door.

Engineering Evaluation ID:

CALC-ANOC-FP-09-00011 Fire Protection Engineering Evaluation of Units 1 & 2 Aux Bldg Elevator Doors Summary:

Purpose:

Evaluate elevator doors as part of the 3-hour fire boundary.

Unit 2:

335' 2040-JJ Fire Area DD 354' 2073-DD Fire Area HH 386' 2136-I Fire Area G 404' 2151-A Fire Area B-2 The elevator doors were previously evaluated in Calculation 85-E-0053-04 however the calculation will be superseded by this evaluation as part of the NFPA 805 transition project.

Basis for Acceptability: Based on the low and moderate (Fire Zone 2136-I only) combustible loading, the availability of the smoke detection systems (and suppression system in Fire Zone 67-U) and the availability of the fire brigade with manual fire fighting equipment, the elevator doors are considered to be adequate for the hazards in the area and acceptable for the 3-hour rated fire barriers.

Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2019-JJ Boric Acid Condensate Tank Room No No No No No No No No No No No No 2032-JJ Spent Resin Storage Tank Room No No No No No No No No No No No No 2040-JJ Corridor No P

No No No No No Yes No No No Yes 2068-DD Hot Machine Shop No Yes No No No No No Yes No Yes No Yes

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-80 Fire Area ID:

DD - Unit 2 General Area 335 Elevation Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary FRE Calculation:

CALC-09-E-0008-07

Title:

ANO-2 Fire Area DD Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs.

Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection system located in Fire Zone 2068-DD was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis.

The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-89 Fire Area ID:

EE-L - South Piping Penetration Area Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Engineering Evaluation ID:

CALC-ANOC-FP-09-00004 Engineering Evaluation of Units 1 &2 Containment Building Penetrations Summary:

Purpose:

This fire protection engineering evaluation is to evaluate the ANO-1 and ANO-2 Reactor and Containment Building penetrations to be used in a 3-hour rated fire area boundary.

Basis for Acceptability: The penetrations are considered adequate for the hazards in the area based on the low probability of a fire starting in the areas of the penetrations, the installed smoke detection and suppression systems (Auxiliary Buildings Electrical Penetration Rooms), the fire resistive materials used in the penetrations and the prompt response by the fire brigade with access to manual firefighting equipment for those areas in the units Auxiliary Buildings.

Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2055-JJ Lower South Piping Penetration Room No Yes No No No No No Yes No Yes No Yes 2084-DD Upper South Piping Penetration Room and Waste Gas Equipment Room No P

No No No No No Yes No No No Yes P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-90 Fire Area ID:

EE-L - South Piping Penetration Area Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-08

Title:

ANO-2 Fire Area EE-L Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs.

Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection system located in Fire Zone 2055-JJ was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis.

The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights DID Maintained:

The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 and shows no additional DID methods are required beyond those inherent to the fire area No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area.

Safety Margin Maintained:

All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.

Comments:

None

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-91 Fire Area ID:

EE-L - South Piping Penetration Area Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs VFDR ID:

EEL-01 VFDR:

Fire damage to cables in the area may impact EFW functions resulting in the following:

a)

Loss of control to normally closed 2CV-1075-1 results in loss of EFW flow from 2P-7B to the credited SG B.

b)

Loss of power and control to 2CV-1025-1 results in loss of control (isolation) of EFW flow to the non-credited SG A.

Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.

Disposition:

This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:

a)

No further action is required for 2CV-1075-1.

b)

No further action is required for 2CV-1025-1.

End of Fire Area EE-L

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-95 Fire Area ID:

EE-U - Lower South Electrical Penetration Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2111-T Lower South Electrical Penetration Room Yes Yes No No No No Yes Yes Yes Yes No Yes P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-96 Fire Area ID:

EE-U - Lower South Electrical Penetration Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-09

Title:

ANO-2 Fire Area EE-U Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs.

Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The suppression and detection systems located in Fire Area EE-U were credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-97 Fire Area ID:

EE-U - Lower South Electrical Penetration Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)

DID Maintained:

The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area.

No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area.

Safety Margin Maintained:

All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.

Comments:

None VFDRs VFDR ID:

EEU-01 VFDR:

Fire damage to cables in the area may impact EFW functions resulting in the following:

a)

Loss of control to 2CV-1036-2 (IN 92-18) resulting in potential spurious closure and loss of EFW flow from 2P-7B to the credited SG B.

Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.

Disposition:

This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-101 Fire Area ID:

FF - Emergency Feedwater Pump Room (Motor Driven)

Compliance Basis:

NFPA 805 Section 4.2.3.2 - Deterministic Approach Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk*

DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2025-JJ EFW Pump Room (Motor Driven)

No Yes No No No No No Yes No Yes No No

• Detection credited in HGL/MCA P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.

VFDRs This fire area is in deterministic compliance and has no VFDRs.

End of Fire Area FF

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-114 Fire Area ID:

G - Unit 2 Alternate Shutdown Areas Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 97-R ANO-1 Cable Spreading Room P

Yes No No No No No No No No No No 129-F Control Room P

Yes No No No No No No No No No No 2199-G Control Room No Yes No No No No No Yes No Yes No Yes 2119-H Printer Room No Yes No No No No No Yes No Yes No Yes 2136-I Health Physics Room P

Yes No No No No Yes Yes No Yes No Yes 2137-I Upper South Electrical Penetration Room Yes Yes No No No No Yes Yes Yes Yes Yes Yes 2150-C Core Protection Calculator Room (Old CPC Room)

No Yes No No No No No Yes No Yes No Yes 2098-C CPC Room Yes Yes No No No No Yes Yes Yes Yes Yes Yes 2098-L Unit 2 Cable Spreading Rooms Yes Yes No No No No Yes Yes Yes Yes Yes Yes P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-115 Fire Area ID:

G - Unit 2 Alternate Shutdown Areas Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-10

Title:

ANO-2 Fire Area G Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions The following equipment is recovered in the post transition baseline case:

2CV-4816 and 2CV-4817 Letdown isolation valves 2P-32A reactor coolant pump 2P-32B reactor coolant pump 2P-32C reactor coolant pump 2P-32D reactor coolant pump New AFW pump and suction & discharge valves 2CV-1036-2 EFW discharge valve 2CV-1075-1 EFW discharge valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

(continued)

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-116 Fire Area ID:

G - Unit 2 Alternate Shutdown Areas Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)

Summary (continued)

The following modifications are area specific and credited to reduce risk in this fire area:

2CV-4698-1 circuit protected to prevent spurious 2SV-4669-1 and 2SV-4670-2 circuits modified to prevent spurious 2CV-1036-2 circuit protected to prevent spurious and eliminate IN 92-18 issues 2CV-1075-1 circuit protected to prevent spurious and eliminate IN 92-18 issues IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. EFW discharge valve 2CV-1036-2 and 2CV-1075-1 modifications resolve the IN 92-18 issues in this fire area.

Additional Fire Area Considerations The available detection systems in Fire Area G were credited in the manual suppression curves in the HGL and MCA analysis to limit the growth of the fire. The suppression systems in Fire Zones 2137-I, 2098-C, and 2098-L are credited in the HGL and the MCA analysis to limit the fire growth in Fire Area G.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-122 Fire Area ID:

G - Unit 2 Alternate Shutdown Areas Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID:

G-03 (continued)

Disposition:

This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:

a.

No further actions for 2CV-1050-2.

b.

No further actions for 2CV-0340-2.

c.

No further actions for 2C-143 (2P-7A).

d.

Defense in depth action only for 2CV-0795-2.

e.

No further actions for 2CV-0711-2.

f.

No further actions for 2CV-1039-1 and 2CV-1076-2.

g.

No further action for 2CV-1026-2 and 2CV-1037-1.

h.

No further actions for 2P-7B.

i.

No further actions for 2CV-1060-2 and 2CV-1090-2.

j.

No further actions for 2CV-1010-1 and 2CV-1040-1.

k.

No further actions for 2CV-1052.

l.

No further actions for 2CV-1066-1.

m.

No further actions for 2CV-1002.

n.

No further actions for 2P-75 and the Condensate pumps 2P-2A, 2P-2B, 2P-2C, and 2P-2D.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-131 Fire Area ID:

GG - Unit 2 North Electrical and Piping Penetration Area Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-11

Title:

ANO-2 Fire Area GG Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions The following equipment is recovered in the post transition baseline case:

2CV-0789-1 EFW pump condensate suction valve Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection system located in Fire Area GG was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights DID Maintained:

The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area.

No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area.

Safety Margin Maintained:

All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.

Comments:

None

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-142 Fire Area ID:

HH - Unit 2 General Area 354 Elevation Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2063-DD Sample Room No No No No No No No No No No No No 2072-R Upper Volume Control Tank Room, Lower Tank & Pump Room No Yes No No No No No Yes No Yes No Yes 2073-DD Access Room, Pump Room, Tank Room (2B62 & Resin Addition Room)

P Yes No No No No Yes Yes No Yes No Yes 2096-M Motor Control Center (2B63)

No Yes No No No No No Yes No Yes No Yes 2106-R Degasifier Vacuum Pump Room No Yes No No No No No Yes No Yes No Yes 2107-N Corridor (North of Stairway 2001)

No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-143 Fire Area ID:

HH - Unit 2 General Area 354 Elevation Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-12

Title:

ANO-2 Fire Area HH Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs.

Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

The following modifications are area specific and credited to reduce risk in this fire area:

2CV-1026-2 circuit modified to prevent spurious operation.

2CV-1076-2 circuit modified to prevent spurious operation.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection systems located in Fire Area HH, except Fire Zone 2063-DD, were credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-160 Fire Area ID:

JJ - Corridor Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-14

Title:

ANO-2 Fire Area JJ Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions The following equipment is recovered in the post transition baseline case:

2A-309 ESF switchgear breaker for offsite power 2CV-1036-2 EFW discharge valve 2CV-1075-1 EFW discharge valve 2PIS-0789 EFW condensate suction Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

The following modifications are area specific and credited to reduce risk in this fire area:

2A3 switchgear DC control power 2A-308 EDG supply breaker 2A-309 Offsite supply breaker 2A-310 crosstie breaker 2B-6 load center 2P-32A, 2P-32B, 2P-32C, and 2P-32D DC control power 2CV-1036-2 circuit modification to resolve IN 92-18 issue 2CV-1075-1 circuit modification to resolve IN 92-18 issue 2CV-4816 and 2CV4817 Letdown valve control circuit

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-161 Fire Area ID:

JJ - Corridor Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)

Summary (continued)

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position. The modifications to 2CV-1036-2 and 2CV-1075-1 will eliminate the IN 92-18 issue for this fire area.

Additional Fire Area Considerations The detection and suppression systems located in Fire Area JJ were credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights DID Maintained:

The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area. The available suppression will provide additional assurances that any unknown weaknesses or uncertainties will not adversely affect the ability to meet nuclear safety performance criteria.

No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area.

Safety Margin Maintained:

All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.

Comments:

None

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-162 Fire Area ID:

JJ - Corridor Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs VFDR ID:

JJ-01 VFDR:

Fire damage to control and power cables in the area may impact EFW functions resulting in the following:

a)

Loss of valves 2CV-1036-2 and 2CV-1075-1 (IN 92-18) which feed the credited SG B.

b)

Loss of valve 2CV-1038-2 which may result in flow diversion to SG A without instrumentation and ADV control capability.

c)

Loss of DC power to the engineered safeguards bus 2A-3 feeding EFW pump 2P-7B.

d)

Loss of MSIV 2CV-1060-2 isolation capability of the credited SG B.

e)

Spurious start of the non-credited EFW pump 2P-7A and loss of Control Room trip function.

f)

Loss of MSIV 2CV-1010-1 isolation capability of the non-credited SG A.

Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.

Disposition:

This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:

a)

Recovery action and modification associated with 2CV-1036-2 and 2CV-1075-1 (see Attachment S).

b)

No further actions are required for 2CV-1038-2.

c)

No further actions are required for 2P-7B.

d)

No further actions are required for 2CV-1060-2.

e)

No further actions are required for 2P-7A.

f)

No further actions are required for 2CV-1010-1.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-163 Fire Area ID:

JJ - Corridor Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDR ID:

JJ-02 VFDR:

Fire damage to control and power cables in the area may impact inventory control functions resulting in the following:

a)

Spurious opening of valve 2CV-5631-2 could result in RWT drain-down to containment sump.

b)

Spurious trip of the charging pumps 2P-36A and C(R) from a SIAS signal.

c)

Spurious start of the HPSI pumps 2P-89A, B and C(R) from a SIAS signal.

d)

Spurious Containment Spray Pumps 2P-35A and B start from a containment spray signal could result in a flow diversion and a loss of DC control power could result in a loss of Control Room trip capability.

e)

Loss of DC control circuit power to Pressurizer Heater banks #1 through #6 resulting in loss of trip capability from the Control Room.

f)

Multiple spurious operations of Pressurizer LTOP relief valves 2CV-4731-2 and 2CV-4740-2 may result in an RCS inventory loss path.

g)

Spurious opening of 2CV-5086-2 and 2CV-5084-1 (SDC RCS isolation) due to impact on power cables for theses high-low pressure interface valves may result in a loss of RCS inventory.

Loss of these functions could challenge the Inventory Control Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.

Disposition:

This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:

a)

Failure of 2CV-5631-2 is not a contributor to core damage sequences in the fire PRA and therefore is not risk significant.

b)

No further action required as 2P-36A and 2P-36C are components in the CVCS system. CVCS is only credited in the Fire PRA to mitigate non-fire induced ATWS scenarios.

c)

No further actions are required for 2P-89A, B and C(R).

d)

Failure of 2P-35A or 2P-35B is not a contributor to core damage sequences in the fire PRA and therefore is not risk significant.

e)

Pressurizer Heaters are associated with failures that may affect inventory control and is modeled in the Fire PRA for sequences that result in RCS inventory loss. The MSO expert panel concluded that spurious actuation of pressurizer heaters will not be a concern as documented within MSO report CALC-ANO2-FP-09-00016.

f)

No further actions are required for 2CV-4731-2 and 2CV-4740-2.

g)

No further actions are required for 2CV-5086-2 and 2CV-5084-1.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-173 Fire Area ID:

KK - South EDG Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2093-P South EDG Room P

Yes No No No No Yes Yes No Yes No Yes 2114-I EDG Air Intake Room No Yes No No No No No Yes No Yes No Yes 2115-I Boric Acid Makeup Tank Room No Yes No No No No No Yes No Yes No Yes P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-174 Fire Area ID:

KK - South EDG Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-15

Title:

ANO-2 Fire Area KK Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs.

Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection system located in Fire Area KK was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights DID Maintained:

The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area.

No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area.

Safety Margin Maintained:

All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.

Comments:

None

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-180 Fire Area ID:

MM - West Battery and DC Equipment Rooms Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-16

Title:

ANO-2 Fire Area MM Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions There are no recovery actions credited in this fire area to reduce the area risk or mitigate the risk of VFDRs.

Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection system located in Fire Area MM was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-182 Fire Area ID:

MM - West Battery and DC Equipment Rooms Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation VFDRs VFDR ID:

MM-01 VFDR:

Fire damage to cables in the area may impact EFW functions resulting in the following:

a)

Loss of DC control cables to valves 2CV-0205-2 (IN 92-18) and 2CV-0340-2 results in loss of steam flow to the EFW 2P-7A pump feeding the credited SG B.

b)

Loss of DC control cables to valve 2-CV-0711-2 (IN 92-18) and 2CV-0795-2 may result in isolation of SW and CST supply to EFW 2P-7A pump.

c)

Loss of DC control cable to valve 2CV-1076-2 results in isolation of the EFW flow to SG B.

d)

Loss of power and control cables to valve 2CV-1026-2 results in loss of control of EFW flow to the non-credited SG A.

e)

Spurious closure of valve 2CV-1039-1 results in isolation of the EFW flow to SG B.

Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.

Disposition:

This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:

a)

No further action is required for 2CV-0205-2 and 2CV-0340-2.

b)

No further action is required for 2-CV-0711-2 and 2CV-0795-2.

c)

No further action is required for 2CV-1076-2.

d)

No further action is required for 2CV-1026-2.

e)

No further action is required for 2CV-1039-1.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-193 Fire Area ID:

OO - Unit 2 Intake Structure Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-18

Title:

ANO-2 Fire Area OO Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions The following equipment is recovered in the post transition baseline case:

2CV-1470-1 sluice gate valve for 2P-4A 2CV-1474-2 sluice gate valve for 2P-4C Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The fire PRA quantification does not credit any suppression or detection systems for Fire Area OO.

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-198 Fire Area ID:

QQ - North Emergency Diesel Compliance Basis:

NFPA 805 Section 4.2.3.2 - Deterministic Approach Required Fire Protection Systems and Features Required?

Installed Separation LA EEEE Risk*

DID Fire Zone Fire Zone ID SUP DET SUP DET SUP DET SUP DET SUP DET SUP DET 2094-Q North EDG Room Y

Yes No No No No Yes Yes Yes Yes No No 2114-I EDG Air Intake Room No Yes No No No No No Yes No Yes No No

• Suppression and/or detection credited in HGL/MCA P - Indicates a partial system is installed.

Separation - Required for Chapter 4 Separation Criteria LA-Required for NRC-Approved Licensing Action EEEE-Required for Existing Engineering Equivalency Evaluation Risk - Required for Risk Significance DID-Required to Maintain an Adequate Balance of Defense-in-Depth in a Change Evaluation or FRE Risk Summary This fire area complies with the deterministic requirements of Section 4.2.3.2 of NFPA 805 and a FRE is not required.

VFDRs This fire area is in deterministic compliance and has no VFDRs.

End of Fire Area QQ

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-202 Fire Area ID:

SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary FRE Calculation:

CALC-09-E-0008-19

Title:

ANO-2 Fire Area SS Risk Evaluation Summary:

The fire risk evaluation has determined that the variances identified for this fire area are acceptable based upon the measured change in CDF and LERF, adequate defense in depth, and maintenance of safety margins with only the global modifications credited to reduce CDF and LERF in all ANO-2 fire areas. This fire area is compliant with the risk-informed, performance-based approach as the results of this fire risk evaluation meet the requirements of NFPA 805 and the guidance of RG 1.205.

Credited Recovery Actions There following equipment is recovered in the post transition baseline case:

2CV-1038-2 EFW discharge valve 2CV-1425-1 ACW isolation valve 2CV-1470-1 Sluice gate for 2P-4A Credited Modifications Listed below are plant modifications that are credited globally to reduce the area CDF and LERF for all Fire PRA scenarios:

New AFW source independent of existing EFW/AFW pumps.

Redundant DC power supply to 2A1, 2A2, 2H1, and 2H2.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-203 Fire Area ID:

SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)

Summary (continued)

The following modifications are area specific and credited to reduce risk in this fire area:

2A-4 circuit reroute modification 2A-409 circuit reroute modification 2B-6 circuit reroute modification 2P-32A, 2P-32B, 2P-32C, and 2P-32D DC control power 2A-3 DC control power and circuit reroute modification 2A-308 DC control power and circuit reroute modification 2A-309 DC control power and circuit reroute modification 2A-310 DC control power and circuit reroute modification 2CV-0789-1(2PIS-0789-1) circuit reroute modification 2CV-1040-1AC power source modification 2D-27 circuit reroute modification 2K-4A circuit reroute modification 2P-16A circuit reroute modification 2P-36A circuit reroute modification 2SV-0724-1 circuit reroute modification 2SV-2809-1 circuit reroute modification 2SV-2810-1 circuit reroute modification 2SV-2811 circuit reroute modification IN-92-18 Concerns There are no recovery actions credited in this fire area to manually position motor operated valves that may have spuriously operated due to fire and failed in a non-recoverable position.

Additional Fire Area Considerations The detection system located in Fire Area SS was credited in the ANO-2 Hot Gas Layer and Multi-Compartment analysis. The detection system is required to support fire brigade response to mitigate the formation of a hot gas layer.

Arkansas Nuclear One - Unit 2 Att. C - NEI 04-02 Table B-3 Fire Area Transition to 2CAN121202 Page C-204 Fire Area ID:

SS - South Switchgear and East DC Equipment and Battery Rooms Compliance Basis:

NFPA 805 Section 4.2.4.2 - Performance Based - Fire Risk Evaluation Risk Summary (continued)

CDF:

Refer to Attachment W Fire PRA Insights LERF:

Refer to Attachment W Fire PRA Insights DID Maintained:

The VFDRs, the associated fire area risks (CDF) and consequences (CCDP) were evaluated to identify general defense-in-depth echelon imbalances. This review is documented in Table 6.2.3 of the FRE and shows no additional DID methods are required beyond those inherent to the fire area.

No procedural changes, modifications, or recoveries are needed for maintenance of DID for this fire area.

Safety Margin Maintained:

All analyses and assessments have been performed utilizing accepted techniques and industry accepted standards that are specifically documented within the FRE calculation.

Comments:

None VFDRs VFDR ID:

SS-01 VFDR:

Fire damage to control and power cables in the area may impact EFW functions resulting in the following:

a)

Spurious start of the non-credited EFW pump 2P-7A and loss of Control Room trip function.

b)

Spurious closure of valve 2CV-1038-2 which feeds the credited SG A.

c)

Loss of remote operation from Control Room to credited EFW pump 2P-7B requires local operation from the switchgear.

Loss of these functions could challenge the Decay Heat Removal Performance Criterion. This condition represents a variance from the deterministic requirements of Section 4.2.3 of NFPA 805. This is a separation issue and evaluation of the additional risk is required in accordance with Section 4.2.4 of NFPA 805.

Disposition:

This VFDR has been evaluated and it was determined that the risk, safety margin, and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with the following actions:

a)

No further actions are required for 2P-7A.

b)

Recovery action associated with 2CV-1038-2.

c)

Modifications associated with switchgear (2A-3) to maintain control of 2P-7B (see Attachment S).

to 2CAN081401 Updated Attachment G, Table G-1 Recovery Actions and Activities

Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition to 2CAN121202 Page G-4 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS B-2 2A-309 2A-3 SUPPLY BREAKER De-energize DC Control Power to 2A-309 at 2A-3, 2A-309 located in Fire Area II, Fire Zone 2101-AA. Then verify open/

manually open 2A-309 in Fire Area II, Fire Zone 2101-AA.

B2-01 RA B-2 2P-32A REACTOR COOLANT PUMP Secure 2P-32A.

B2-03 RA B-2 2P-32B REACTOR COOLANT PUMP Secure 2P-32B.

B2-03 RA B-2 2P-32C REACTOR COOLANT PUMP Secure 2P-32C B2-03 RA B-2 2P-32D REACTOR COOLANT PUMP Secure 2P-32D.

B2-03 RA B-3 2CV-0789-1 EFW PUMP 2P-7B CONDENSATE SUCTION VALVE De-energize 2CV-0789-1 at panel 2B-5, breaker 2B-514, located in Fire Area II, Fire Zone 2101-AA. Verify open/

manually open 2CV-0789-1 in Fire Area FF, Fire Zone 2025-JJ.

B3-01 RA G

2CV-4816 LETDOWN THROTTLE CV Locally vent air from the actuator for 2CV-4816 to fail the valve closed in Fire Area EE, Fire Zone 2084-DD.

G-02 RA G

2CV-4817 LETDOWN THROTTLE CV Locally vent air from the actuator for 2CV-4817 to fail the valve closed in Fire Area EE, Fire Zone 2084-DD.

G-02 RA G

2P-32A REACTOR COOLANT PUMP De-energize DC Control Power to 2P-32A at 2H-11 located in Fire Area B-2, Fire Zone 2200-MM. Then verify tripped/

manually trip 2H-11 in Fire Area B-2, Fire Zone 2200-MM.

G-02 RA G

2P-32B REACTOR COOLANT PUMP De-energize DC Control Power to 2P-32B at 2H-21 located in Fire Area B-2, Fire Zone 2200-MM. Then verify tripped/

manually trip 2H-21 in Fire Area B-2, Fire Zone 2200-MM.

G-02 RA G

2P-32C REACTOR COOLANT PUMP De-energize DC Control Power to 2P-32C at 2H-22 located in Fire Area B-2, Fire Zone 2200-MM. Then verify tripped/

manually trip 2H-22 in Fire Area B-2, Fire Zone 2200-MM.

G-02 RA G

2P-32D REACTOR COOLANT PUMP De-energize DC Control Power to 2P-32D at 2H-12 located in Fire Area B-2, Fire Zone 2200-MM. Then verify tripped/

manually trip 2H-12 in Fire Area B-2, Fire Zone 2200-MM.

G-02 RA G

NEW AFW PUMP NEW AUXILIARY FEEDWATER (AFW) PUMP Manually start and align 2P-75B AFW pump locally to establish primary to secondary heat removal.

G-03 G-04 G-05 RA

Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition to 2CAN121202 Page G-5 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS G

2CV-1036-2 2CV-1075-1 2P-7B DISCHARGE TO SG-B 2P-7B DISCHARGE TO SG-B Locally open Emergency Feedwater discharge valves following power failure.

G-03 RA G

2P-89B HIGH PRESSURE SAFETY INJECTION (HPSI) PUMP Locally open breaker 2A-406 to prevent start of HPSI pump.

G-02 RA*

G 2P-60B LOW PRESSURE SAFETY INJECTION (LPSI) PUMP Locally open breaker 2A-405 to prevent start of LPSI pump.

N/A RA*

G 2P-35B CONTAINMENT SPRAY PUMP Locally open breaker 2A-404 to prevent start of Containment Spray Pump.

G-02 RA*

G 2CV-5630-1 2CV-5631-2 RWT OUTLET VALVES Close both RWT outlet valves locally.

G-02 RA*

G 2CV-4920-1 2CV-4921-1 BORIC ACID MAKEUP TANK (BAMT) GRAVITY FEED VALVES Open both BAMT Gravity Feed valves locally.

G-02 RA*

G 2CV-4873-1 VOLUME CONTROL TANK (VCT) OUTLET VALVE Close VCT outlet valve locally.

N/A RA*

G 2T-1 PRESSURIZER HEATERS Turn OFF and operate pressurizer heaters as necessary.

G-01 RA*

G 2P-36A/B/C CHARGING PUMPS Stop and operate Charging pumps locally as needed.

G-02 RA*

G 2K-4B EMERGENCY DIESEL GENERATOR #2 (EDG #2)

Place EDG #2 in LOCKOUT locally.

G-04 RA*

G 2A-4 4160V VITAL POWER SWITCHGEAR De-energize 2A-4 locally to prevent spurious operation.

Energize 2A-4 to restore power to vital equipment.

G-04 RA*

G 2B-6 480V VITAL POWER LOAD CENTER De-energize 2B-6 locally.

G-04 RA*

Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition to 2CAN121202 Page G-6 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS G

2D-24-2 2D-24-4 2D-24-6 2D-24-8 2D-24-9 2D-24-10 2C108 POWER SUPPLY 2A4 DC CONTROL POWER 2E21 POWER SUPPLY 2C108 POWER SUPPLY 2B6 DC CONTROL POWER HI POINT VENT PNL 2C336-2 Open breakers to remove DC power to various equipment.

G-02 G-04 RA*

G 2CV-4840-2 CHARGING HEADER ISOLATION Locally verify open Charging header isolation.

G-02 RA*

G 2CV-1504-2 EDG #2 SERVICE WATER OUTLET Locally verify open EDG #2 Service Water outlet.

G-05 RA*

G 2CV-4950-2 RWT SUCTION VALVE Verify RWT suction valve open for Charging capability if necessary.

G-02 RA*

G 2P-4C SERVICE WATER PUMP Align Loop 2 Service Water header locally.

G-05 RA*

G 2CV-0795-2 2P-7A SUCTION MOV Verify open 2P-7A Condensate suction MOV locally.

G-03 RA*

G 2VUC-30 SPDS COMPUTER ROOM COOLER Transfer SPDS Room Cooler 2VUC-30 to ANO-1 power source if required.

G-06 RA*

GG 2CV-0789-1 EFW PUMP 2P-7B CONDENSATE SUCTION De-energize 2CV-0789-1 at panel 2B-53, breaker 2B-53D2, located in Fire Area B-3, Fire Zone 2091-BB. Verify open/

manually open 2CV-0789-1 in Fire Area FF, Fire Zone 2025-JJ.

GG-02 RA JJ 2A-309 2A-3 SUPPLY BREAKER De-energize DC Control Power to 2A-309 at 2A-3 located in Fire Area II, Fire Zone 2101-AA. Verify closed/manually close 2A-309 in Fire Area II, Fire Zone 2101-AA.

JJ-04 RA JJ 2CV-1036-2 2CV-1075-1 2P-7B DISCHARGE TO SG-B Locally open EFW discharge valve following power failure.

2CV-1075-1 and 2CV-1036-2 located in Fire Area GG, Fire Zone 2081-HH.

JJ-01 RA JJ 2PIS-0789 EFW PUMP 2P-7B CONDENSATE SUCTION De-energize and manually open 2CV-0789-1 in Fire Area FF, Fire Zone 2025-JJ, prior to starting an EFW pump.

JJ-01 RA

Arkansas Nuclear One - Unit 2 Att. G - Recovery Actions Transition to 2CAN121202 Page G-7 Table G Recovery Actions and Activities Fire Area Component Component Description Actions VFDR RA/PCS OO 2CV-1470-1 SERVICE WATER (SW) TO 2P-4A De-energize 2CV-1470-1 at panel 2B-54, breaker 2B-54E4, located in Fire Area II, Fire Zone 2101-AA. Verify open/

manually open 2CV-1470-1 in Fire Area OO, Fire Zone INTAKE.

Note: Valve operator is installed external to the intake structure and not in the impacted area.

OO-01 RA OO 2CV-1474-2 SW TO 2P-4C De-energize 2CV-1474-2 at panel 2B-62, breaker 2B-62H3, located in Fire Area HH, Fire Zone 2073-DD. Verify open/

manually open 2CV-1474-2 in Fire Area OO, Fire Zone INTAKE.

Note: Valve operator is installed external to the intake structure and not in the impacted area.

OO-01 RA SS 2CV-1038-2 EFW FROM 2P-7B TO SG-A ISOLATION De-energize 2CV-1038-2 at panel 2B-63, breaker 2B-63H3, located in Fire Area HH, Fire Zone 2096-M. Verify open/

manually open 2CV-1038-2 in Fire Area EE-L, Fire Zone 2084-DD.

SS-01 RA SS 2CV-1425-1 AUXILIARY COOLING WATER (ACW) ISOLATION De-energize 2CV-1425-1 at panel 2B-54, breaker 2B-54D5, located in Fire Area II, Fire Zone 2101-AA. Verify closed/

manually close 2CV-1425-1 in Fire Area OO, Fire Zone INTAKE.

SS-05 RA SS 2CV-1470-1 SW TO 2P-4A De-energize 2CV-1470-1 at panel 2B-54, breaker 2B-54E4, located in Fire Area II, Fire Zone 2101-AA. Verify open/

manually open 2CV-1470-1 in Fire Area OO, Fire Zone INTAKE.

SS-05 RA TT 2A-309 2A-3 SUPPLY BREAKER De-energize DC Control Power to 2A-309 at 2A-3 located in Fire Area II, Fire Zone 2101-AA. Verify closed/manually close 2A-309 in Fire Area II, Fire Zone 2101-AA.

TT-01 RA TT 2CV-1036-2 EFW FROM 2P-7B TO SG-B ISOLATION De-energize 2CV-1036-2 at panel 2B-63, breaker 2B-63H1, located in Fire Area HH, Fire Zone 2096-M. Verify open/

manually open 2CV-1036-2 in Fire Area GG, Fire Zone 2081-HH.

TT-01 RA TT 2CV-1075-1 EFW FROM 2P-7B TO SG-B FLOW CONTROL VALVE De-energize 2CV-1075-1 at panel 2B-53, breaker 2B-53J2, located in Fire Area B-3, Fire Zone 2091-BB. Verify open/

manually open 2CV-1075-1 in Fire Area GG, Fire Zone 2081-HH.

TT-01 RA RA - Recovery Action RA* - Defense in Depth Measure PCS - Primary Control Station

to 2CAN081401 Updated Attachment S, Table S-1 Plant Modifications

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-1 S.

Plant Modifications and Items to be Completed During Implementation Table S-1, Plant Modifications, provided below includes a description of the modifications along with the following information:

A problem statement, Risk ranking of the modification, An indication if the modification is currently included in the FPRA, Compensatory measure in place, and A risk-informed characterization of the modification and compensatory measure.

The following ranking legend should be used when reviewing the table:

High = Modification which would have an impact on FPRA and affect multiple Fire Areas.

Med = Modification which would have an impact on FPRA and affect individual Fire Areas, or include IN 92-18 modifications.

Low = Modification which would have no or insignificant impact on risk.

Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-1 Med (PRA) 2 In Fire Area HH, a separation issue was identified on the EFW valves 2CV-1026-2 and 2CV-1076-2. During a fire induced circuit failure the feedwater valves may be impacted by a fire in Fire Zone 2096-M.

LAR Source:

Attachment C ANO plans to relocate interposing relays and affected cables associated with 2CV-1026-2 and 2CV-1076-2 from Fire Area HH, Fire Zone 2096-M, to the adjacent room in Fire Area G, Fire Zone 2098-C. Circuits for 2CV-1026-2 and 2CV-1076-2 are currently routed through Fire Area G and no new impacts will be generated by this modification.

Yes Yes This modification is specifically credited from a PRA perspective.

Modification reduces the risk in Fire Area HH of a fire induced circuit failure for EFW valves 2CV-1026-2 and 2CV-1076-2 in Fire Zone 2096-M.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-2 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-2 High (PRA) 2 In Fire Area JJ, a separation issue was identified that impacts the DC power cables control wiring on both trains. If a fire event occurred, this could result in the loss of equipment that would otherwise be available. Additional considerations are potential spurious operations at switchgear 2A-3 that may result in a loss of power to the safety bus.

LAR Source:

Attachment C ANO plans to modify the circuits as described to eliminate impacts in Fire Area JJ associated with these components.

2A-3, 2A-308, 2A-309, and 2A-310 - The red train 125V DC panel 2D-23 that supplies control power for 2A-3 and 2B-5 is planned for relocation to Fire Area MM from Fire Area JJ. Control power cables are planned to be rerouted using embedded conduits from Fire Area MM to Fire Area II to avoid Fire Areas JJ and SS.

This allows post-fire control of 2A-3 bus from the control room.

2CV-1036 Auxiliary relays 2CR1036A, B, C, and D are currently installed in MCC 2B-61 and are planned to be relocated to MCC 2B-63. This would eliminate cables that are routed through Fire Area JJ associated with this valve. This eliminates a loss of 2CV-1036-2 due to a fire in Fire Area JJ.

(continued)

Yes Yes This modification is specifically credited from a PRA perspective and affects multiple fire areas.

The modification limits the risk of a potential spurious operation and a loss of DC power to safety bus for switchgear 2A-3 due to a fire induced circuit failure.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-3 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-2 2CV-1075 The reroute of DC control power to bus 2A-3 and load-center 2B-5 listed above assures MCC 2B-53 remains available to power this valve. Control cables from 2C-17 to MCC 2B-53 are planned to be rerouted using an embedded conduit between Fire Area G and II to avoid Fire Areas JJ, SS, and TT.

New dedicated fuses are planned to be installed in 2C-17 for 2CV-1075-1 control relays so that failure of cables in scheme 2S113 will not impact 2CV-1075-1.

2B Cables are planned to be rerouted to control room panel 2C33-2 from 2B-6 using an embedded conduit between Fire Zone 2100-Z to the cable spreading room Fire Area G.

This eliminates an impact in Fire Area JJ.

2CV-4816 & 2CV-4817 - A reroute of cable 2I016N is planned by using embedded conduit C4080 that is located between Fire Area G (cable spreading room) to Fire Area EE-L. Cable 2I016N is also planned to be separately fused in panel 2C-09 to prevent failure due to a loss of cable 2I016P.

This eliminates circuit impacts in Fire Areas TT, JJ, and EE-U.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-4 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-3 High (PRA) 2 In Fire Area MM, fire induced circuit failure could impact DC power cables feeding circuit breakers at switchgear 2A-1, 2A-2, 2H-1, and 2H-2. The failure of 2A-1 and 2A-2 could prevent alignment to an offsite power source.

The failure of 2H-1 and 2H-2 could prevent tripping the reactor coolant pumps (RCPs) from the control room.

LAR Source:

Attachment C ANO plans to install backup DC control power to switchgear 2A-1, 2A-2, 2H-1 and 2H-2 with automatic transfer capability in the event the normal DC control power source is lost.

The new backup DC power source will be located completely within Fire Area B-2 in proximity to the switchgear either on elevation 372 or below at elevation 354. This eliminates impacts to switchgear DC control power due to a fire in any other ANO-2 fire area and allows tripping of the RCPs in those areas.

Inclusive in this modification will be changes to the control power circuits for switchgear 2H-1 and 2H-2 to allow tripping the RCPs in a scenario where a fire originates internally to a switchgear cubicle. This design will prevent fire damage to a load cubicle from disabling the ability to trip the line breakers and remove power to the RCPs. The opposite scenario where fire damages the line breakers would not prevent the RCP load breakers from being tripped. This modification will require the line and load breakers to be separately fused and fed as described:

(continued)

Yes Yes This modification is specifically credited from a PRA perspective and affects multiple fire areas.

Modification to install an alternate DC power source reduces the risk of a fire induced circuit failure to the DC power cables feeding RCP circuit breakers 2H-1 and 2H-2 which could prevent tripping the RCPs from the control room.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-5 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-3 2H Internal DC control wiring jumpers will be removed to isolate the line and load cubicles. The DC control power for line breakers 2H-13, 2H-14, and 2H-15 will be isolated from the DC control power for the load breakers 2H-10, 2H-11, and 2H-12.

2H Internal DC control wiring jumpers will be removed to isolate the line and load cubicles. The DC control power for line breakers 2H-23, 2H-24, and 2H-25 will be isolated from the DC control power for the load breakers 2H-20, 2H-21, and 2H-22.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-6 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-4 High (PRA) 2 In Fire Area TT, a separation issue was identified that impacts the power cables for EFW, chemical and volume control system (CVCS), and service water (SW) components.

LAR Source:

Attachment C ANO plans to modify the circuits as described to eliminate impacts in Fire Area TT associated with these components.

2CV-1036 Auxiliary relays 2CR1036A, B, C, and D are currently installed in MCC 2B-61 and are planned to be relocated to MCC 2B-63. This would also eliminate cables that are routed through Fire Area TT associated with this valve. This eliminates a loss of 2CV-1036-2 due to a fire in Fire Area TT.

2CV-1075 Cables for this valve between panels 2C-39 to 2C-17 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room. Control cables from 2C-17 to MCC 2B-53 are planned to be rerouted using an embedded conduit between Fire Area G and II to avoid Fire Areas JJ, SS, and TT. New dedicated fuses are planned for installation in 2C-17 for 2CV-1075-1 control relays so that failure of cables in scheme 2S113 will not impact 2CV-1075-1.

(continued)

Yes Yes This modification is specifically credited from a PRA perspective and affects multiple fire areas.

The modification reduces the risk of a fire induced circuit failure for EFW/CVCS/SW components and power cables (2B-5, 2CV-0789-1, 2CV-1036-2, 2CV-1075-1, 2CV-4816, 2CV-4817, and 2P-7B) in Fire Area TT.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-7 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-4 2P-7B - Cables for this pump between panels 2C-39 to 2C-17 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room. New conduits are also planned to be installed.

2CV-0789 Cables for this valve between panels 2C-39 to 2C-17 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room. Control cables from 2C-17 to MCC 2B-53 are planned to be rerouted using an embedded conduit between Fire Area G and II to avoid Fire Areas JJ and SS.

2CV-4816 & 2CV-4817 - A reroute of cable 2I016N is planned by using embedded conduit C4080 that goes between Fire Area G (cable spreading room) to Fire Area EE-L. Cable 2I016N is also planned to be separately fused in panel 2C-09 to prevent failure due to a loss of cable 2I016P of cable. This eliminates circuit impacts in Fire Areas TT, JJ, and EE-U.

(continued) 2B Cables for this load center between panels 2C-39 to 2C-33-1 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-8 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-4 2B Cables for this load center between panels 2C-39 to 2C-33-1 that are currently routed through Fire Area TT are planned to be rerouted to remain exclusively in the cable spreading room.

S1-5 High (PRA) 2 In Fire Area SS, a fire induced circuit failure could impact the DC power on both trains resulting in the loss of the various components.

LAR Source:

Attachment C ANO plans to modify the circuits as described to eliminate impacts in Fire Area SS associated with these components.

2A-3 and 2A-310 - The red train 125V DC panel 2D-23 that supplies control power for 2A-3 and 2B-5 is planned to be relocated from Fire Area JJ to Fire Area MM. Control power cables are planned to be rerouted using embedded conduits from Fire Area MM to Fire Area II to avoid Fire Areas JJ and SS. This allows post-fire control of 2A-3 bus from the control room.

2A-4, 2A-409, & 2B Cables are planned to be rerouted to control room panel 2C33-2 from 2A-4 and 2B-6 using an embedded conduit between Fire Zone 2100-Z to the cable spreading room Fire Area G. This eliminates an impact in Fire Zone 2097-X and Fire Area JJ.

(continued)

Yes Yes This modification is specifically credited from a PRA perspective and affects multiple fire areas.

The modification reduces the risk of a fire induced circuit failure that could result in the loss of DC power for both trains.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-9 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-5 The 125V DC control power from 2D-24 to 2A-4 is planned to be rerouted using a new conduit to avoid an impact against cables G2D2404A and B in Fire Zone 2097-X.

2CV-0789-1 & 2PIS-0789 The power cable for 2PIS-0789-1 (for 2CV-0789-1) is planned to be re-routed using an embedded conduit from Fire Area G to Fire Area II to avoid Fire Area SS.

2CV-1040 This valve is not directly impacted but is failed due to a loss of AC. The red train 125V DC panel 2D-23 that supplies control power for 2A-3 and 2B-5 is planned to be relocated from Fire Area JJ to Fire Area MM. Control power cables are planned to be routed using embedded conduits from Fire Area MM to Fire Area II to avoid Fire Areas JJ and SS. This assures 2CV-1040-1 will have a source of power and eliminates an impact in Fire Area SS.

(continued)

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-10 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-5 2A-308, 2A-309, 2D-27, 2K-4A, 2P-16A, 2P-36A, 2SV-0724-1, 2SV-2809-1, 2SV-2810-1, and 2SV-2811 - The cables associated with these components are planned to be re-routed to avoid Fire Area SS by using embedded conduits and as required the installation of a new raceway in Fire Area B-2 directly under Fire Area SS on elevation 372. The new raceway in Fire Area B-2 is planned to be installed above the vertical zone of influence for any postulated fire source. This eliminates impacts for 2A-308, 2A-309, 2D-27, 2K-4A, 2P-16A, 2P-36A, 2SV-0724-1, 2SV-2809-1, 2SV-2810-1, and 2SV-2811 in Fire Area SS.

S1-6 Med (92-18) 2 Motor Operated Valves (MOVs) will be modified to meet requirements per IN 92-18. The NPO assessment determined that any one of the RCS drop line valves can fail in a closed and unrecoverable position resulting in a loss of SDC.

LAR Source:

Attachment D NPO-RCS-SDC ANO plans to modify the control circuit for 2CV-5038-1 to prevent spurious closure.

This is planned to be similar to the inhibit circuit modification on CV-1275 for ANO-1.

Procedural controls to secure power by opening breakers are planned to be implemented for 2CV-5084-1 and 2CV-5086-2.

No Yes The NPO modification reduces the risk of fire induced MOV circuit failures (hot shorts, open circuits and short to ground). This MOV modification can prevent a non-recoverable position failure resulting in the loss of shutdown cooling.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-11 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-7 Med (PRA) 2 In Fire Area G, MOVs will be modified to meet requirements per IN 92-18.

Four EFW discharge valves can fail in an unrecoverable position.

LAR Source:

Attachment C ANO plans to modify the control circuit for MOVs 2CV-1075-1 and 2CV-1036-2 to prevent fire induced spurious operation from the main control room, Fire Area G. This will be accomplished by separating the cable conductors, inclusive of internal panel wiring, that can cause spurious valve operation and protecting them with suitable barriers to prevent inadvertent energizing of target conductors. This will prevent contact with potentially energized conductors from both intracable and intercable hot shorts.

MOV 2CV-1075-1 control cables R2B53J2C and R2B53J2N that enter panel 2C-17 or 2C-39 from floor penetrations have been identified as the cables of concern applicable to this modification.

MOV 2CV-1036-2 control cable G2B63H1E that enters panel 2C-40 from a floor penetration has been identified as the cable of concern applicable to this modification.

Yes Yes This modification is specifically credited from a PRA perspective.

The modification reduces the risk of fire induced MOV circuit failures (hot shorts, open circuits and short to ground). This MOV modification can prevent a non-recoverable position failure.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-12 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-8 High (PRA) 2 In Fire Areas B-3 and G, spurious opening of MOV 2CV-4698-1 pressurizer low temperature -

overpressure (LTOP) relief can result from a fire in motor control center (MCC) 2D-27 and cabinet 2C-09.

LAR Source:

Attachment C ANO plans to modify the control circuit for 2CV-4698-1 to prevent fire induced spurious opening in Fire Areas B-3 and G. This will be accomplished by separating the cable conductors, inclusive of internal panel wiring, that can cause spurious opening and protecting the conductors with suitable barriers to prevent inadvertent energizing of target conductors. This will prevent contact with potentially energized conductors from both intracable and intercable hot shorts. Control cable R2D27A3J that enters MCC 2D-27 in Fire Area B-3 and the other end of cable that enters cabinet 2C-09 in Fire Area G from the floor penetrations have been identified as the cable of concern applicable to this modification.

Yes Yes This modification is specifically credited from a PRA perspective.

The modification in Fire Areas B-3 and G protects the valve control cable in MCC 2D-27 and cabinet 2C-09 which reduces the risk of fire induced circuit failures (such as spurious opening). This modification can prevent a non-recoverable position failure.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-13 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-9 High (PRA) 2 RCS High Point Vent Valves 2SV-4670-2 and 2SV-4669-1:

Solenoid valves are provided in the RCS system for a means of venting.

These energize to open solenoids provide a vent path from the top of the pressurizer. The final two, in parallel, solenoids determine the vent path, either to the containment atmosphere (2SV-4670-2) or the quench tank (2SV-4669-1).

In Fire Area G, PRA determined that solenoid valves 2SV-4670-2 and 2SV-4669-1 control circuits shall be protected with metallic sleeves and/or barriers to prevent a spuriously opening of 2SV-4670-2 or 2SV-4669-1 due to a fire induced hot short circuit failure resulting in an uncontrolled RCS vent path release.

LAR Source:

Attachment C ANO plans to modify the RCS vent solenoid valve control circuits with the installation of metallic sleeves and/or barriers as described to eliminate impacts in Fire Area G associated with the following components:

2SV-4670 Control circuit in cabinet 2C-336-2 modification is planned to prevent conductor (wire P3) of cable G2SI122N from contacting energized conductors by installing a grounded metallic sleeve and/or barriers up to load side of hand-switch 2HS-4670-2.

2SV-4669 Control circuit in cabinet 2C-336-1 modification is planned to prevent conductor (wire P3) of cable R2SI121N from contacting energized conductors by installing a grounded metallic sleeve and/or barriers up to load side of hand-switch 2HS-4669-1.

Yes Yes This modification is specifically credited from a PRA perspective.

The modification in Fire Area G to install grounded metallic sleeves and/or barriers protects control circuit cable G2SI122E in cabinet 2C-336-2 and control circuit cable R2SI121E in cabinet 2C-336-1 to reduce the risk of fire induced circuit failures (such as spurious opening).

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-14 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-10 Med (PRA) 2 In Fire Area B-4 an incipient fire detection system is not installed in control element drive mechanism (CEDM) room panels 2C-70, 2C-71, 2C-72, 2C-73, 2C-75, 2C-80, and 2C-409.

However an early warning fire detection system in accordance with NFPA 72, Fire Alarm Detection Code, is required by the PRA in accordance with FRE CALC-09-E-0008-05.

LAR Source:

Attachment C ANO plans to provide a modification in the CEDM room in Fire Area B-4 to install incipient detection in cabinets 2C-70, 2C-71, 2C-72, 2C-73, 2C-75, 2C-80, and 2C-409.

Fire detection signal cable is planned to be routed from each air sampling detector to the control room fire panel 2C-343-3.

Yes Yes This modification is specifically credited from a PRA perspective.

The early warning fire detection system modification in Fire Area B-4 reduces the risk of a fire induced circuit and equipment failures that could result in the loss of CEDM room panels 2C-70, 2C-71, 2C-72, 2C-73, 2C-75, 2C-80, and 2C-409.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

S1-11 High (PRA) 2 At ANO the availability of feedwater to ANO-2 SGs was identified as an issue by PRA.

Also identified by PRA was ANOs inability to perform high risk and time sensitive actions, such as control of auxiliary feedwater (AFW), outside of the ANO-2 Control Room.

LAR Source:

Attachment C (listed globally in multiple Fire Areas for new AFW pump in Risk Summaries)

ANO plans to install a new AFW pump capable of feeding either of the ANO-2 SGs.

The AFW would be designed to meet or exceed the flow requirements of ANO-2 Emergency Feedwater (EFW)

Pump 2P-7B (380 gpm @

1100 psig).

The new pump, controls and motor operated valves would be designed to be installed in a manner that protects the assumptions in the PRA. The preferred source of suction for the new pump is planned to be from an available source (i.e.

Condensate Storage Tank).

(continued)

Yes Yes The AFW modification is specifically credited from a PRA perspective to provide a reliable additional source of feedwater.

The local control panel modification is specifically credited from a PRA perspective to provide an alternate means to perform required actions outside the ANO-2 Control Room.

This modification reduces the risk of not being able to perform necessary operator actions to shutdown the plant, if either Control Room cant be manned.

(continued)

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-15 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-11 The discharge piping is planned to be routed through the Turbine Building to ANO-2 Auxiliary Building Rooms 2081 and 2084 for the tie-ins to the EFW System piping. The AFW tie-ins are planned to discharge into the ANO-2 EFW downstream of all EFW injection valves to ensure a single area fire does not disable AFW.

The AFW pump would be designed to have the capability to be operated from the ANO-2 Control Room and also locally.

The design will ensure electrical isolation from Control Room functions to prevent a fire in the ANO-2 Control Room from affecting local control of AFW components.

The AFW pump and associated motor operated valves would be designed to be powered by diverse non-safety related power sources to prevent a single failure from disabling equipment operation.

The AFW pump would be designed to include controls and monitoring instrumentation to ensure proper water flow to the SGs. The local controls and monitoring instrumentation are planned to be located and powered with redundant power supplies in a manner that protects the assumptions in the PRA.

Also, the local control panel modification reduces the risk of availability issue with of feedwater supply to the ANO-2 SGs.

Manual actions are credited in fire areas that contain redundant safe shutdown equipment. These actions have been demonstrated feasible and are therefore considered adequate compensatory measures until compliance can be achieved by transitioning to a 10CFR50.48(c) licensing basis.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-16 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-12 Med (PRA) 2 In Fire Area B-3, excessive temperatures have been identified in Fire Zone 2091-BB following a loss of ventilation.

LAR Source:

Attachment C ANO plans to modify the control wiring for fans 2VEF-63 and 2VEF-64 to isolate the control room and allow the local controls to override a stop signal generated from within Fire Area G, either from handswitch positioning or fire-induced circuit damage. This eliminates fire impacts in Fire Area G and assures either 2VEF-63 or 2VEF-64 will remain available except for a fire in Fire Area B-3, Fire Zone 2091-BB.

Yes No This modification supports a basic assumption from a PRA perspective.

S1-13 Med (PRA) 2 In Fire Area MM, excessive temperatures have been identified in Fire Zone 2099-W following a loss of ventilation.

LAR Source:

Attachment C ANO plans to provide a modification to fire door DR 265 to allow normally open positioning with automatic closure features in the event of a fire. This allows natural circulation to prevent long term room overheating impact on equipment located in Fire Zone 2099-W, West DC Equipment Room, by allowing an opening to Fire Zone 2109-U, Corridor, in Fire Area JJ.

Yes No This modification supports a basic assumption from a PRA perspective.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-17 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-14 Low (Code)

C With regard to NFPA 50A, Gaseous Hydrogen Systems, code non-compliance issues were identified in the Hydrogen Gas Bottle Storage Room related to inadequate vent piping and room ventilation. The hydrogen storage room light switch was identified as not meeting Article 501 for Class I, Division 2 locations of the National Electric Code (NEC).

LAR Source:

Attachment A, Section 3.3.7.1 ANO plans to modify the ventilation for the Hydrogen Bottle Storage area to ensure compliance with NFPA 50A.

In addition, electrical equipment and wiring changes will be made to meet the requirements of NFPA 70 (NEC), Article 501 for Class I, Division 2.

No No The subject hydrogen gas system bottle storage area is not credited by the PRA.

This modification will be completed to meet NFPA 805 code requirements.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-18 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-15 Med (PRA)

C NFPA 805 non-compliance issues were encountered when smaller fire areas were defined such that multiple walls, dampers, penetration seals, and doors were credited and used in the PRA model as rated fire barriers in the NRC regulatory basis for NFPA 805.

Multiple walls and doors barriers will require upgrading to comply with NFPA 805.

LAR Source:

Attachment A, Section 3.11.2 ANO plans to provide an adequate-for-the-hazard evaluation and if necessary a modification to upgrade fire barrier walls, dampers, penetration seals, and doors to rated barriers for those barriers credited for deterministic compliance and subsequently credited in the Fire PRA analysis.

These barriers have been previously identified as NRC regulatory basis to ensure compliance with NFPA 805 and have compensatory measures established. The barriers to be addressed as identified by EC-1956 are 2005-2, 2005-3, 2067-4, 2082-3, 2091-1, 2091-2, 2091-3, 2091-4, 2107-4, 2110-2, 2110-4, 2110-7, 2112-2, 2112-8, 2112-10, 2133-5, 2133-6, 2147-8, 2148-4, 2148-5, 2149-5, 2152-2, 2154-2, 2154-3, 2154-5, 2158-10, 2224-2, 2224-3, 2228-10, 2239-4, 2239-5, 2256-4, 2256-5, 2256-6, 2256-8, 2134-1, and 2155-1.

Yes Yes This modification will be completed to meet NFPA 805 code requirements.

In accordance with station directives, compensatory measures per OP-1003.014 have been established as appropriate.

Arkansas Nuclear One - Unit 2 Att. S - Plant Modifications and Items to be Completed to 2CAN121202 Page S-19 Table S-1 Plant Modifications Item Rank Unit Problem Statement Proposed Modification In FPRA Comp Measure Risk Informed Characterization S1-16 Low (Code)

C NFPA 10 non-compliance issues (such as incorrect number of fire extinguishers for travel distance, incorrect type and size for the hazard area) were identified with ANO portable fire extinguishers.

LAR Source:

Attachment A, Section 3.7 ANO plans to provide a modification to resolve the NFPA 10 code deficiencies identified in CALC-ANOC-FP-09-00009.

In general, this modification would involve portable fire extinguisher physical relocation, substitution of existing extinguishers, and documentation updates to reflect these plant changes.

The results will ensure the proper number of fire extinguishers to meet travel distance requirements in coverage areas, adequately sized fire extinguishers, and the correct type of extinguisher that is rated for the fire hazard in each area.

No No The subject fire extinguishers are not credited in the FPRA.

This modification will be completed to meet NFPA 805 code requirements.

to 2CAN081401 Updated Attachment V Fire PRA Quality

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-1 V.

Fire PRA Quality V.1 ANO-2 Fire PRA Quality Review In accordance with RG 1.205 position 4.3:

The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard supporting requirements important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable. Licensees should justify use of Capability Category I for specific supporting requirements in their NFPA 805 risk assessments, if they contend that it is adequate for the application. Licensees should also evaluate whether portions of the PRA need to meet Capability Category III, as described in the PRA Standard.

The ANO-2 Fire PRA (FPRA) has undergone a RG 1.200, Revision 2, Peer Review against the ASME PRA Supporting Requirements (SRs) by a team of knowledgeable industry (vendor and utility) personnel. The review was conducted by the Westinghouse Owners Group in June 2009 under LTR-RAM-II-09-046, Fire PRA Peer Review against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the Arkansas Nuclear One, Unit 2 Fire Probabilistic Risk Assessment. The conclusion of the review was that the ANO-2 FPRA methodologies being used were appropriate and sufficient to satisfy the ASME/ANS PRA Standard RA-Sa-2009. The review team also noted that the staff appeared to be applying the NUREG/CR-6850 methodologies correctly.

The ANO-2 FPRA has also been subject to three additional focus scope peer reviews. During the evolution of the NFPA 805 project, some changes to the fire scenario methodologies were applied to both refine the model and results, and to more fully comply with approved methods.

The first of the focused reviews primarily evaluated changes associated with fire scenario selection (FSS) technical elements. Two ignition frequency (IGN) elements were also studied.

This focused-scope peer review was conducted by the Westinghouse Owners Group in October 2011 under LTR-RAM-I-11-064, Focused Scope Fire PRA Peer Review for Arkansas Nuclear One Unit 2.

The second additional peer review focused solely on FSS technical elements. This review was performedin November 2012 by Kazarians & Associates (Mardy Kazarians) and is documented in 5384.R02.121122.DRAFT B, Focused Scope Peer Review ANO-2 Fire PRA, FSS-A, C, D, E and H. None of the other FPRA SRs involved changes in the Probabilistic Safety Assessment (PSA) Methodology as defined in ASME/ANS RA-Sa-2009.

The third focused scope peer review was performed by Jan Grobbelaar and Kaydee Kohlhepp of Curtiss-Wright Scientech in June 2014. The scope of this peer review was limited to high level requirements HLR-HRA-C and HR-G as applicable to fire Human Reliability Analysis (HRA). This peer review found that the ANO-2 fire HRA was performed consistent with the guidance set forth in NUREG/CR-1921, and specifically with NUREG/CR-1921, Attachment B Detailed Quantification of Fire Human Failure Events Using the EPRI Fire HRA Methodology.

The summary of the original peer review findings exhibited the following statistics for the evaluation of elements to the combined PRA Standard. For the ANO-2 FPRA, 81.3% of the SRs were assessed at Capability Category II or higher, including 7.6% of the SRs being

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-2 assessed at Capability Category III. The ANO-2 FPRA had an additional 3.5% of the applicable SRs assessed at the Capability Category I (CC-1) level. The FPRA was found to not meet 15.3% of the applicable SRs.

The Westinghouse Peer Group concluded that the ANO-2 FPRA is consistent with the ASME/ANS PRA Standard and supports risk-informed applications.

This attachment provides a detailed assessment of each of the findings identified by the Peer Review team. Table V-1 lists all findings of the original peer review and provides the ANO-2 disposition of each finding. Table V-2 lists each focus scope review finding and provides the ANO-2 disposition of each finding. Table V-3 lists those SRs that were assessed as CC-I and provides the disposition for CC-I acceptability in the NFPA application.

The Peer Review Reports will be available for NRC review via the CERTREC portal.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-3 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition CF-A1-01 Incorrect probabilities used in FRANC AlteredEvents Table Closed The FRANC AlteredEvents Table, Appendix F in the ANO-2 Scenario Report (ERIN doc. # 0247-06-0006.05 Rev. 0) appears to have incorrect probabilities in the following instances:

1) Events related to the letdown flow control valves 2CV4816

& 2CV4817 apparently used the wrong probability from NUREG/CR-6850 Table 10-2 for fire zones 2098-C &

2108-S. In these fire zones, database xx2.mdb shows that the applicable cables (2I016Q and 2I016N/P, respectively) are routed through trays instead of conduits as apparently assumed in Appendix F. NUREG/CR-6850 Table 10-2 allows probabilities in the range of 0.02 to 0.1 for inter-cable shorts in trays rather than the 0.01 cited in Appendix F which would be acceptable if the cables were in conduits.

2) Appendix F states the 0.0006 probability used for event YMP2P35BFF in fire zones 2100-Z and 2109-U-C comes from the HFE AHF2CSAS1P. In the HRA toolbox spreadsheet hfe_cp New Fire PRA HFEs (4-23-09).xlsm provided shows that the HEP for this HFE is actually 3.2E-4.

3) Appendix F states the 0.22 probability used for event EB12A3XXXF in fire zone 2100-Z comes from the HFE EHF2A309XP. In the HRA toolbox spreadsheet hfe_cp New Fire PRA HFEs (4-23-09).xlsm provided shows that the HEP for this HFE is actually 2.2E-2.

Use of incorrect probabilities in the FRANC AlteredEvents Table would introduce errors in results obtained using FRANC.

Correct errors in probabilities used or document justification for values currently found in the FRANC AlteredEvents Table.

The altered events table was revised to correct, and provide a summary of the basis for, the assigned probabilities.

1) Spurious actuation of these valves requires two, concurrent, proper polarity shorts to another 4-20mA cable.

Table 10-2 of NUREG/CR-6850 provides a probability range of 0.02 to 0.1 for an inter-cable short from one multi-conductor cable to another in a tray for a DC circuit.

Assuming an average probability of 0.06 for each short, the overall probability is 0.06

• 0.06 or 0.0036. This discussion is documented in Section 12 of the Fire Scenarios Report (PRA-A2-05-003).

2) YMP2P35BF was originally included in the model as an MSO for depletion of the RWT. However, it has since been removed due to the length of time required to deplete the RWT, the operator cues for rapid RWT depletion, reactor building (RB) Spray pump operation, and the alignment to the sump, if required.

Therefore, this event is removed from the Altered Events list.

3) EHF2A309XP has been added to the fault tree. Therefore, the altered event for EB12A3XXXF has been removed.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-4 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition CF-B1-01 Other Affected SR FSS-A3, HRA-B1 Problems with AlteredEvents Table &

Appendix F in the ANO-2 Scenario Report Closed The FRANC AlteredEvents Table, Appendix F, in the ANO2 Scenario Report (ERIN doc. # 0247-06-0006.05 Rev. 0) makes recurring reference to ANO Engineering Change EC13540, "ANO2 CABLE ROUTING EXCLUSIONS TO SUPPORT FIRE PRA FOR NFPA-805" to document that cables associated with a given event do not go through the cited fire zone. A review of EC13540 shows that it does identify the component and associated cables, but documents different fire zones than those listed in Appendix F. For example, events QAV200798C and QSV200798C for valve 2CV-0798-1 are assigned a probability of zero because the cables for that valve do not pass through fire zones 2007-LL and 2024-JJ. EC13540 shows that cables for valve 2CV 0798-1 do not pass through fire zone 2101-AA without any mention of fire zones 2007-LL and 2024-JJ. A review of PDMS data shown in database file xx2.mdb confirm that 2CV-0798-1 cables do not pass through fire zones 2007-LL and 2024-JJ, but EC13540 cannot be used as a reference to support that conclusion. This is applicable to almost all of the events set to zero with reference to EC13540.

In three instances, an event was incorrectly set to zero based on EC13540 when cables associated with the component actually do pass through the cited fire zone based on PDMS information found in database xx2.mdb:

1) Event QSV200798R for valve 2SV-0798-1 has an associated cable R2D2301A identified in EC13540 passing through fire zone 2099-W in tray EC122.

2) Event PMV210401R for valve 2CV-1040-1 has an associated cable R2B53B1E identified in EC13540 passing through fire zone 2200-MM in tray EC114.

3) Event ECB2A409XD for breaker 2A409 has associated cables 2A409A thru H and J thru M identified in EC13540 passing through fire zone 2200-MM in trays DA020, 030 thru 035, 037 & 039.

(continued)

The altered events table in Attachment F of the Fire Scenarios Report (PRA-A2-05-003) has been revised so that the reference to EC13540, ANO-2 Cable Routing Exclusions to Support Fire PRA for NFPA-805, has been removed for the items mentioned in the finding. Where EC 13540 is referenced, the events can be tied to the EC directly. Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301)

(ML13312A877).

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-5 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition CF-B1-01 (continued)

Event ECB2A409XR for breaker 2A409 has a probability of 0.1 assigned for fire zone 2200-MM-B while two other events for the same breaker in that fire zone are assigned probabilities of 0. Note that EC13540 does not address this fire zone at all.

Inadequate justification provided for setting numerous event probabilities to zero and in three instances event probabilities inappropriately set equal to zero. Use of incorrect probabilities in the FRANC AlteredEvents Table would introduce errors in results obtained using FRANC.

Revise EC13540 to address the fire zones listed in the FRANC AlteredEvents Table and fix probabilities where re-analysis of cable routing shows that the cables actually do pass through the cited fire zones.

CS-A7-01 Other Affected SR ES-B1, FQ-D1 LERF/

Containment Bypass Closed Although cables are considered for one containment bypass event (MOV 2CV-3200-2 TRANSFERS OPEN), there are no other containment isolation failures considered in the LERF model. The containment purge system is not considered (it is screened from the internal events model) and other containment bypass paths may exist, but are not identified.

There is currently an open internal event PRA F&O (LE-D6-01) on the containment isolation system, expressing no confidence that the analysis represents a realistic assessment.

Cable and circuit failure modes affecting containment bypass are not completely considered or dispositioned in the Fire PRA plant response model for LERF.

Review potential containment bypass pathways beyond those in the current internal events PRA, including containment purge, that could lead to a LERF.

A review of the containment piping and venting paths were reviewed for impact on large early release. No LERF paths were identified for the internal events. Only two potential LERF paths were identified due to multiple spurious operations of the isolation valves: the containment purge supply and return lines (penetrations 2V-1 and 2V-2).

However, each of these lines includes two AOVs and an MOV for isolation. The valves are in series with different power supplies (i.e, red and green powered), have independent circuits, and power supplies that are separated by fire zone with the exception of the control room. The controls within the control room divided between two cabinets (2C16 and 2C17). These cabinets have sufficient separation and barriers to prevent a fire from spreading from one cabinet to another.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-6 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition CS-A8-01 Other Affected SR FSS-C5 Component and Cable Selection Report Thermoplastic Cable Closed ER-ANO-2003-0450-000 Rev 0 is credited in the self-review as determining that only 1% of cables at ANO-2 are thermoplastic and therefore of minimal significance. This ANO-generated document is not sufficiently complete to justify this conclusion. This engineering reply document states that 380 thermoplastic cables were identified in PDMS, but listed only 10 specific cables as being in "safety significant fire zones". A brief review of safe shutdown components conducted with ANO engineering personnel identified a safe shutdown neutron monitoring system as having thermoplastic cables which is not mentioned in the engineering reply document cited above. A complete plant-wide review of thermoplastic cable used for safe-shutdown components should be performed, documented and referenced in the Fire PRA Component and Cable Selection Report.

SR requirement to include treatment of thermoplastic cable failure was not satisfied.

CALC-ANOC-FP-09-00019, Rev. 0, EC-6964, Safe Shutdown Cable Jacket Insulation Types at ANO, evaluates the cables at both units and concludes that of the over 4600 cables reviewed, less that 0.3%

have thermoplastic insulation. This calculation also confirms that thermoplastic cables are not used in power supply circuits.

Thus, thermoplastic insulation at ANO is of minimal significance. CALC-ANOC-FP 0019 is referenced in the Fire PRA Scenarios Report.

The resolution of ANO-2 CS-A8-01 was reviewed and accepted by the ANO-1 Peer Review Team.

CS-A9-01 Other Affected CS-A5 DC Circuits Proper Polarity Hot Shorts on Ungrounded DC Circuits Closed The ANO Fire PRA guidance was contradictory with respect to cable selection with respect to proper polarity hot shorts on ungrounded dc circuits, and it was unclear whether these hot shorts were appropriately considered.

"Safe Shutdown Cable Analysis," Calculation Number 85-E-0087-24, Revision 1 was designated as the reference for the scope. Section 4.2.5 states: "For ungrounded DC circuits, two hot shorts of the proper polarity (without grounding) causing spurious operation is not considered credible except for high-low pressure interface components." However, the ANO-2 self-assessment for CS-A9 states: "Postulated proper polarity hot shorts on ungrounded DC circuits (e.g., RCS head vents) were considered as documented in the PDMS database."

Further, in CALC-85-E0087-24, Safe Shutdown Cable Analysis, conflicting criteria was identified in Section 4.2 of the calculation. Criteria 4.2.2 states that all DC grounded and ungrounded circuits must consider any and all shorts, hot shorts, shorts-to ground and open circuits. Criteria 4.2.3 states that all ungrounded circuits (both AC and DC) will be analyzed as if the circuit is grounded to account for the possibility of experiencing a ground fault.

(continued)

The F&O indicates that Criteria 4.2.2, 4.2.3, and 4.2.5 in CALC-85-E-0087-24 are contradictory and that this guidance for cable analysis should be changed. However, the criteria in CALC-85-E-0087-24 come from NUREG-1778 and Generic Letter 86-10, and provide a consistent approach to circuit analysis. This approach keeps the analyst from making false assumptions based upon a single failure safety analysis instead of more appropriately considering multiple concurrent failures expected as the result of a fire. The criteria in question are concerned with DC circuit analysis and ungrounded AC circuits which are similar to ungrounded DC in how they respond to circuit failures. The 125 VDC power and control circuits at ANO are ungrounded and typically only shields in low voltage DC instrumentation loops are grounded.

(continued)

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-7 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition CS-A9-01 (continued)

However, criteria 4.2.5 then states that for ungrounded DC circuits, two hot shorts of proper polarity (without grounding) are not considered credible except for high-low pressure interface components. Criteria 4.2.5 conflicts with 4.2.2 and 4.2.3 for several reasons: (a) 4.2.2 criterion requires that any and all hot shorts be considered, and (b) 4.2.3 criterion requires that ungrounded dc circuits be grounded, and it would only take one hot short of proper polarity from the same DC source to spuriously operate the high/low pressure interface.

These criteria need to be aligned with each other.

Additional contradictory evidence is found in the "NFPA-805 Fire PRA Modeling of Multiple Spurious Operations (MSO)"

document CALC-ANO2-FP-09-00016, Rev. 0. For example, the disposition of PWR MSO 10 states that inter-cable hot shorts are not postulated due to the use of thermoset cables.

The guidance and discussion of proper polarity hot shorts on ungrounded dc circuits is contradictory and the correct disposition of this requirement could not be determined.

Clarify the position on proper polarity hot shorts in ungrounded dc circuits and document that the ANO-2 Fire PRA includes consideration of proper polarity hot shorts on ungrounded dc circuits, besides high-low interface components.

Not grounding DC systems allows a single fault on either the positive or negative side of the circuit to occur without jeopardizing the function of the DC system. The criteria used for ungrounded systems provide a methodical approach as discussed below.

Criterion 4.2.2 of CALC-85-E-0087-24 is a definition of the scope of DC circuits considered and prevents the exclusion of any circuit associated with safe shutdown equipment without engineering review/analysis.

Criterion 4.2.3 of CALC-85-E-0087-24 is a simplified methodology to DC circuit analysis.

Since safe shutdown deals with multiple failures it is possible to have failure in a separate circuit create the initial ground fault.

Requiring that all ungrounded circuits (both AC and DC) be analyzed as if the circuit is grounded simplifies the analysis to determine circuit failure. A single short/fault in the circuit being reviewed can result in failure since a ground is already assumed to be established. This criterion requires all circuits, grounded and ungrounded, to be evaluated in the same manner.

Criterion 4.2.5 of CALC-85-E-0087-24 will allow an exclusion of a specific circumstance where a cable in the subject circuit faults to a second cable (two hot shorts of proper polarity). This unique intercable short requires that it be of the appropriate DC voltage, proper polarity (positive to positive/negative to negative), occurs exclusive of any fire induced grounds, and not be a high-low pressure boundary. This criterion is the DC equivalent of Criterion 4.2.4 which is for 3-phase AC.

(continued)

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-8 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition CS-A9-01 (continued)

To summarize, Criterion 4.2.2 provides a baseline, Criterion 4.2.3 a simplified methodology, and Criterion 4.2.5 is an exclusion for a specific set of circumstances.

Therefore, the criteria are consistent with each other and no revision to CALC-85-E-0087-24 is necessary.

Also, the F&O indicates that additional contradictory evidence is found in the "NFPA-805 Fire PRA Modeling of Multiple Spurious Operations (MSO)" document CALC-ANO2-FP-09-00016, Rev. 0. For example, the disposition of PWR MSO 10 states that inter-cable hot shorts are not postulated due to the use of thermoset cables. This inaccurate wording has been removed from CALC-ANO2-FP-09-00016.

Inter-cable hot shorts are postulated in the fire PRA model and NSCA as described in the MSO report.

CS-C4-01 Cable Selection/

Overcurrent Protection Closed Electrical distribution system over-current coordination and protection analysis was performed using the methodology of ANO Upper Level Document "Electrical Protection/

Coordination", ULD-0-TOP-12 Revision 3, 11/25/2002. This upper level document references the ANO2 plant specific analysis found in engineering calculation 84-E-0103-01, Revision 8, "General Criteria for Safety Bus", 01/21/2000.

No references or discussion of this review is provided in the fire PRA documentation. Provide in the appropriate portion(s) of the fire PRA documentation references to the review of electrical distribution system over-current coordination and protection analysis.

References to the review of electrical distribution system over-current coordination and protection analysis are required in Fire PRA documentation to satisfy this SR.

Provide in the appropriate portion(s) of the Fire PRA documentation references to the review of electrical distribution system over-current coordination and protection analysis.

Section 4.4 of the Component and Cable Selection Report, provides documentation that all circuits and electrical distribution buses credited in the fire PRA have been analyzed for proper over-current coordination and protection. A description of the processes used is included via reference to ULD-0-TOP-12, Upper Level Document ANO-1 and ANO-2 Electrical Protection/

Coordination, Rev. 3. Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301)

(ML13312A877).

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-9 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition ES-C1-01 HRA Instrumentation Closed Task 12, Post Fire HRA was reviewed for detailed analysis performed for credited operator actions. For many of the recovery actions control room instrumentation and annunciators were credited as cues. In many cases, these instruments and annunciators (along with supporting cables and power supplies) were not added to the equipment list and validated for function. In some cases, the cues stated may not be necessary since it is expected that the operator takes the manual actions regardless of cues stated. In other cases, the operators would not know the actions are required unless the cues were available. Example of instrumentation needed:

RHGISORCPBO (Operators Fail to Isolate RCP Bleed-off) credits RCP bleed off temperature from annunciator 2K11 for knowledge that spurious operation has occurred. Example of instrumentation (annunciators) credited, though may not be needed include EHF2A309XP (Operator fails to locally close 2A-309 to restore off-site power to 2A-3). This action credits annunciators for low bus voltage and EDG (Emergency Diesel Generator) trouble. However, it appears that the step to verify 2A1 energized from Off Site power would be followed if verification could not take place due to loss of instrumentation.

For each HFE, check credited instrumentation and add as appropriate to the equipment list. Ensure satisfaction of ES-B4 such that supporting power supplies are included in equipment/cable selection.

At the time of the peer review, most of the Post Fire HRAs included information related to the instrumentation for operator cues in the Fire Scenarios report. Since the peer review, the HRA information was moved to a separate report (PRA-A2-05-007, Rev. 0).

The HRA Notebook, Section 4.2 describes the correlation of instrumentation for HFEs to Appendix R instruments that have been confirmed to be available following a fire.

The existing HFEs were evaluated to determine if non-Appendix R instrumentation is required. The additional detail on the HRA instrumentation had a minor impact on HRA probabilities and HRA credit in the Fire Scenarios. The addition of detail for the HRAs represents an enhancement of the justification for HRA credit in the fire scenarios and does not impact the Fire PRA methodology.

PRA-A2-05-002, ANO2 Fire PRA New Human Failure Events, was revised to indicate the appropriate operator cues.

Event RHGISORCPBO was renamed to RHF2BLOFFP to reflect the ANO basic event naming convention, however, this event is not currently credited in the Fire PRA model.

EHF2A309XP has sufficient cues to ensure operator actions are taken; additional instrumentation is not necessary.

ES-C1-02 HRA Instrumentation Closed Appendix R instrumentation for Steam Generator pressure and level were credited for manual operation of AFW based upon adequate instrumentation demonstrated by the Appendix R analysis. The Appendix R analysis only validated that a minimum of one train was available. Since the PRA model may credit other trains, validation that adequate instrumentation for that train is available to support manual operations was not performed.

Evaluate instrumentation availability either within the PRM or through separate evaluation and document.

As stated in the F&O, one train of level instrumentation will be available to the operators for manual operation of the AFW.

Also, additional level and pressure indication is to be installed at a proposed AFW control and instrumentation panel outside the control room. The culmination of these instruments will ensure sufficient cues for all fire scenarios. In addition, the operator cues (continued)

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-10 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition ES-C1-02 (continued) for manual start of the AFW pump is based on both SG instrumentation and failure of the MFW and EFW pumps to operate. Since the SG instrumentation is not the only cue used for this operator action and no fire scenario damages all SG instrumentation, the HEP for manual start of AFW is not impacted by a failure of one train of Appendix R instrumentation.

ES-D1-01 Other Affected SR HRA-B3 HRA Instrumentation Closed ANO Fire Scenarios Report, 0247-06-0006.05, Appendix E has numerous documentation issues:

1) Some Basic Events (e.g. AHF2CSAS1P) do not have entries under "Operator Cues and Components Credited with Providing Operator Cues for Pose Fire Safe Shutdown" column, even though the "Required Instrumentation Available" column is marked Yes.

2) Column titles contain references to footnotes, but the footnotes are not in the document

3) The instrumentation identified as required for operator actions are listed in noun format, not instrumentation identifiers (e.g. "Steam generators pressure indications")

4) No explanation is given regarding whether or not all instrumentation listed in column for operator cues &

components

5) No link to documentation is provided to basis of conclusion in "Required Instrumentation Available" column Since there is information missing from Appendix E that is necessary for determining whether or not SR ES-C1 is met, this is a documentation finding. It also has an impact on the determination of SR HRA-B3.

Revise Appendix E of the report to include missing information and to clearly explain the basis of conclusion of availability of a component; this could be achieved by adding a column to reference the Safe Shutdown Capability Assessment or Safe Shutdown Equipment List - whichever document is more relevant.

At the time of the peer review, most of the Post Fire HRAs included information related to the instrumentation for operator cues in the Fire Scenarios report. Since the peer review, the HRA information was moved to a separate report (PRA-A2-05-007, Rev. 0).

1) The column "Operator Cues and Components Credited with Providing Operator Cues for Pose Fire Safe Shutdown" has been populated for all credited HFEs.

2) The footnotes associated with this Appendix have been moved to the Assumptions section of the HRA report.

These assumptions address the justification of Fire HRA multipliers and the definitions of accessible and simple actions to account for fire impacts on HRAs.

3) The instruments for each operator action used in the Fire PRA are provided using the component identifiers in the Cue Instruments / Components column.

4) The Cue Instruments / Components column has been added to the table in Appendix A to provide the instruments used.

(continued)

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-11 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition ES-D1-01 (continued)

5) The availability of at least one train of Appendix R instrumentation has been verified through the deterministic safe shutdown analysis. The availability of non-Appendix R instruments is documented based on discussions with Operations.

ES-D1-02 Other Affected SR ES-C2, HRA-A3 HRA Instrumentation Closed Attachment C of the Fire Scenario Report 0247-06-0006.05 is the Fire PRA Simulator Review, which assesses the instrumentation needs of the operators during / following a fire event. The review was conducted by a simulator operator and a Senior Reactor Operator and is formatted such that all of the appropriate questions are addressed to ensure that the SR is met. However, the basis for the conclusions is not included in the report. The condition for meeting the Category II requirement is answered by the questions, "Do the operators only rely on the specifically identified instruments?" and "Which tiles cause the operator to take and immediate action?"

The answers are that they do not rely on only one indicator and that there are no immediate actions taken due to annunciation. There is no other supporting documentation to demonstrate that these responses are correct.

The ANO team confirmed that the analysis had verified the responses, but there is no documentation of the basis of the conclusions; therefore, it is very difficult to recreate the same conclusions. Since critical documentation is missing, this is considered to be a finding as opposed to a suggestion.

Revise Attachment C of the report to include either a discussion regarding the basis for the answers or at a minimum to reference the applicable document. Also, clarify the relevancy of the unanswered questions or provide a response.

The HRA Notebook, PRAA205007 was revised to include the information discussed in this F&O. Appendix A includes all of the instruments that would provide indications for the operator actions. In addition, OP-2203.049 verifies the Appendix R instruments that are available for the worst case fire in each fire area. The operators are trained to rely on these instruments for cues to perform the operator actions.

COPD-001, Operations Expectations and Standards Section 5.4.2 provides the following direction; Numerous events within the industry have occurred because the Operator performing the reactivity change only focused on one parameter to determine core response, such as, only monitoring RCS average temperature following a reactivity addition and ignoring other critical parameters. This practice must be avoided.

The individual as well as the Operations team must use diverse and multiple indications to monitor changes to the core as a result of reactivity manipulations.

Therefore, a basis is available to support the conclusions of the simulator review.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-12 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition FSS-A4-01 Targets in ZOI - missing a specific conduit Closed Plant drawing E-2872 was reviewed to identify some sample targets (EC3021, EC3022, B4124, B4125, and J4519) within the zone of influence of 2D-35 (Scenario 2109-U-B). All the targets were listed in Attachment A, except B4124. A review and plant walk-down should be performed to determine if B4124 should be added to the list of targets in Attachment A of ERIN Report 0247-06-0006.05.

Because this is a conduit, the location of conduits may not necessarily match the location in the drawing. If this conduit is located in the zone of influence of 2D-35, it will need to be listed as a target.

Verify PDMS information is consistent with walk-down results and drawings.

Conduit B4124 was determined to be within the zone of influence of 2D-35. Therefore, conduit B4124 was added to the list of targets for scenario 2109-U-B in Attachment A of the Scenario Report. This finding is associated with a specific scenario with a missing target.

A review of drawings identified targets that were difficult to locate physically in the field.

This review was performed for all scenarios in congested fire zones. The walkdown target list includes notes that identify some of the targets that were identified from the drawing review versus the original.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301) (ML13312A877).

FSS-A5-01 Ignition Frequency Closed The ignition frequency used in scenario 2111-T-B is for the electrical panels and not a battery charger. The CDF quantification for this scenario will need to be revised to use the correct ignition frequency.

A related F&O FSS-D10-01 identified that a battery charger was missed from the ignition source walk-down sheets and the CDF quantification results were not correct.

The CDF quantification for 2111-T-B has been revised to use the appropriate ignition frequency. See -PRA-A2-05-003 - the ANO-2 Fire Scenarios Report for details.

The ignition source data sheet and walkdown sheet for compartment 2111-T was revised in CALC-08-E-0016-01 to include battery charger 2D-33.

FSS-B1-02 Control room abandonment Closed The glass partition between the two control room fire compartments (129-F and 2199-G) is not a fire rated barrier.

The calculation of the control room abandonment CDF for Unit 2 only considers the fire initiating frequency for Unit 2, and does not consider a fire in the Unit 1 control room. The fire ignition frequency for the Unit 2 control room abandonment is under estimated.

Include the fire ignition frequency for a Unit 1 control room fire in the Unit 2 control room abandonment CDF.

The Evaluation of Unit 2 Control Room Abandonment Times at the Arkansas Nuclear One Facility [ANO2-FP-09-00013] defines the conditions that are assumed to cause control room abandonment and reliance upon alternate shutdown actions. Abandonment times are assessed for ANO-2 for fire scenarios originating in the ANO-1 control room and the sensitivity study also assumes failure of the glass barrier separating the ANO-1 and ANO-2 control rooms.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-13 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition FSS-B2-01 Shutdown following MCR abandonment Closed The control room abandonment (CRA) scenario for Unit 2 (2199-G-B) assumes a CCDP of 0.1 for an alternate shutdown scenario. The basis for this CCDP in the Fire Scenarios Report states that this is "a conservative CCDP, probability of failure of shutdown from outside the control room, given that a deterministic analysis has been performed and feasibility of required actions have been assessed to ensure feasibility of this shutdown scenario for a worst case fire". This is not a sufficient basis to validate a CCDP value of 0.1.

A determination of a bounding risk for the MCR abandonment scenario is not demonstrated with the assumed CCDP.

A detailed Control Room Abandonment analysis [CALC-09-E-0008-10] was performed to calculate the CCDP for MCR Abandonment.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301) (ML13312A877).

FSS-C5-01 Thermoplastic Cables Closed Per ER-ANO-2003-0450-000, Revision 0, approximately 1% of the cables in the plant are thermoplastic. These cables are primarily COAX used to support instrumentation. The subject ER is limited in scope and does not identify Fire PRA cables they may be thermoplastic and require a lower damage temperature.

Significance of effects of different types of cables could be enough to impact results.

Determine which (if any) fire PRA credited cables are thermoplastic and use appropriate damage threshold in fire damage scenarios.

CALC-ANOC-FP-09-00019, Rev. 0, EC-6964, Safe Shutdown Cable Jacket Insulation Types at ANO, evaluates the cables at both units and concludes that of the over 4600 cables reviewed, less that 0.3%

have thermoplastic insulation. This calculation also confirms that thermoplastic cables are not used in power supply circuits.

Thus, thermoplastic insulation at ANO is of minimal significance. CALC-ANOC-FP 0019 is referenced in the FPRA Scenarios Report.

FSS-C8-01 Fire wraps -

Fire-Resistance Rating Closed ANO has not confirmed that the wrap will not be subjected to direct flame impingement or validation that the wrap is qualified for fire impingement was not performed.

Lack of basis for wrap validation is significant information; therefore, F&O is finding rather than suggestion.

Review credited raceway wraps and either validate that wrap is qualified for fire impingement or validate that wrap is not subjected to fire impingement.

The FPRA analysis was revised such that it does not take credit for the 1-hour (Hemyc) fire wrap on service water pump cables in areas OO and B-6 or on charging pump cables in area DD. As shown in the Scenarios Report, in base scenarios for these areas, the wrapped cables are assumed to fail. For specific fire scenarios in these areas, the wrapped cables are assumed to fail if they are within the zone of influence of the fixed or transient ignition source. Thus, the FPRA results for these areas were calculated assuming that the 1-hour fire wrap does nothing to mitigate the extent of fire damage.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-14 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition FSS-D3-01 Fire growth and propagation within cable trays Closed Implementation of the generic fire modeling methods did not consider the effect of fire growth and propagation within cable trays. A review of fire modeling applications within fire compartment 2109-U shows that the HGL determination was based upon bounding estimates for cabinet heat release rates and did not validate that the value would be bounding with the inclusion of fire spread to the cable trays.

This is a finding because the analysis may not be bounding if the heat release rate contribution of the cable trays would be involved in a cabinet fire scenario.

Consider the effect of fire growth and propagation within cable trays to validate that the value would be bounding with the inclusion of fire spread to the cable trays.

The generic fire modeling methods were updated to perform a case to account for hot gas layer effects on cable trays located one foot above the electrical panel. See Section 2.2 of the Multi-Compartment Analysis (PRA-ES-05-004). Therefore, the fire growth and propagation within the cable trays are addressed in the fire scenarios associated with bounding cabinet heat release rates such as with fire compartment 2109-U.

FSS-D7-01 Site-specific suppression system unavailability Closed Site-specific suppression system unavailability values were not evaluated. Surveillance requirements are incorporated into site procedures, but specific unavailability values in the cable spreading room are not tracked for Fire PRA purposes. The SR is not met per NUREG/CR-6850 guidance because the methodology does not include maintenance contributions to unavailability, credit for manual actuation of the system, dependent failures, and plant specific data.

F&O is a finding since the analysis did not include attributes necessary to meet the NUREG/CR-6850 guidance for uncertainty evaluations.

Incorporate the required plant specific data regarding unavailability per NUREG/CR-6850 guidance.

The ANO-2 Fire Scenarios Report (PRA-A2-05-003) discusses the review of the fire suppression systems credited in the Fire PRA. Explicit credit for suppression and detection systems is taken for the cable spreading room fire scenario only. The smoke detection system is credited for early detection of a fire in the cable spreading room. A review of impairments associated with the cable spreading room detection system indicated that only individual detectors were out of service for a limited period of time during the period from 2007 through 2009. Therefore, the unavailability of these systems is very low and is considered to be enveloped by the system unreliability data taken from NUREG/CR-6850.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-15 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition FSS-D10-01 Other Affected SR FSS-H10 Verification of ignition sources via walkdowns Closed Compartment 2111-T was one of those areas that was not walked down by ERIN, but was walked down by the FPRA Peer Review. A battery charger 2D33 is located in compartment 2111-T and is not listed in the Walk-down sheet or in the Ignition source data. Further review of the fire scenarios listed in ERIN Report 0247-06-0006.05 showed a fire scenario that takes into account battery charger 2D33 as the ignition source (FRANC scenario 2111-T-B in Attachment A). It is recommended that the fire areas that were not walked down (listed in Table 2-1 of CALC-08-E-0016-01) have the ignition sources re-verified by a walkdown or compared against the list of fire scenarios.

Because a fire scenario for the battery charger 2D33 was evaluated using the wrong ignition source frequency (see F&O FSS-A5-01), the battery charger needs to be added to the plant walk-down and the data sheets to ensure the correct ignition frequency is used in the CDF quantification.

Because the missing data was due to not performing the plant walk-down, it is recommended that the list of compartments in Table 2-1 of CALC-08-E-0016-01 be re-considered for walk-downs to ensure that all the applicable ignition sources are adequately evaluated as fire scenarios.

The ignition source data sheet and walkdown sheet for compartment 2111-T were revised in CALC-08-E-0016-01 to include the battery charger. The walkdown sheets in CALC E-0016-01 were developed during initial ignition frequency walkdowns. Subsequently, additional walkdowns were performed for fire scenario development. During the later walkdowns, access was obtained to some zones previously inaccessible. The notes from the subsequent walkdowns were reviewed to determine if components in addition to those in CALC-08-E-0016-01 were noted. A review of the walkdown sheets resulted in some additional components being added based upon the walkdowns. In addition, other changes were identified and incorporated addressing new information relating to the components identified.

FSS-E2-01 Other Affected SR UNC-A2 Justification for assumption of no target damage from ventilated cabinet fire within 30 minutes Closed Section 7.1.2.1 of the ANO Fire Scenarios Report 0247 0006.05 evaluates the severity factor for ventilated cabinets located in fire zones equipped with automatic detection. The evaluation assumes that no target damage will result if the fire is suppressed within 30 minutes. This is an application of expert judgment in lieu of plant-specific data or generic estimates. The justification of the basis for this application has not been adequately developed or supported.

Justification has not been provided for the basis of the specific expert judgment discussed above and therefore may not be considered to be valid.

Perform a more extensive study to justify the use of the 30-minute non-damage time based on fire brigade response.

The non-suppression probabilities for electrical cabinet fires were changed based upon a methodology that has been submitted to the EPRI Fire PRA Methods Panel. These values are based on Panel voltage ratings and do not include any assumptions for suppression times.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301) (ML13312A877).

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-16 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition FSS-G2-01 Multi-Compartment Analysis Fire barrier failure probability Closed Multi-Compartment Screening Analysis methodology includes a manual suppression probability based upon the rating of the fire barriers and that they would withstand a fire for a minimum of 90 minutes. A barrier failure probability is used based upon the bounding assumption that each barrier includes a normally closed door (7.4E-3). In the screening method, the value of manual suppression is multiplied by a barrier random failure probability. Considering that the manual suppression probability is based upon 90 minutes, these two values should not be ANDed.

Given that the method is intended to be bounding for purposes of screening, this method may artificially reduce the CDF by up to 3 orders of magnitude.

This F&O was written on the analysis when credit for manual non-suppression probability was based on the rating of the barrier. This approach was corrected in the latest revision of the MCA/HGL analysis. The current analysis credits a manual non-suppression probability based on the time required to generate a hot gas layer. The original approach would be susceptible to a dependency between the non-suppression probability and the barrier failure probability.

In the current approach the non-suppression probability is not related to the barrier rating and therefore has no significant dependency with respect to the barrier failure probability.

FSS-H7-01 Other Affected SR FSS-E2 Fire brigade response time Closed Section 7.1.2.1 - Severity Factor for Ventilated / Open Cabinets from the Fire Scenarios Report describes values credited in fire brigade suppression prior to damage external to the effected cabinet. The assumptions related to this method have not been justified.

The method assumes that fire detection will occur 30 minutes prior to external damage to cabinets. This assumption has not been justified. NUREG/CR-6850 indicates a bounding assumption that in-cabinet detectors may provide an additional 5 minutes of time for detection. Therefore, without justification 30 minutes should not be used. In addition, the 30 minutes does not account for plant specific response times to the alarm. This value would also need to be considered. The lower branch does provide a 15 minute delay to fire brigade response. The 15 minutes included in NUREG/CR-6850 is loosely based upon control room indication of fire due to failed equipment that results from a developed fire. Justification for credit of this value would need to be provided on how operations would know to send fire brigade based upon a fire limited to a single cabinet with no fire alarm.

The concern identified in this F&O no longer applies. The non-suppression probabilities for electrical cabinet fires were changed based upon a methodology that has been submitted to the EPRI Fire PRA Methods Panel. These values are based on Panel voltage ratings and do not include any assumptions for suppression times.

Brigade response is only credited for transient fires in the cable spreading room.

This is described in Section 9.0 of the Fire Scenarios Report (PRA-A2-05-003).

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-17 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition FSS-H7-02 Fire brigade capability to alter hot gas layer development Closed Section 9.1 of the Fire Scenarios Report contains the method used to evaluate overall compartment damage through hot gas layer development. This section of the report includes credit for fire brigade capability to alter hot gas layer development in 20 minutes which in essence results in a 1.0 probability that if a HGL (625 deg F) does not develop within 20 minutes it will not develop. This is an input/ assumption that is not clearly identified in the report. In addition, the basis for this value is not provided. The model employed does not discuss impacts that would support this time including fire brigade response time, fire characterization, and probabilities for failure.

Brigade response is only credited for transient fires in the cable spreading room.

This is described in Section 9.0 of the Fire Scenarios Report (PRA-A2-05-003). This evaluation assumes that manual suppression would have to occur within 5 minutes and before damage to any cable trays. In this case, a hot gas layer would not have time to develop.

HRA-A2-01 Other Affected SR HRA-B3 New fire specific HRA Procedures Open Identification of new operator actions for each fire scenario did not include fire area specific EOPs (Emergency Operating Procedures). This can result in the failure to identify operator actions which may de-energize equipment, lead to actions different from those assumed in the Fire PRA, and/or more direct symptom-based procedural actions. For example, for a fire in Fire Area B-3, the operator is specifically directed (by FIRES IN AREAS AFFECTING SAFE SHUTDOWN, 2203.049, Rev. 7) to verify the RCP breakers are OPEN with DC Control Power removed. This is different from the HRA write-up which assumes that 10 minutes will be used to attempt to trip the RCPs in the CR and that "when they [the operators] tried to trip the RCPs from the control room, they would not be successful and would decide to open the associated breakers." This would be evaluated differently if procedurally guided.

It was noted by the Fire PRA staff that the fire-related procedures were planned to be changed to make them more symptom-based and remove many of the steps currently in the procedures.

The SR specifically requires, for each fire scenario, identification of any new fire-specific safe shutdown actions called out in the plant fire response procedures. The Fire PRA staff noted that the fire-related procedures were expected to be changed and did not want to add actions to the PRA that would have to be removed before applying the fire PRA.

(continued)

The new HRA events identified for the Fire PRA are documented in calculation PRA-A2-05-002. This calculation evaluates the probabilities for these new operator actions and any new combinations between these actions and the other HRA events. Each of the HFEs that were assessed for the transition to NFPA-805 and contain a reference to the guiding procedures within the HRA worksheets. Revisions to OP-2203.049 will be required prior to transition to NFPA-805 that will provide additional guidance to the operators for fire in each area of the plant.

The transition to NFPA-805 will result in removal of many operator actions currently considered necessary under Appendix R.

Therefore, many actions deemed necessary in the current version of OP-2203.049 are not required post transition to 805.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301) (ML13312A877).

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-18 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition HRA-A2-01 (continued)

SR HRA-A2 was assessed as Not Met for the reasons discussed above. To Meet this SR, ANO-2 should update the fire-related procedures as the fire PRA staff indicated to be planned. Following this update to the fire-related procedures, perform and document a systematic review of the plant fire response procedures and identify fire-specific safe shutdown actions that may be taken by the operator for inclusion in the PRA. It is recommended that all HFEs included in the model be tied directly to either the fire-related procedures or the normal EOPs.

HRA-A4-01 HRA documentation of operations or training personnel to confirm HRA modeling is consistent with plant operation Closed Attachment C of the Scenarios Report (0247-06-0006.05 Revision 0) provides the results of talk-and walk-throughs with operations personnel regarding instrumentation and indication issues during fire scenarios. However, there is no evidence of reviews or talk-throughs with operations or training personnel to confirm that the interpretation is consistent with plant operational and training practices during fire scenarios.

SR HRA-A4 requires at least a review of the interpretations of procedures associated with actions in the PRA with operations or training personnel to meet Capability Category I.

SR HRA-A4 was assessed as Capability Category I for the reasons discussed above. To achieve Capability Category II/III, ANO-2 needs to perform and document the review or talk-through with operations or training personnel to confirm the interpretation of procedures is consistent with both operational and training practices. This should be performed after the fire-related procedures are revised. (See F&O HRA-A2-01)

Although this SR only meets CC-I, the issue identified does not impact the FPRA results and is ultimately considered a documentation issue associated with ensuring that the information provided by the Shift Managers involved with the interviews are familiar with the Ops and training requirements and expectations. ANO has a high degree of confidence that the Shift Managers are familiar with these expectations since it is one of the responsibilities of the Shift Managers.

In addition to the general talk-throughs of fire scenarios and simulator observations in Attachment C of the Fire HRA calculation (PRA-A2-05-007), ANO-2 has documented operator interviews of operator actions identified for the Fire PRA in calculation PRA-A2-05-002 and has identified operator cues and instruments required for each operator action.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-19 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition HRA-D2-01 Other Affected SR PRM-B9 HRA /

availability of action during fire Closed The ECCS vent path is modeled as a spurious open vent path (2CV-4740-2 and 2CV-4698-1) equivalent to a medium LOCA.

The model logic (Gate FIRE003) includes an operator recovery action (RHFPZRVNTISO) from the control room to isolate the vent path by closing valve 2CV-4740-2. However, 2CV-4740-2 is one of the spuriously opened valves, and may not be remotely operable.

The ECCS vent path may not be remotely isolable given a spurious opening due to a fire. The recovery action assigned in the model may not be effective.

A detailed circuit analysis of 2CV-4740-2 should be performed.

Operator recovery action RHFPZRVNTISO, to close valve 2CV-4740-2, has been deleted from the FPRA model. Additionally, all valves requiring manual operation post fire have been evaluated for 92-18 concerns. If a valve has been identified to require manual manipulation after spurious operation and is indeed a 92-18 concern, the valve has been identified for modification within the FREs.

IGN-A5-01 Ignition frequencies correlated to availability factor Closed Table 3-2: some IEFs are not corrected to a "per reactor-year" basis. Some ignition frequencies are based on all modes of operation (e.g., batteries) and are correctly updated with calendar years as described in Section 3.1 of Entergy Calculation CALC-08-E-0016-01 Revision 0. However for those ignition frequencies, plant availability must be factored into the initiating event frequency or it will be conservative by approximately 10%.

IGN-A5 requires generic fire ignition frequencies or plant-specific fire frequency updates on a reactor-year basis. This is not done for the following ignition frequency bins: 1, 4, 8-10, 12-19, 23, 26, and 30.

Multiply the frequencies of all-mode Ignition Frequencies by the availability factor (critical years/calendar years).

IGN-A5 is Met because the plant-specific fire frequency updates were revised to reflect a reactor-year basis. The plant availability was used in determining the frequencies by the fraction of time the plant was at-power.

CALC-08-E-0016-01 Table 3-2 has been changed to show that all bins were updated on a reactor-year basis.

IGN-B4-01 Other Affected SR IGN-A6 Ignition Frequency/

Bin/frequency error Closed It appears that the 5th and 95th frequencies in Table 3-2 of Entergy Calculation CALC-08-E-0016-01 Revision 0, does not match up with the 5th and 95th frequencies from Table C-3 of NUREG/CR-6850. Additionally, Bin 16 has been subdivided to include isophase bus ducts bins (16c and 16d) based on FAQ 0035, but this is not reflected in Table 3-2.

(continued)

Table 3-2 of CALC-08-E-0016-01 was revised to reflect the data from NUREG/CR-6850 Table C-3, as well as data from FAQ 07-0035 for bins 16c and 16d and data from FAQ 06-0017 for bins 16a and 16b.

Per e-mail from the peer reviewer for this SR (L. Shanley) to J. Renner, dated 9/2/09, 1:36 pm; the portion of this F&O related to the uncertainty information was in error. No action is required.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-20 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition IGN-B4-01 (continued)

Table 3-2 of Entergy Calculation CALC-08-E-0016-01 Revision 0 does not accurately reflect the data used to develop ignition frequencies and compartment initiating event frequencies. Additionally, a spot check of the uncertainty information associated with the generic frequencies provided by the PRA team (FIF Bart Input Template (Updated 11-14-2008).xls) found one error (Bin 1 Range Factor is listed as 11.0 and it should be 10.0); other data appears to be correct.

Correct Table 3-2 to reflect the correct information used for the Bayesian Update. Review all data used for the Bayesian update and correct any additional errors (beyond Bin 1), if found. It is also recommended that the document specify the prior distribution parameters used in the BART calculations (i.e., mean and range factor) to allow for reproduction of the update results. Provide details on isophase bus duct frequencies (bins 16c and 16d) from FAQ 0035.

PRM-B2-01 Internal Events PRA Open Items Closed The majority of the Internal Event deficiencies are open. The items have not been dispositioned such that their impact on the Fire PRA could be determined. A review of the items indicate one item in particular (timing for securing RCPs following a loss of CCW) that could have impact on the Fire PRA results. ANO-2 needs to review the disposition of open items from their level 1 internal initiator PRA to ensure that their disposition remains valid in view of the unique aspects of fires.

The ANO-2 Internal Events Peer Review (LTR-RAM-II-08-020 identified 26 findings against the Internal Events supporting requirements. These findings are discussed in Attachment U of this LAR submittal. A qualitative review indicated that most of these findings would have no or minor impact on fire risk.

ANO-2 has created a new Fire HFE associated with tripping the RCP outside the control room. The system time window for this HFE uses a shorter available time than that assumed in the internal events analysis.

This new available time is based upon WCAP16175-P-A Model for Failure of RCP Seals Given Loss of Seal Cooling in CE NSSS Plants.

(continued)

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-21 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition PRM-B2-01 (continued)

The system time window for the HFE associated with tripping the RCP from within the control room (developed as part of the internal events PRA) will be revised in the next revision to the internal events PRA. The difference in failure probability using the shorter available time for this HFE does not change significantly due to the contribution of the Caused Based Method to the resulting failure probability.

PRM-B9-01 Other Affected SR HRA-D2 Plant

Response

Model Added Components for HRA actions Closed Components credited to support fire recovery actions have not been reviewed per the requirements of the applicable internal events standard. For example, valves added to support recovery actions that isolate spurious operation flow diversions do not always included credited power supply dependencies and random failure modes for the equipment.

Review added components credited in HRA recovery and add appropriate basic events and links within the PRM. Document as required.

All new fire recovery actions in the ANO-2 FPRA model have been reviewed per the requirements of the applicable internal events standard and appropriate basic events and links were added to the plant response model to ensure that dependencies are modeled.

Therefore, plant response model (PRM)-B9 is Met.

PRM-C1-01 Plant

Response

Model Documentation of PRM development Closed The Supporting Requirement requires the Fire plant response model be documented consistent with HLR-IE-D, HLR-AS-C, HLR-SC-C, HLR-SY-C, and HLR-DA-E and their SRs in Section 2. No specific documentation could be found that explains the development of the Plant Response Model consistent with the SR requirements above.

The supporting requirement to document the PRM (PRM-C1) is not met. The work supporting the PRM (including modifications to the internal events PRA) needs to be tied together in a summary document.

Prepare a calculation report that documents the PRM in accordance with SR PRM-C1.

This is a documentation issue only. The following discussion was added to Section 1 of the Component and Cable Selection reports linking the requirements of PRM-C1 to the internal events model.

The FPRA model uses the ANO Internal Events model as its basis. The methodology used to update the ANO Internal Events model to generate the FPRA model uses the same methodology used in the development and documentation of the Internal Events model. The FPRA documentation may differ from that used for the internal events model due to FPRA specific requirements, however, the basic intent of the internal events documentation requirements are met.

(continued)

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-22 Table V-1 Fire PRA Peer Review - Findings and Observations SR Topic Status1 Finding/Observation Disposition PRM-C1-01 (continued)

The FPRA therefore meets the same model development and documentation requirements applicable to the internal events model. Therefore, requirements of the ANSI/ASME standard applicable to the PRA model specified in PRM-C1 are met via the internal event model meeting the associated model requirements.

Additional details on development of the PRM from the Internal Events PRA model can be found in Section 4 of the ANO-2 Component and Cable Selection Report.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301) (ML13312A877).

UNC-A1-01 Other Affected SR FQ-F1, UNC-A2 Uncertainty of CDF and LERF results Closed The impact of parameter uncertainties were not estimated or propagated as required in SR QU-E1 and for LERF (LE-F3).

Additionally, the uncertainties are not characterized as required by SR QU-E4. Appendix D lists and discusses sources of uncertainty; however the characterization of these uncertainties is not detailed.

Uncertainty must be evaluated to include "ESTIMATE the uncertainty interval of the CDF results" per QU-E1, Capability Category I or II (different details, but both require uncertainty interval to be estimated). This is also required for LERF (LE-F3).

Propagate or estimate the impact of parameter uncertainties on CDF and LERF. Alternatively, a defined basis can be developed to support the non-applicability of this SR.

Appendix D of the Fire PRA Summary Report (PRA-A2-05-004) addresses sensitivity to the sources of uncertainty for the Fire PRA tasks.

Additionally, PRA-A2-05-006, ANO-2 Fire PRA Uncertainty/Sensitivity Analysis has been developed to calculate the uncertainty associated with the CDF and LERF values of the Fire PRA. The information provided in these two documents satisfies the requirements identified in this F&O.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301) (ML13312A877).

MU-B4-01 Software upgrades Closed There is no reference to a peer review for upgrades. Did not find a section which addressed upgrades (not updates) to the PRA specifically involving changes to key PRA software.

EN-DC-151 (PSA Maintenance and Update) procedure was revised to require an industry peer review if a PSA upgrade per ASME/ANS RA-Sa-2009 has occurred.

1 The Status of Closed indicates that the Peer Review Finding has been addressed in the NFPA-805 evaluation and incorporated into the Fire PRA model and documentation.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-23 Table V-2 Focused Scope Fire PRA Peer Review - Findings and Observations SR Topic Status Finding/Observation Disposition FSS-D8-01 Fire Detection and Suppression System Open This SR requires an assessment of fire detection and suppression system effectiveness in the context of analyzed fire scenarios. The ANO-2 fire PRA credits the smoke-actuated dry pipe sprinkler system in the cable spreading room for preventing target damage from a transient fire. That is, the fire PRA only models target damage if the system fails. ANO-2 did not perform (or at least did not document) a technical basis that the suppression system is capable of preventing target damage. With the current model assuming a 317 kW fire, a relatively short fire growth phase, and the relative proximity of overhead cable trays to the postulated fire, the suppression system may not be capable of suppression prior to target damage.

Suggested Resolution - Provide a technical basis that the cable spreading room suppression system can effectively extinguish a transient fire prior to target damage.

As an alternate resolution, consider making this area a zero-transient area per EN-DC-161 such that a 69 kW transient (versus the current 317 kW) can be modeled per the ANO-2 transient methodology, in addition to being able to credit prompt suppression if the procedure revision requires a continuous fire watch when transient combustibles are present. The technical basis that the suppression system can extinguish the 69 kW fire prior to target damage should also be provided per this Supporting Requirement.

The lower heat release rate has been justified as applicable in the cable spreading rooms and will be supported by a revision to EN-DC-161, which will restrict the combustibles allowed within the affected rooms. A Work tracking item (LO-WTANO-2010-00222 CA-00011) has been issued to ensure that this zone is included as a no transient zone. This action requires an update to the procedure for control of combustibles [EN-DC-161] to ensure "zero-transient" zones in the ANO-1 and ANO-2 Fire Scenarios Reports are designated and maintained. As part of the transition to an NFPA-805 licensing basis, the ANO-1 and ANO-2 Fire Scenarios Reports specify that certain fire zones must be maintained as "zero-transient" zones.

Credit for manual suppression of a 69 kW transient fire has been credited for this fire zone. No specific credit for the automatic suppression systems has been taken.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301) (ML13312A877).

FSS-H1-01 Transient Fire size documentation Open The ANO-2 fire PRA documentation is generally adequate and this SR is met with an F&O.

Procedure EN-DC-161 Control of Combustibles Rev 5 was reviewed against Table 8-1 of the Fire Scenarios Report. This review identified that four fire zones (2111-T, 2096-M, 2112-BB, and 2182-J) allowed transient combustibles (in some cases up to 300 pounds) with a roving fire watch. However, the fire PRA model postulates 69 kW transient fires in these areas consistent with its methodology for zero-transient areas.

(continued)

A corrective action has been written to revise EN-DC-161 to address this discrepancy. Work tracking item #LO-WTANO-2010-00222 CA-00011 has been issued to ensure that this zone is included as a no transient zone. This action requires an update to the procedure for control of combustibles [EN-DC-161] to ensure "zero-transient" zones in the ANO-1 and ANO-2 Fire Scenarios Reports are designated and maintained. As part of the transition to an NFPA-805 licensing basis, the ANO-1 and ANO-2 Fire Scenarios Reports specify that certain fire zones must be maintained as "zero-transient" zones. (continued)

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-24 Table V-2 Focused Scope Fire PRA Peer Review - Findings and Observations SR Topic Status Finding/Observation Disposition FSS-H1-01 (continued)

After some discussion, it was identified that ANO-2 plans to revise EN-DC-161 to require continuous fire watch for any transient storage in these areas. The Fire Scenarios Report should be revised to reference an ANO-2 action item or CR tracking implementation of this procedure revision.

On a related note, ANO-2 may consider crediting prompt suppression for these areas.

Suggested Resolution - Revise the Fire Scenarios Report to reference an ANO-2 action item or CR tracking implementation of revision to ENDC-161.

While the action item has been written and incorporated into the Paperless Condition Reporting System (PCRS), it was not deemed prudent to incorporate a temporary work tracking item into the Fire Scenario Report.

IGN-A7-01 Fire Ignition Frequency Apportioning Closed This SR relates to the fire ignition frequency apportioning methodology. ANO-2 meets this SR for transient ignition sources based on preponderance of evidence; however, one deficiency was noted with the turbine building transient fire frequency. In the turbine building, ANO-2 postulated 12 transient fire scenarios that could affect PRA targets. The area factor for each source was calculated as 100 sq. ft.

divided by the total turbine building floor area. The total area factor for all turbine building transients was therefore 1,200 sq.

ft. divided by the total turbine building floor area. Based on inspection of plant drawings, the total floor area containing PRA targets of concern is greater than 1,200 sq. ft., and the ANO-2 model is therefore not accounting for some fraction of the turbine building transient fire frequency. The underlying cause of this error was using a nominal 100 sq. ft. for each transient and not postulating enough scenarios to cover the risk-relevant floor area. Alternatively, the area factor for each transient could be inflated to ensure the entire risk-relevant floor area is considered.

Suggested Resolution - Adjust the turbine building transient area factors such that the total floor area containing risk-relevant targets is considered.

The floor areas assumed for the transient fires were adjusted based upon walkdowns and drawings to better reflect the area of the scenarios being developed. This approach increased the size of the floor area used in calculating the ignition frequency for each of the scenarios considered. This new information is documented in the Scenario Report Attachment H and Attachment D.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301) (ML13312A877).

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-25 Table V-2 Focused Scope Fire PRA Peer Review - Findings and Observations SR Topic Status Finding/Observation Disposition HR-G3-01 Evaluation of Instrumentation Cues Closed All assumptions that impact feasibility of operator actions need to be validated before the HEPs can be applied, specifically assumptions regarding availability of instrumentation credited for diagnosis should be verified available in the fire scenarios where the HFE is credited. This includes instrumentation credited as diverse to instrumentation that is rendered unavailable due to fire. Note that these assumptions may be implicit in analyses where cues and indications are stated without any statements regarding availability in the fire scenarios where the HFE is credited.

Related to the above are assumptions made with regards to explicit modeling of instrumentation. Assumptions regarding procedure changes also need to be validated.

Credit for instrumentation is documented in the post fire safe shutdown analysis. This instrumentation is correlated to the credited HEPs in the Detailed Fire HRA for Selected ANO-2 HFE, Appendices B and C.

HR-G7-01 Availability of Resources to Support HEPs Closed Based on a review of Figure 5 in PRA-A2-01-003S03, the dependency approach does not consider availability of resources, which can be important for fire PRA.

A review was performed of the most challenging scenario with respect to manpower requirements, control room abandonment, which confirmed that available manpower is sufficient to support multiple HEPs in a cutset for the control room abandonment cutsets. The review is documented in ANO-2 PRA-A2-05-007, Rev. 1, Fire Probabilistic Risk Assessment Human Reliability Analysis (HRA) Notebook, Section 4.8.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-26 Table V-3 Fire PRA-Category I Summary SR Topic Status PP-B2 Credit for non-rated fire barriers The ANO Plant Partitioning and Fire Ignition Frequency Development report uses the plant areas that are identified in the plant FPP. The ANO-2 Peer Review team noted two instances where this approach allows credit for non-rated fire barriers: 1) the glass partition between the ANO-1 and ANO-2 Control Rooms and 2) the Turbine deck, which does not have a rated barrier between the units. The MCR Abandonment analysis provides an evaluation of fires that would fail the partition and require both Control Rooms to be abandoned. The MCA evaluated fires on the turbine deck to determine if a fire in this area would spread to the other unit. The MCA did not identify any fires on the turbine deck that would spread to the other unit. Detailed evaluation of the impact to fires on the barriers was completed. The detailed evaluation of the non-rated barriers ensures the Category I limitation does not impact the results or conclusions of the model.

PP-B3 Credit for spatial separation as a partitioning feature PP-B5 Credit for active fire barriers The ANO Plant Partitioning was performed without taking credit for active fire barriers. Active barriers are used to separate divisional rooms. Not crediting the active barriers for partitioning only shifts the burden for fire risk evaluation to the Fire Scenario development process. Therefore, this assumption has a minor impact on overall fire risk. Active fire barriers were appropriately credited/modeled in scenario development. While this approach only meets the Category I requirement for portioning (PP), it does not significantly impact the overall model results or conclusions.

CS-B1 Analyze Electrical Buses for Overcurrent Coordination Section 4.4 of the Component and Cable Selection Report (PRA-A2-05-005) documents the Electrical Coordination/Protection for ANO-2. This document refers to Upper Level Document ULD-0-TOP-12 for FPRA components in the SSEL. Components that are not in the SSEL are evaluated in Table 4-3 of the calculation. The difference between Category I and Category II is the use of an existing document (Cat I) instead of completing a new analysis for all modeled buses (Cat II). The Category I classification only partially applies. An existing analysis was used for FPRA components in the SSEL. Evaluation of non-SSEL buses meets the Category II requirement. Though the current method only partially meets the Category II requirement, it was judged to be acceptable for the NFPA 805 application.

Also see Entergy response to NRC PRA RAI 01, dated November 7, 2013 (2CAN111301)

(ML13312A877).

IGN-A10 Provide uncertainty intervals for fire ignition frequencies Section 6.2 of the ANO-2 FPRA Uncertainty/Sensitivity Analysis (PRA-A2-05-006) discusses the uncertainty intervals used for the fire ignition frequencies for propagating uncertainty. The uncertainty from ignition frequency development is primarily from the NUREG guidance. ANO strictly followed NUREG/CR-6850 guidance in developing ignition frequencies using numbers included in the document.

Arkansas Nuclear One - Unit 2 Att. V - Fire PRA Quality to 2CAN121202 Page V-27 References

1.

LTR-RAM-II-09-046, Fire PRA Peer Review against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the Arkansas Nuclear One, Unit 2 Fire Probabilistic Risk Assessment, September 2009.

2.

LTR-RAM-I-11-064, Focused Scope Fire PRA Peer Review for Arkansas Nuclear One Unit 2, December 2011.

3.

Echelon Calculation PRA-A2-05-003, Fire Scenarios Report NUREG/CR-6850 Tasks 8 and 11, Rev. 1.

4.

Entergy Engineering Change EC13540, ANO2 Cable Routing Exclusions to Support Fire PRA for NFPA-805.

5.

Echelon Calculation PRA-A2-05-005, Rev. 1, ANO-2 Fire Probabilistic Risk Assessment Component and Cable Selection Report,

6.

ANO Calculation CALC-ANOC-FP-09-00019, Rev. 0, Safe Shutdown Cable Jacket Insulation Types at ANO, July 2009.

7.

ANO Calculation CALC-85-E-0087-24, Rev. 1, Safe Shutdown Cable Analysis, April 2006.

8.

Echelon Calculation PRA-A2-05-020, Detailed Fire Human Reliability Analysis (HRA) for Selected ANO-2 Human Failure Events (HFE).

9.

Echelon Calculation PRA-A2-05-007, Rev. 1, ANO-2 Fire Probabilistic Risk Assessment Human Reliability Analysis (HRA) Notebook.

10. ANO Calculation PRA-A2-05-013, Rev. 0, ANO2 Fire Probabilistic Risk Assessment Plant Partitioning and Fire Ignition Frequency.

11. ANO Calculation ANO2-FP-09-00013, Rev. 1, Evaluation of Unit 2 Control Room Abandonment Times at ANO Facility.

12. ANO Calculation CALC-09-E-0008-010, Rev. 1, ANO-2 Fire Area G Risk Evaluations.

13. Echelon Calculation PRA-A2-05-014, Rev. 0, Multi-Compartment / Hot Gas Layer Analysis.

14. Echelon Calculation PRA-A2-05-004, Rev. 2, Fire Probabilistic Risk Assessment Summary Report, NUREG/CR-6850 Task 16.

15. Echelon Calculation PRA-A2-05-006, Rev. 2, ANO-2 Fire PRA Uncertainty / Sensitivity Analysis.

16. Nuclear Management Manual Procedure EN-DC-151, Rev. 2, PSA Maintenance and Update, January 2011.

17. Nuclear Management Manual Procedure EN-DC-161, Rev. 7, Control of Combustibles, November 2012.

18. ASME/ANS RASa-2009 - ASME and ANS combined PRA Standard Standard for Level 1 /

Large Early Release Frequency Probabilistic Risk Assessment.

19. Arkansas Nuclear One - Unit 2, Fire HRA Peer Review Report, issued by Curtiss-Wright Scientech, June 2014.

to 2CAN081401 Updated Attachment W Fire PRA Insights

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-1 W.

Fire PRA Insights W.1 Fire PRA Overall Risk Insights Risk insights were documented as part of the development of the FPRA. The total plant fire CDF/LERF was derived using the NUREG/CR-6850 methodology for FPRA development and is useful in identifying the areas of the plant where fire risk is greatest. The risk insights generated were useful in identifying areas where specific contributors might be mitigated via modification.

A detailed description of significant risk sequences associated with the fire initiating events that contribute above 1% of the calculated fire risk for the plant was prepared for the purposes of gaining these insights and an understanding of the risk significance of MSO combinations.

These insights are provided in Table W-1.

Fire Scenario Selection Fire scenarios were selected based on the definition of significant accident sequence from RG 1.200, Revision 2:

Significant accident sequence: A significant sequence is one of the set of sequences, defined at the functional or systemic level that, when ranked, compose 95% of the CDF or the LERF/LRF, or that individually contribute more than ~1% to the CDF or LERF/LRF.

There are 82 fire scenarios comprising 90% of the cumulative fire CDF and 119 fire scenarios comprising 95% of the cumulative fire CDF. Of these, 24 scenarios contribute 1% or greater on an individual basis. These 24 scenarios (all scenarios contributing 1% or greater on an individual basis) are presented in Table W-1. There is a strong correlation between CDF and LERF. There are six scenarios that contribute 1% or more to the total CDF, but do not similarly contribute to LERF. An additional three scenarios contribute 1% or more to the LERF total, but do not similarly contribute to CDF. These three scenarios have been added to Table W-1.

Modifications Several modifications were identified in the FREs that contributed to reducing CDF and LERF to within the acceptable criteria. The risk benefits of these proposed modifications are reflected in the CDF and LERF risk values presented in Table W-2.

See Attachment S for a complete list of all modifications including additional details of each.

Recovery Actions Recovery actions were reviewed for adverse impact on the FPRA. Each human action credited in the FPRA model was evaluated in the ANO-2 Fire PRA Human Reliability Analysis Notebook (PRA-A2-05-007, Revision 1). None of the modeled actions was found to have an adverse impact on the FPRA. Recovery actions were not credited given a fire in the room in which the action occurs, or through which the operators must pass to perform the action. Also, for the main control room, recovery actions were not credited given a fire in a panel needed to complete the action.

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-2 Safe Shutdown Analysis actions were also reviewed in developing the Variances from Deterministic Requirements (VFDRs) to be used for assessing actions adverse to risk.

The risk associated with each VFDR was evaluated in the FRE process. The potential risk of each recovery action is bounded by the CDF and LERF provided in Table W-2. Also, the additional risk of recovery actions for an area was conservatively determined and is provided in Table W-2.

See Attachment G for the recovery actions credited in each area.

W.2 Risk Change Due to NFPA 805 Transition In accordance with the guidance in Regulatory Position 2.2.4.2 of RG 1.205, Revision 1:

The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions).

The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines. If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk-neutral or represent a risk decrease.

The change in CDF and LERF for each fire area is provided in Table W-2.

W.2.1 Methods Used to Determine Changes in Risk Variances from Deterministic Requirements (VFDRs)

For a fire area that is not deterministically compliant under NFPA 805, Section 4.2.3.2, deterministic compliance strategies were identified as compensatory measures for the variances from the deterministic requirements (VFDRs). These strategies include NRC-granted exemptions, evaluations that determine specific equipment is free of fire damage (see discussion of embedded conduit below), application of electrical raceway fire barrier systems (wrap), or manual manipulations of controls and equipment in the control room and power block.

The VFDRs are lack-of-separation issues and span the NFPA 805 nuclear safety performance criteria: reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries, and process monitoring. The compensatory measures already in place for the VFDRs mitigate the risk of the associated failures, thereby ensuring that the risk from fire damage prior to transition to NFPA 805, although not quantified, is acceptable.

For the transition to NFPA 805, the risk from fire damage is quantified. Each fire area that is not deterministically compliant under NFPA 805, Section 4.2.3.2, is evaluated in the risk-informed, performance-based approach under NFPA 805, Section 4.2.4.2, by the Fire Probabilistic Risk Assessment (FPRA). The FPRA provides a current state-of-the-art fire PRA analysis performed using accepted methods.

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-3 The aggregate change in risk associated with the VFDRs in a given fire area is evaluated by comparing all fire scenarios in the given fire area against two distinct plant models within the FPRA:

the compliant plant model, and the post-transition plant model Compliant Case Analysis for a Fire Area The compliant case for an area represents the existing as-built, as-operated plant if all of the VFDRs in the area were eliminated; in other words, if the area was deterministically compliant.

Thus, the compliant case for each area was analyzed as follows.

The FPRA scenarios for the area were reviewed to determine which VFDR related components are modeled and for those modeled, which VFDR related components are failed by each specific fire scenario. VFDR related components not modeled within the FPRA (e.g., those involving HVAC systems not required to meet PRA success criteria) were determined to have no impact on the safety functions modeled in the FPRA and thus, no contribution to core damage frequency (CDF) or large early release frequency (LERF).

For each scenario, the specific VFDR related components, which if protected would eliminate the VFDR, were set to their random failure probability instead of to failed by the fire. Setting these components to their random failure probability provides an estimate of the fire risk if individual modifications were made to protect or reroute the components, thereby eliminating the VFDRs. The other components in the FPRA model that are impacted by the fire scenario are set to failed by the fire.

Recovery actions (outside control room manual actions to mitigate the direct failure of VFDRs listed in Attachment G) were not credited in the compliant case. Non-recovery actions (manual actions to mitigate non-VFDR failures) were credited in the compliant case. This ensures that the compliant case represents the as-built, as-operated plant except for the eliminated VFDRs in the area, allowing for direct comparison with the post-transition plant model, which credits recovery actions.

As a rule, proposed modifications (listed in Attachment S) were not credited in the compliant case. This ensures that the compliant case represents the as-built, as-operated plant, except for the eliminated VFDRs in the area, allowing for direct comparison with the post-transition plant model, which credits the modifications. One noted exception is Modification S1-3, Backup DC control power to switchgear 2A-1, 2A-2, 2H-1 and 2H-2, which is conservatively credited in both the compliant plant model and the post transition plant model.

Post Transition Case Analysis for a Fire Area The post transition case for a fire area represents the plant if the recoveries listed in Attachment G and the modifications listed in Attachment S are used to protect the plant from core damage, mitigating the risk imposed by the VFDRs.

For each scenario, the specific VFDR related components and other components in the FPRA model that are impacted by the fire scenario were set to failed by the fire. Some examples of the methods used in the post transition case are provided below.

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-4 VFDR components with 1-hour fire wrap; specifically, Service Water pump cables in Fire Areas OO and B-6 and charging pump cables in Fire Area DD were set to failed by the fire. Since these wrapped components were set to failed by the fire in the post transition model, but not in the compliant model, they are included in the risk calculations. Also, since the resulting risk meets the acceptance criteria, the fire wrap does not need to be maintained. See VFDRs B6-01, DD-03 and OO-01 in Attachment C.

VFDR related components with NRC-granted exemptions were set to failed by the fire.

Since these components were set to failed by the fire in the post transition model and the resulting risk meets the acceptance criteria, the NRC-granted exemptions for these components do not need to be maintained in the post transition licensing basis. See Licensing Actions sections of Attachment C for details.

VFDR related components in embedded conduits were not set to failed by the fire since they were determined to be protected from fire damage as documented in Attachment J. Thus, they have no impact on the risk calculations.

Recovery actions (outside control room manual actions to mitigate the direct failure of VFDRs listed in Attachment G) were credited in the post transition case. Non-recovery actions (manual actions to mitigate non-VFDR failures) were also credited in the post transition case.

The proposed modifications (listed in Attachment S) were credited in the post transition case.

This ensures that the post transition case represents the plant following transition to NFPA 805 and allows comparison, with the compliant case which does not credit the modifications. The exception is Modification S1-3, Backup DC control power to switchgear 2A-1, 2A-2, 2H-1 and 2H-2, which is conservatively credited in both the compliant plant model and the post transition plant model.

FPRA model changes to incorporate the proposed modifications (listed in Attachment S) and recovery actions (listed in Attachment G) were made using accepted methods. FPRA peer reviews were performed to assess the adequacy of the FPRA model and the results of the peer reviews are described in Attachment V.

To confirm the availability of operator cues for the recovery actions, the actions were correlated to fire safe shutdown analysis instrumentation. Since one train of fire safe shutdown analysis instrumentation is demonstrated to be available via the conservative deterministic post-fire analysis, these cues will remain available post-fire. Current fire procedures provide guidance to the operators for use of the fire safe shutdown analysis instrumentation as cues for evaluation of the need to perform actions. Confirmation of the availability of operator cues is documented in the ANO-2 Fire PRA Human Reliability Analysis calculations (PRA-A2-05-007, PRA-A2-05-020, and PRA-A2-05-002).

Change in Risk for a Fire Area Each scenario for a fire area is evaluated as described above, with the following exceptions.

If a scenario does not impact VFDR related components, no analysis was performed since the post transition case is equivalent to the compliant case, and the CDF and LERF for that scenario are zero.

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-5 If a scenario contains a VFDR and is low risk (< 1E-08 CDF, < 1E-09 LERF), the scenario CDF and LERF are conservatively approximated by assuming the CDF/LERF for the compliant case is equal to zero. Therefore, the delta risk is bounded by assuming the delta risk is equal to the total risk of the post transition case results for that scenario.

The change in risk for a fire area is computed as the sum of the post-transition plant scenario results minus the sum of the compliant plant scenario results.

ANO-2 Control Room Analysis ANO-2 does not have a Primary Control Station outside the Control Room. Therefore, if Control Room abandonment is required, the current Alternate Shutdown Procedure for Control Room abandonment requires Operations personnel to control the plant via local control of components. Plant monitoring is performed from the Technical Support Center using the Safety Parameter Display System (SPDS).

Post transition to NFPA-805, the plant response to a fire in Fire Area G will differ from the current response. The primary differences are due to the proposed modifications and the insights gained from the fire scenarios analyzed as part of the transition to NFPA-805.

Incorporating the proposed modifications into the Fire PRA analysis allows for analyzing the risk impact of maintaining Reactor Coolant System (RCS) integrity and providing primary-to-secondary heat removal.

The ANO-2 Control Room is one of several fire zones included in Fire Area G. In order to support the transition to NFPA-805, the ANO-2 Control Room abandonment scenario, the non-abandonment scenarios, and additional scenarios associated with other fire zones within Fire Area G have been analyzed to determine the delta risk for the compliant cases vs. post transition cases. The information provided in the previous sections describing the process used for delta risk determination also applies to the methodology used for Fire Area G. The VFDRs for Fire Area G are provided in Attachment C.

In order to calculate the compliant cases for Fire Area G, the pertinent VFDRs were identified that would be affected in each of the scenarios (i.e. both abandonment and non-abandonment).

This process allowed for a compliant case to be developed and analyzed for each of the Fire Area G scenarios similar to other fire areas.

The Post Transition case was analyzed by failing the components affected by the fire and using the Fire PRA model with modifications and recoveries as necessary to determine the risk.

In addition to the proposed modifications and recoveries identified as part of the risk analysis, additional Defense in Depth actions, which are listed in Attachment G, have been identified to enhance plant control and reduce the likelihood that additional equipment is damaged due to spurious operation.

Additional Risk of Recovery Actions In the fire area risk evaluations, credit was taken for plant modifications in addition to the credited recovery actions to meet the acceptance criteria. These proposed modifications were developed and scoped to reduce risk. One modification in particular, the newly proposed

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-6 Auxiliary Feedwater (AFW) pump, is a significant modification and has been developed to be implemented with more reliable and redundant power supplies. Additionally, this proposed modification will have fewer dependencies than the currently configured Emergency Feedwater (EFW) pumps. The design of the AFW pump will be developed to be more reliable than the existing configuration for the EFW systems when mitigating a post-fire transient. The risk reduction, after implementing the AFW pump in the PRA, results in a significant decrease in plant risk. This large decrease in risk enveloped the positive risk increase of the recovery actions and resulted in an overall negative delta CDF and LERF.

Regulation Guide 1.205, Section 2.2.4.1 (Revision 1), requires that the risk increase associated with the recovery actions be reported to the NRC as part of the license amendment request (LAR). Therefore, the risk increase from crediting the identified recovery actions, which is enveloped in the overall negative delta CDF/LERF of the fire risk evaluations, was determined as described below.

The additional risk of recovery actions for an area was conservatively determined by calculating the difference in CDF and LERF using the post transition case model (with modifications incorporated into the model) in accordance with the guidance in FAQ 07-0030 Establishing Recovery Actions.

The difference between the two cases of the same scenario (one recovery set to its nominal HEP values, and the other set to zero failure probability) provides an estimated, yet bounding, evaluation of the change in risk of the recovery actions in the area. By starting with the post-transition model, the modifications are credited in both cases of the risk of recovery analysis. This method of analysis removes the modification offset reported in the fire risk evaluation CDF and LERF in Table W-2.

W.2.2 Risk Acceptance Criteria The change in CDF and LERF for each fire area is provided in Table W-2.

Total Change in CDF and LERF The total change in CDF for this application is calculated to be -1.29E-04/yr (the sum of the calculated delta risk from Table W-2) and the total change in LERF is calculated to be

-4.72E-06/yr. These values include credited recovery actions and plant modifications (documented in Attachments G and S, respectively). These changes in the plant CDF and LERF meet the RG 1.174 criteria as the total change in risk associated with the transition to NFPA 805 results are well within the acceptance criteria and the total plant fire risk is below 1E-04/yr for CDF and 1E-05/yr for LERF.

Site Risk from Internal Events Although RG 1.174 does not require calculation of total CDF and LERF, if the increases are below the delta CDF and delta LERF of 1E-06/yr and 1E-07/yr respectively, it does recommend that if there is an indication that the CDF is considerably higher than 1E- 04/yr or if LERF is considerably higher than 1E-05/yr, then the focus should be on finding ways to decrease CDF or LERF.

The total CDF including Fire and Internal events has a value of 7.7E-05/yr (Internal Events CDF (9.5E-7/year) + Internal Floods (8.0E-07/yr) + Fire CDF (7.5E-05)), and the total LERF has a value of 1.9E-06/yr (Internal Events LERF (1.1E-07/year) + Internal Floods (5.6E-08/yr) + Fire LERF (1.7E-06/yr)). Both values are below the RG 1.174 criteria of 1E-04/yr (CDF) and 1E-05/yr (LERF).

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-7 The aforementioned total CDF and LERF values do not include contribution from external events. Therefore, the contribution to risk from external events is captured below.

Site Risk from External Events Seismic - The Operating Basis Earthquake for ANO is 0.1g and the Design Basis Earthquake for ANO is 0.2g. As part of the IPEEE submittal, ANO-2 performed a Seismic Margin Analysis (SMA). The results of the walkdowns that were performed as part of the SMA verified that the equipment, tanks, distribution systems, structures, and relays are able to withstand the 0.3g Review Level Earthquake at the plant and still provide for its safe shutdown. Based on an updated seismic hazard curve provided by EPRI, the likelihood of a seismic event exceeding 0.3 g peak ground acceleration is 9.28E-06/yr. Given the low seismic frequency with no seismic design outliers, the seismic CDF is estimated to be well below 1E-5/yr and LERF is estimated to be well below 1E-6/yr.

Flooding and other External Events - High winds, floods, or off-site industry facility accidents do not contribute significantly to ANO-2 site risk. For the external events 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.

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 1E-5/yr. A total bounding estimate for LERF external events is assumed to be 0.1 of the total CDF, which is less than 1E-6/yr.

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-8 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2199-G/A Post Transition Baseline Case 16.44%

16.44%

Fire scenario 2199-G/A is the control room abandonment scenario. This scenario assumes fire damage to all cable targets in the control room. No in-control room actions are credited. Top cutsets are associated with the ex-control room actions to trip the reactor coolant pumps (RCPs), isolate letdown, and start the new AFW pump at the local control station. The local control station for the AFW pump will include local Steam Generator (SG) instrumentation to assist the operators in shutting down the plant following an abandonment scenario.

8.20E-05 1.51E-01 1.24E-05 1.55E-07 2099-W/HGL Base scenario -

Severe Fire 3.95%

20.39%

Fire scenario 2099-W/HGL is a base scenario representing a severe fire in the West DC Equipment room, which impacts all targets in the room. Both the turbine-driven and motor-driven EFW pumps are unavailable. The new DC modification is credited to provide the DC power needed to trip the RCPs from the control room. The top cutsets are associated with random failure to trip RCPs in the control room resulting in an RCP seal Loss of Coolant Accident (LOCA) due to the Engineered Safety Feature (ESF) signal induced isolation of Component Cooling Water (CCW) to the RCP seals. High Pressure Safety Injection (HPSI) is unavailable due to spurious Recirculation Actuation Signal (RAS) prior to sufficient inventory in the containment sump and is not available for LOCA mitigation or feed and bleed. The top cutsets also include random failure of the new AFW pump and the existing AFW pump (2P-75) to provide cooling to the SGs.

4.98E-04 5.98E-03 2.98E-06 7.40E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-9 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2101-AA/HGL Base scenario 3.29%

23.67%

Fire scenario 2101-AA/HGL represents the severe fire contribution for all fires in the North Switchgear room. This fire scenario is assumed to impact all targets in the zone, which results in failure of vital 4 kV bus 2A-3 (red train power) and components fed by that bus. Both the turbine-driven EFW pump (2P-7A) and the new AFW pump are available for secondary heat removal.

6.44E-04 3.85E-03 2.48E-06 8.15E-08 2199-G-AJ/A 2C33, 2C33-1, 2C33-2 3.14%

26.81%

Fire scenario 2199-G-AJ/A is a fire impacting the Service Water (SW) and Boron Management vertical board panel 2C33-1 and 2C33-2 in the Main Control Room (MCR). No divisional separation is credited for a fire in this panel. Failure of breaker interlocks can result in a loss of offsite power and loss of Emergency Diesel Generators (EDGs). The new AFW pump has an independent power supply and is available to provide feed to the SGs. RCS integrity is maintained following an in-control room trip of the RCPs before seal damage occurs. HPSI is unavailable due to SW (for HPSI pump room cooler) and power supply failures.

1.38E-04 1.72E-02 2.37E-06 5.67E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-10 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2150-C/A Base scenario -

Severe Fire 2.94%

29.76%

Fire scenario 2150-C/A is a base scenario in which all components in the Core Protection Calculator (CPC) room are impacted. The primary failures in this scenario are associated with failure of ESF Actuation System (ESFAS) due to fire-induced failure of the SG level and pressure instrumentation.

HPSI is unavailable due to the potential for a spurious RAS resulting from fire-induced failure of the Refueling Water Tank (RWT) level instrumentation and the power supply to the Plant Protection System (PPS). The resulting top cutsets are associated with random failure to trip RCPs in the control room resulting in an RCP seal LOCA due to assumed loss of CCW. HPSI is unavailable due to fire related effects and is not available for LOCA mitigation. The new AFW pump is available and the EFW pumps are available with an action for EFAS override.

3.96E-04 5.61E-03 2.22E-06 5.44E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-11 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2136-I-TN/A Transient Base Scenario 2.89%

32.65%

Fire scenario 2136-I-TN/A is a very large transient fire due to the potential for a large amount of intervening combustible material to be stored within this health physics and corridor zone. The entire zone is considered a designated storage area in which the combustible material could exceed the NUREG/CR-6850 317 kW fire. As a result, conservative fire modeling assumes instantaneous full zone burnout. Both the turbine-driven and motor-driven EFW pumps are unavailable. Offsite power and diesel generators are impacted in this fire area. The new DC modification is credited to provide the DC power needed to trip the RCPs from the control room. The top cutsets are associated with random failure to trip RCPs in the control room resulting in an RCP seal LOCA due to assumed loss of CCW. HPSI is unavailable due to fire related effects and is not available for LOCA mitigation or feed and bleed. The new AFW pump is the only source of feedwater credited in the scenario.

1.90E-04 1.14E-02 2.18E-06 7.48E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-12 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2109-U-B1/A 2D35 2.76%

35.40%

Fire scenario 2109-U-B1/A is a fire starting in battery eliminator (charger) 2D-35 which damages conduits and targets up to, and including, the nearest tray target above the panel. This scenario results in loss of 2P-7A discharge motor-operated valves (MOVs)

(IN 92-18) without the ability to reposition.

The motor-driven EFW pump (2P-7B) and existing AFW pump (2P-75) are failed due to air operated valve (AOV) 2CV-0714-1 (flow diversion valve to startup and blow-down demineralizer system). 2CV-0714-1 can be isolated from the control room, but this recovery action has conservatively not been credited. Therefore, the scenario relies on the new AFW pump to provide secondary heat removal. The RCS integrity is maintained with HPSI available to provide RCS inventory loss mitigation.

1.88E-04 1.11E-02 2.08E-06 6.96E-08 2199-G-O/A 2C17 2.49%

37.90%

Fire scenario 2199-G-O/A is a fire impacting ESF and PPS vertical board panel 2C17 in the MCR. The new AFW pump provides the main source of feedwater and the existing AFW pump is available via manual action.

The EFW flow path is available via a recovery action for the AC operated EFW valve that is being modified to remove the IN 92-18 concerns (Attachment S, Modification S1-7).

RCS integrity is maintained following an in-control room trip of the RCPs before seal damage occurs. HPSI is unavailable due to failure of the cold leg injection path from headers A and B.

1.42E-04 1.32E-02 1.88E-06 4.01E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-13 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2137-I-TN1/A Transient Base Scenario 2.43%

40.32%

Fire scenario 2137-I-TN1/A is a designated storage area (DSA) fire that burns out the entire compartment in the Upper South Electrical Penetration (USEP) room. The fire modeling is consistent with the approach described in scenario 2136-I-TN/A. The new AFW pump is the only pump available to provide secondary heat removal. The top cutsets are associated with random failure to trip RCPs in the control room resulting in an RCP seal LOCA due to assumed loss of CCW. HPSI is unavailable due to fire related effects and is not available for LOCA mitigation or feed and bleed.

1.56E-04 1.17E-02 1.83E-06 6.19E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-14 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2100-Z/HGL Base Scenario 2.35%

42.67%

Fire scenario 2100-Z/HGL represents the severe fire contribution for all fires in the South Switchgear room. This fire scenario is assumed to impact all targets in the zone, which results in failure of vital 4 kV bus 2A-4 (green train power) and components fed by that bus. Also, cables impacted in 2100-Z could result in loss of red train SW. Red train service water could be lost due to a spurious hot short causing the red train sluice gate (2CV-1470-1) to transfer closed or loss of control of the Auxiliary Cooling Water (ACW) isolation valves causing a flow diversion.

Loss of red train service water SW would subsequently result in failure of the red train EDG. The new AFW pump is unaffected as it has an independent power supply, and the motor-driven EFW pump (2P-7B) is available, but manual action to open the green train powered isolation MOV may be required if that valve is closed at the time green train power is lost. RCS integrity is maintained following an in-control room trip of the RCPs before seal damage occurs. HPSI is available for HPSI recirc mode with action to mitigate SW failures.

6.82E-04 2.59E-03 1.77E-06 4.71E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-15 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2091-BB/HGL Base scenario -

Severe Fire 1.80%

44.47%

Fire Scenario 2091-BB/HGL is a severe fire contribution in the North Electrical Equipment room. In this scenario, all cable targets in 2091-BB are damaged. The motor-driven EFW pump (2P-7B) and the new AFW pump are available to provide feed to the SG.

2P-7B is available with operator action to open the EFW discharge MOVs after a fire-induced loss of power. The top risk contributing cutsets include random failures of the new AFW pump and the motor-driven EFW system. RCP seal LOCA is a concern since CCW is assumed lost, but HPSI is available for LOCA mitigation.

4.41E-04 3.09E-03 1.36E-06 4.48E-08 2076-HH/A Base scenario 1.68%

46.15%

Fire scenario 2076-HH/A is the Severe Fire contribution for all ignition sources in electrical equipment room 2076-HH resulting in all targets being damaged within the fire zone. The motor-driven EFW pump and the new AFW pump are available to provide secondary heat removal. Top cutsets are associated with failure of the manual control room action to trip the RCPs since CCW is assumed lost. HPSI is available for feed and bleed and LOCA mitigation.

2.44E-03 5.20E-04 1.27E-06 4.09E-08 2108-S/HGL Base scenario 1.50%

47.65%

Fire scenario 2108-S/HGL is the Severe Fire contribution for all ignition sources in electrical equipment room 2108-S resulting in all targets being damaged in Fire Zone 2108-S. The existing AFW pump (2P-75) and the new AFW pump are credited to provide secondary heat removal. The DC modification is credited to provide DC power to trip the RCPs from the control room since CCW is assumed lost.

Fire-induced failures result in loss of HPSI for LOCA mitigation.

3.09E-04 3.66E-03 1.13E-06 2.23E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-16 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2068-DD/A Base scenario 1.47%

49.12%

Fire scenario 2068-DD/A is a full zone burn out of the Hot Machine Shop conservatively impacting all cables in the fire zone, many of which are routed in the overhead. The motor-driven EFW pump (2P-7B) and the new AFW pump are available to provide secondary heat removal. Top cutsets are associated with failure of the manual control room action to trip the RCPs since CCW is assumed lost.

Fire-induced failures also result in loss of HPSI for LOCA mitigation.

4.80E-04 2.31E-03 1.11E-06 1.27E-08 2073-DD-F/A 2B-62 1.42%

50.54%

Fire Scenario 2073-DD-F/A is a fire in motor control center (MCC) 2B-62. The EFW, existing AFW and new AFW pumps are available to provide secondary heat removal.

Top cutsets are associated with failure of the manual control room action to trip the RCPs since CCW is assumed lost. Fire-induced failures result in loss of HPSI injection valves for LOCA mitigation.

4.71E-04 2.28E-03 1.07E-06 1.18E-08 2199-G-B/A 2C04 (incl.

2C01, 2C02, 2C03, 2C09) and adj panels 2C100 1.40%

51.95%

Fire Scenario 2199-G-B/A is the Main Control Board (MCB) fire in the MCR. An Appendix L factor is credited using a zero distance from Figure L-1 in NUREG/CR-6850. The top cutsets are associated with the ex-control room actions to trip the RCPs and to isolate letdown, as well as with the in-control room action to use the new AFW pump and random failures of the new AFW pump.

7.03E-06 1.50E-01 1.06E-06 1.32E-08 2109-U-H/A 2B-51 1.26%

53.21%

Fire scenario 2109-U-H/A is a fire in MCC 2B-51. The motor-driven EFW and new AFW pumps are available to provide secondary heat removal. Top cutsets are associated with failure of the manual control room action to trip the RCPs since CCW is assumed lost.

Fire-induced hot short failures could result in loss of HPSI for LOCA mitigation.

5.56E-04 1.71E-03 9.52E-07 3.15E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-17 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2200-MM-CW/A 2A-1, 2A-2 1.26%

54.47%

Fire scenario 2200-MM-CW/A is a fire at switchgear 2A1 and 2A2 in the turbine building. This scenario results in a fire-induced Loss of Offsite Power (LOSP) and credits the diesel generators for providing power to the vital red and green train 4kV switchgear (2A3 and 2A4). The new DC modification will provide trip capability of the RCPs if this fire precludes access to 6.9 kV switchgear 2H-1 and 2H-2. The EFW, existing AFW and new AFW pumps are available for feedwater. Also, the redundant power supply to the new AFW pump is independent of the EDGs and unaffected by fire impact to the offsite power source from 2A-1. Top cutsets are associated with failure of the manual control room action to trip the RCPs since CCW is assumed lost. HPSI is available provided the emergency diesel generators load properly.

1.30E-03 7.32E-04 9.51E-07 2.39E-08 2111-T/HGL Base scenario 1.25%

55.72%

Fire scenario 2111-T/HGL is the Severe Fire contribution for all ignition sources in the Lower South Electrical Penetration (LSEP) room resulting in all targets being damaged within the fire zone. The new AFW pump and existing AFW pump are credited for secondary heat removal. The motor-driven and turbine-driven EFW pumps are failed due to fire-induced failures. Top cutsets are associated with failure of the manual control room action to trip the RCPs since CCW is assumed lost. HPSI is available via EDG power.

1.94E-04 4.87E-03 9.44E-07 2.15E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-18 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2199-G-N/A 2C16 1.22%

56.94%

Fire scenario 2199-G-N/A is a fire impacting ESF and PPS vertical board panel 2C16 in the MCR. The new AFW pump provides the main source of feedwater but the existing AFW pump is available with manual action.

The AFW flow path is through the AC operated EFW discharge lines that are being modified to remove the IN 92-18 concerns (see Attachment S, Modification S1-7). RCS integrity is maintained following an in-control room trip of the RCPs before seal damage occurs. HPSI is unavailable due to failure of the cold leg injection and suction paths from headers A and B.

9.49E-05 9.73E-03 9.23E-07 1.54E-08 2098-C/HGL Base scenario -

Severe Fire 1.18%

58.13%

Fire scenario 2098-C/HGL is the Severe Fire contribution for all ignition sources in the CPC room resulting in all targets being damaged within the fire zone. The HALON suppression system is credited for the CPC panel fire and this scenario represents the failure of the HALON system to actuate before additional cable target damage. In this scenario, only the new AFW pump is available to provide secondary heat removal. Actions to trip the RCPs and isolate letdown are credited to maintain RCS integrity and HPSI is unavailable due to fire-induced failures.

7.62E-05 1.17E-02 8.93E-07 3.02E-08 2040-JJ-Y/A 2B-52 1.16%

59.29%

Fire scenario 2040-JJ-/AY is a fire in MCC 2B-52. The EFW, existing AFW, and new AFW pumps are available to provide secondary heat removal. Top cutsets are associated with failure of the manual control room action to trip the RCPs since CCW is assumed lost. Fire-induced failures result in loss of HPSI injection valves for LOCA mitigation.

3.85E-04 2.28E-03 8.76E-07 9.61E-09

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-19 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2101-AA-H-NS/A 2A3 1.11%

60.40%

Fire scenario 2101-AA-H-NS/A represents a fire at the 4 kV vital bus 2A-3. The risk insights are the same as in 2101-AA/HGL described above.

2.19E-04 3.83E-03 8.39E-07 2.76E-08 2199-G-M/A 2C14 1.03%

61.43%

Fire scenario 2199-G-M/A is a fire impacting the Plant Auxiliary Systems vertical board panel 2C14 in the main control room. The new AFW pump provides the main source of feedwater, but EFW and the existing AFW pumps are available with manual override following a spurious Main Steam Isolation Signal (MSIS) signal. RCS integrity is maintained following an in-control room trip of the RCPs before seal damage occurs. HPSI is unavailable due to a spurious RAS signal.

9.49E-05 8.23E-03 7.80E-07 1.05E-08 2073-DD-TN5/A1 Transient Base Scenario 0.94%

62.37%

Fire scenario 2073-DD-TN5/A is a designated storage area transient fire in Fire Zone 2073-DD. This scenario burns the entire compartment. The new AFW pump is the only pump available for secondary heat removal. Top cutsets are associated with failure of the manual control room action to trip the RCPs since CCW is assumed lost.

5.53E-05 1.28E-02 7.08E-07 2.43E-08 2033-K/A1 Base scenario 0.82%

63.19%

Fire scenario 2033-K/A represents a large fire within containment, North Side. No walk-downs were performed for the scenario development and the high risk targets were identified from drawing references. The primary failures include failures to SG and RCS instrumentation. The new AFW pump remains available for feedwater injection and HPSI is available with a manual override of a spurious Safety Injection Actuation Signal (SIAS) signal.

2.40E-03 2.57E-04 6.18E-07 2.05E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-20 Table W-1 Fire PRA CDF & LERF Significant Fire Initiating Events (Individually Representing > 1% of the Calculated CDF)

Scenario Description Contribution Risk Insights IGF CCDP CDF LERF

% of Total Cumulative (rx/yr) 2136-I/HGL1 Base scenario 0.67%

63.86%

Fire scenario 2136-I/HGL represents the severe fire contribution from the fixed ignition sources in Fire Zone 2136-I. The risk insights are the same as those for 2136-I-TN/A described above.

4.44E-05 1.14E-02 5.08E-07 1.75E-08 (1) - Sequences included due to LERF contribution. These sequences contribute less than 1% of CDF, but account for 1% or more of the LERF total.

CCDP - Conditional Core Damage Probability IGF - Ignition Frequency (includes severity factor and probability of non-suppression, where applicable)

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-21 Table W-2 ANO-2 Fire Area Risk Summary Fire Area Area Description NFPA 805 Basis Fire Area CDF Fire Area LERF VFDR (Yes/No) RAs Fire Risk Eval. CDF Fire Risk Eval. LERF Additional Risk of RAs (CDF/LERF) 2MH01E concrete manhole east 4.2.4.2 9.62E-09 1.42E-10 Yes No 9.62E-09 1.42E-10 N/A 2MH02E concrete manhole east 4.2.4.2 9.62E-09 1.42E-10 Yes No 9.62E-09 1.42E-10 N/A 2MH03E concrete manhole east 4.2.4.2 3.13E-07 9.76E-09 Yes No

-5.05E-06

-1.69E-07 N/A 2MH01W concrete manhole west 4.2.3.2 2.95E-08 7.47E-10 No N/A N/A N/A N/A 2MH02W concrete manhole west 4.2.3.2 2.95E-08 7.47E-10 No N/A N/A N/A N/A 2MH03W concrete manhole west 4.2.3.2 2.95E-08 7.47E-10 No N/A N/A N/A N/A AA Fire Zone 2007-LL (B HPSI, LPSI, and Containment Spray Pump room and gallery) 4.2.4.2 4.49E-07 5.21E-09 Yes No

-1.29E-07

-1.38E-08 N/A AAC Fire Zones SBOD and 2MH12 (alternate AC diesel and nearby manhole) 4.2.3.2 3.50E-08 7.78E-10 No N/A N/A N/A N/A Admin administration building 4.2.3.2 n/a n/a No N/A N/A N/A N/A B-2 miscellaneous turbine building fire compartments 4.2.4.2 3.84E-06 9.49E-08 Yes Yes

-2.94E-05

-1.01E-06 7.73E-07 /

1.15E-08 B-3 Fire Zones 2091-BB, 2112-BB and 2183-J (electrical penetration rooms) 4.2.4.2 1.47E-06 4.77E-08 Yes Yes 5.87E-07 1.89E-08 4.90E-07 /

1.64E-08 B-4 Fire Zone 2154-E (CEDM equipment room) 4.2.4.2 2.89E-07 4.30E-09 Yes No

-8.39E-06

-2.48E-07 N/A B-5 Fire Zones 2149-B and 2158-F (stairwells 2001 and 2055) 4.2.3.2 4.14E-09 1.13E-10 No N/A N/A N/A N/A

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-22 Table W-2 ANO-2 Fire Area Risk Summary Fire Area Area Description NFPA 805 Basis Fire Area CDF Fire Area LERF VFDR (Yes/No) RAs Fire Risk Eval. CDF Fire Risk Eval. LERF Additional Risk of RAs (CDF/LERF)

B-6 Fire Zones 2006-LL, 2010-LL, 2011-LL, and 2014-LL (general access, C HPSI pump room, tendon gallery access, and A HPSI, LPSI and Containment Spray Pump room) 4.2.4.2 2.85E-07 3.54E-09 Yes No

-4.51E-08

-1.59E-09 N/A CC Fire Zone 2024-JJ (turbine-driven emergency feedwater pump room) 4.2.3.2 3.21E-09 9.63E-11 No N/A N/A N/A N/A DD Fire Zones 2019-JJ, 2032-JJ, 2040-JJ, and 2068-DD (boric acid condensate tank room, spent resin storage tank room, corridor, and hot machine shop) 4.2.4.2 2.45E-06 3.13E-08 Yes No

-1.84E-06

-6.72E-08 N/A EE-L Fire Zones 2055-JJ and 2084-DD (piping penetration rooms) 4.2.4.2 4.95E-07 1.24E-08 Yes No

-6.83E-08

-2.18E-09 N/A EE-U Fire Zone 2111-T (lower south electrical penetration room) 4.2.4.2 9.49E-07 2.17E-08 Yes No

-1.82E-06

-6.11E-08 N/A FF Fire Zone 2025-JJ (motor-driven emergency feedwater pump room) 4.2.3.2 1.62E-07 5.37E-09 No N/A N/A N/A N/A G

Fire Zones 2199-G, 2119-H, 2136-I, 2137-I, 2150-C, 2098-C, 2098-L, 129-F, and 97-R (control room and other alternate shutdown areas) 4.2.4.2 3.88E-05 7.74E-07 Yes Yes

-3.92E-05

-1.66E-06 1.76E-05 /

2.88E-07 GG Fire Zones 2076-HH and 2081-HH (electrical equipment room and upper north and lower north piping penetration room) 4.2.4.2 1.65E-06 4.52E-08 Yes Yes

-3.87E-06

-1.37E-07 1.15E-06 /

3.79E-08

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-23 Table W-2 ANO-2 Fire Area Risk Summary Fire Area Area Description NFPA 805 Basis Fire Area CDF Fire Area LERF VFDR (Yes/No) RAs Fire Risk Eval. CDF Fire Risk Eval. LERF Additional Risk of RAs (CDF/LERF)

HH Fire Zones 2063-DD, 2072-R, 2073-DD, 2096-M, 2106-R, and 2107-N (sample room, VCT room, 2B-62 room, 2B-63 room, degasifier vacuum pump room, and corridor) 4.2.4.2 3.60E-06 8.62E-08 Yes No

-1.01E-06

-2.59E-08 N/A II Fire Zone 2101-AA (north switchgear 2A-3 room) 4.2.4.2 3.93E-06 1.29E-07 Yes No

-2.12E-05

-7.17E-07 N/A JJ Fire Zone 2109-U (corridor) 4.2.4.2 4.57E-06 1.36E-07 Yes Yes

-1.31E-06

-4.22E-08 9.97E-07 /

3.31E-08 K

Fire Zones 16-Y and 2020-JJ (clean waste receiver tank room and boron holdup tank vault) 4.2.3.2 6.13E-10 1.45E-11 No N/A N/A N/A N/A KK Fire Zones 2093-P, 2114-I, and 2115-I (south EDG room, EDG air intake room, and boric acid makeup tank room) 4.2.4.2 9.64E-08 2.93E-09 Yes No N/A L

Fire Zone TKVLT (diesel fuel storage vault) 4.2.3.2 2.70E-08 6.69E-10 No N/A N/A N/A N/A MM Fire Zones 2099-W and 2103-V (west DC equipment room and west battery room) 4.2.4.2 3.02E-06 7.49E-08 Yes No

-1.76E-06

-8.40E-08 N/A NN Fire Zones 2032-K and 2033-K (containment building south side and containment building north side) 4.2.4.2 1.53E-06 4.34E-08 Yes No N/A OO Intake Structure 4.2.4.2 3.01E-07 3.76E-09 Yes Yes 2.61E-07 2.48E-09 7.80E-09 /

1.64E-10

Arkansas Nuclear One - Unit 2 Att. W - Fire PRA Insights to 2CAN121202 Page W-24 Table W-2 ANO-2 Fire Area Risk Summary Fire Area Area Description NFPA 805 Basis Fire Area CDF Fire Area LERF VFDR (Yes/No) RAs Fire Risk Eval. CDF Fire Risk Eval. LERF Additional Risk of RAs (CDF/LERF)

QQ Fire Zones 2094-Q and 2114-I (north EDG room and EDG air intake room) 4.2.3.2 1.99E-07 6.31E-09 No N/A N/A N/A N/A SS Fire Zones 2097-X, 2100-Z and 2102-Y (east DC equipment room, south switchgear room and east battery room) 4.2.4.2 2.83E-06 8.07E-08 Yes Yes

-9.84E-06

-3.41E-07 2.64E-06 /

7.60E-08 TT Fire Zone 2108-S (electrical equipment room) 4.2.4.2 1.17E-06 2.30E-08 Yes Yes

-4.70E-06

-1.58E-07 4.02E-07 /

1.37E-08 YD YARD 4.2.3.2 2.02E-06 5.78E-08 No N/A N/A N/A N/A Various Unit 1 - specific fire areas1 4.2.3.2 8.64E-07 2.02E-08 No N/A N/A N/A N/A TOTAL 7.54E-05 1.72E-06

-1.29E-04

-4.72E-06 2.41E-05 /

4.77E-07 Indicative of an immeasurable change in risk from the impact of the VFDR on Fire PRA model.

1 ANO-1-specific fire areas were conservatively assessed to contribute to ANO-2 CDF/LERF. Fires in these areas typically do not impact circuits for ANO-2 components and are not expected to cause, or require, an ANO-2 plant trip. (Fires in the ANO-1 control room and cable spreading room are not included in this value since they are included in the Fire Area G results.)

2 The Fire PRA quantification results for CDF, LERF, CDF/LERF are documented in PRA-A2-05-021.

References

1.

Echelon Calculation PRA-A2-05-002, Rev. 0, ANO-2 Fire PRA New Human Failure Events, October 2011.

2.

Echelon Calculation PRA-A2-05-007, Rev. 1, ANO-2 Fire Probabilistic Risk Assessment Human Reliability Analysis (HRA) Notebook, July 2014.

3.

Echelon Calculation PRA-A2-05-020, Rev. 0, Detailed Fire Human Reliability Analysis (HRA) for Selected ANO-2 Human Failure Events (HFE), July 2014.

4.

PRA-A2-05-021, Rev. 0, Fire PRA Quantification Changes to Support Attachment W of the License Amendment Request (LAR).